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{{#Wiki_filter:ATTACHMENT 4 TO TXX-15101EMERGENCY ACTION LEVELTECHNICAL BASES(REDLINE AND STRIKEOUT VERSION)(276 PAGES) | |||
LuminantComanche Peak Nuclear Power PlantEPP-201Emergency Action Level Technical Bases DocumentPrepared by:Print NameSignatureDateTechnical Reviewer:Print NameSignatureDateReviewer:Print NameSignatureDateApproval:Print NameSignatureDateEffective Date:DRAFT D2 6/24/15I Page 1 of 276 I TABLE OF CONTENTSSECTION PAGE1.0 PURPOSE .............................................................................................................................. 32.0 DISCUSSION .......................................................................................................................... 32.1 Background ......................................................................................................................... 32.2 Fission Product Barriers .................................................................................................. 42.3 Fission Product Barrier Classification Criteria ................................................................ 42.4 EAL Organization ........................................................................................................ 52.5 Technical Bases Information ........................................................................................... 72.6 Operating M ode Applicability ......................................................................................... 82.7 Unit Designation .......................................................................................................... 83.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS .......................................... 93.1 General Considerations .................................................................................................. 93.2 Classification Methodology ........................................................................................... 1 | |||
==04.0 REFERENCES== | |||
..................................................................................................................... 134.1 Developm ental ................................................................................................................... 134.2 Im plementing ..................................................................................................................... 135.0 DEINITIONS, ACRONYMS & ABBREVIATIONS ............................................................. 146.0 CPNPP TO NEI 99-01 Rev. 6 EAL CROSS-REFERENCE .............................................. 207.0 ATTACHM ENTS ................................................................................................................... 241 Emergency Action Level Technical Bases ........................................................... 25Cate-gory R Abnormal Rad Release / Rad Effluent .................................... 25Category E ISFSI ...................................................................................... 65Cate-gory C Cold Shutdown / Refueling System Malfunction ...................... 68Category H Hazards ...................................................................................... 112Cate-gory S System Malfunction ........... ; ........................................................ 155Cate-gory F Fission Product Barrier Degradation .......................................... 2002 Fission Product Barrier Loss / Potential LossMatrix and Bases .................................................................................................... 2053 Safe Operation & Shutdown Areas Tables R-3 & H-2 Bases ................................. 258Page 2 of 276 1.0 PURPOSEThis document provides an explanation and rationale for each Emergency Action Level (EAL)included in the EAL Upgrade Project for Comanche Peak Nuclear Power Plant (CPNPP). Itshould be used to facilitate review of the CPNPP EALs and provide historical documentationfor future reference. Decision-makers responsible for implementation of EPP-201,"Assessment of Emergency Action Levels, Emergency Classification and Plan Activation,"may use this document as a technical reference in support of EAL interpretation. Thisinformation may assist the Emergency Coordinator in making classifications, particularly thoseinvolving judgment or multiple events. The basis information may also be useful in trainingand for explaining event classifications to off-site officials.The expectation is that emergency classifications are to be made as soon as conditions arepresent and recognizable for the classification, but within 15 minutes or less in all cases ofconditions present. Use of this document for assistance is not intended to delay theemergency classification.Because the information in a basis document can affect emergency classification decision-making (e.g., the Emergency Coordinator refers to it during an event), the NRC staff expectsthat changes to the basis document will be evaluated in accordance with the provisions of 10CFR 50.54(q).2.0 DISCUSSION2.1 BackgroundEALs are the plant-specific indications, conditions or instrument readings that are utilized toclassify emergency conditions defined in the CPNPP Plant Radiological EmergencyResponse Plan (RERP).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) Revisions 4 and 5 were 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 address conditionsthat may be postulated to occur at permanently Defueled Stations and IndependentSpent Fuel Storage Installations (ISFSIs).* Simplifying the fission product barrier EAL threshold for a Site Area Emergency.Subsequently, Revision 6 of NEI 99-01 has been issued which incorporates resolutions tonumerous implementation issues including the NRC EAL Frequently Asked Questions (FAQs).Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Action Levelsfor Non-Passive Reactors," November 2012 (ADAMS Accession Number ML12326A805) (ref.4.1.1), CPNPP conducted an EAL implementation upgrade project that produced the EALsdiscussed hereinPage 3 of 276 2.2 Fission Product BarriersFission product barrier thresholds represent threats to the defense in depth design conceptthat precludes the release of radioactive fission products to the environment. This conceptrelies on multiple physical barriers, any one of which, if maintained intact, precludes therelease of significant amounts of radioactive fission products to the environment.Many of the EALs derived from the NEI methodology are fission product barrier thresholdbased. That is, the conditions that define the EALs are based upon thresholds that representthe loss or potential loss of one or more of the three fission product barriers. "Loss" and"Potential Loss" signify the relative damage and threat of damage to the barrier. A "Loss"threshold means the barrier no longer assures containment of radioactive materials. A"Potential Loss" threshold implies an increased probability of barrier loss and decreasedcertainty of maintaining the barrier.The primary fission product barriers are:A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains thefuel pellets.B. Reactor Coolant System (RCS): The RCS Barrier includes the RCS primary side and itsconnections up to and including the pressurizer safety and relief valves, and otherconnections up to and including the primary isolation valves.C. Containment (CNTMT): The Containment Barrier includes the containment building andconnections up to and including the outermost containment isolation valves. This barrieralso includes the main steam, feedwater, and blowdown line extensions outside thecontainment building up to and including the outermost secondary side isolation valve.Containment Barrier thresholds are used as criteria for escalation of the ECL from Alertto a Site Area Emergency or a General Emergency2.3 Fission Product Barrier Classification CriteriaThe following criteria are the bases for event classification related to fission product barrierloss or potential loss:Alert:Any loss or any potential loss of either Fuel Clad or RCS barrierSite 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 4 of 276 2.4 EAL OrganizationThe CPNPP EAL scheme includes the following features:0 Division of the EAL set into three broad groups:o EALs applicable under any 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, Hot Standby,Startup, or Power 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, Refueling orDefueled mode.The purpose of the groups is to avoid review of hot condition EALs when the plant is ina cold condition and avoid review of cold condition EALs when the plant is in a hotcondition. This approach significantly minimizes the total number of EALs that must bereviewed by the EAL-user for a given plant condition, reduces EAL-user reading burdenand, thereby, speeds identification of the EAL that applies to the emergency.0 Within each group, assignment of EALs to categories and subcategories:Category and subcategory titles are selected to represent conditions that are operationallysignificant to the EAL-user. The CPNPP EAL categories are aligned to and represent the NEI99-01"Recognition Categories." Subcategories are used in the CPNPP scheme as necessaryto further divide the EALs of a category into logical sets of possible emergency classificationthresholds. The CPNPP EAL categories and subcategories are listed below.Page 5 of 276 EAL Groups, Categories and SubcategoriesEAL Group/Category EAL SubcategoryAny Operating Mode:R -Abnormal Rad Levels / Rad Effluent I -Radiological Effluent2 -Irradiated Fuel Event3 -Area Radiation LevelsH -Hazards and Other Conditions 1 -SecurityAffecting Plant Safety 2 -Seismic Event3 -Natural or Technological Hazard4 -Fire5 -Hazardous Gases6 -Control Room Evacuation7 -Emergency Coordinator JudgmentE -ISFSI I -Confinement BoundaryHot Conditions:S -System Malfunction 1 -Loss of Emergency AC Power2 -Loss of Vital DC Power3 -Loss of Control Room Indications4 -RCS Activity5 -RCS Leakage6 -RPS Failure7 -Loss of Communications8 -Containment Failure9 -Hazardous Event Affecting Safety SystemsF -Fission Product Barrier Degradation NoneCold Conditions:C -Cold Shutdown / Refueling System 1 -RCS LevelMalfunction 2 -Loss of Emergency AC Power3 -RCS Temperature4 -Loss of Vital DC Power5 -Loss of Communications6 -Hazardous Event Affecting Safety SystemsThe primary tool for determining the emergency classification level is the EAL ClassificationMatrix. The user of the EAL Classification Matrix may (but is not required to) consult the EALTechnical Bases Document in order to obtain additional information concerning the EALsunder classification consideration. The user should consult Section 3.0 and Attachments I & 2of this document for such information.I Page 6 of 276 2.5 Technical Bases InformationEAL technical bases are provided in Attachment 1 for each EAL according to EAL group (Any,Hot, Cold), EAL category (R, C, H, S, E and F) and EAL subcategory. A summary explanationof each category and subcategory is given at the beginning of the technical bases discussionsof the EALs included in the category. For each EAL, the following information is provided:Category Letter & TitleSubcategiorv Number & TitleInitiating Condition (IC)Site-specific description of the generic IC given in NEI 99-01 Rev. 6.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 each EALidentifier:1. First character (letter): Corresponds to the EAL category as described above (R, C,H, S, E 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). If acategory does not have a subcategory, this character is assigned the number one(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)Exact wording of the EAL as it appears in the EAL Classification MatrixPage 7 of 276 Mode ApplicabilityOne or more of the following plant operating conditions comprise the mode to which eachEAL is applicable: 1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown, 5 -Cold Shutdown, 6 -Refueling, D -Defueled, or Any. (See Section 2.6 for operating modedefinitions)Definitions:If the EAL wording contains a defined term, the definition of the term is included in thissection. These definitions can also be found in Section 5.1.Basis:A Plant-Specific basis section that provides CPNPP-relevant information concerning theEAL. This is followed by a Generic basis section that provides a description of the rationalefor the EAL as provided in NEI 99-01 Rev. 6.CPNPP Basis Reference(s):Site-specific source documentation from which the EAL is derived.2.6 Operating Mode Applicability (ref. 4.1.8)1 Power OperationKeff greater than or equal to 0.99 and reactor thermal power greater than 5%2 StartupKeff greater than or equal t 0.99 and reactor thermal power-< 5%3 Hot StandbyKeff less than 0.99 and average coolant temperature greater than or equal to 3501F4 Hot ShutdownKeff less than 0.99 and average coolant temperature 3501F greater than Tavg greaterthan 200 OF and all reactor vessel head closure bolts fully tensioned5 Cold ShutdownKeff less than 0.99 and average coolant temperature < 200OF6 RefuelinqOne or more reactor vessel head closure bolts are less than fully tensionedD DefueledAll reactor fuel removed from reactor pressure vessel (full core off load during refuelingor extended outage).The plant operating mode that exists at the time that the event occurs (prior to any protectivesystem or operator action being initiated in response to the condition) should be compared tothe mode applicability of the EALs. If a lower or higher plant operating mode is reached beforethe emergency classification is made, the declaration shall be based on the mode that existedat the time the event occurred.2.7 Unit DesignationThe specific unit designator (1 or 2) is represented within these instructions by the symbol "u".The appropriate unit digit may be substituted for this symbol to obtain the unit specificequipment number (Example u-FK-121 represents 1-FK-121 for Unit 1 and 2-FK-121 for Unit2). For equipment or components that are common or non unit-specific the "X" designator isused. (Example X-RE-6272 represents a radiation monitor that is common to both units).Page 8 of 276 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS3.1 General ConsiderationsWhen making an emergency classification, the Emergency Coordinator must consider allinformation having a bearing on the proper assessment of an Initiating Condition (IC). Thisincludes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability,Notes, and the informing basis information. In the Recognition Category F matrices, EALs arebased on loss or potential loss of Fission Product Barrier Thresholds.3.1.1 Classification TimelinessNRC regulations require the licensee to establish and maintain the capability to assess,classify, and declare an emergency condition within 15 minutes after the availability ofindications to plant operators that an emergency action level has been exceeded and topromptly declare the emergency condition as soon as possible following identification of theappropriate emergency classification level. The NRC staff has provided guidance onimplementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, EmergencyPlanning for Nuclear Power Plants" (ref. 4.1.11).3.1.2 Valid IndicationsAll emergency classification assessments shall be based upon valid indications, reports orconditions. A valid indication, report, or condition, is one that has been verified throughappropriate means such that there is no doubt regarding the indicator's operability, thecondition's existence, or the report's accuracy. For example, verification could beaccomplished through an instrument channel check, response on related or redundantindicators, or direct observation by plant personnel.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) by directobservation by plant personnel, such that doubt related to the indicator's operability, thecondition's existence, or the report's accuracy is removed. Implicit in this definition is the needfor timely assessment.3.1.3 Imminent ConditionsFor ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), theEmergency Coordinator should not wait until the applicable time has elapsed, but shoulddeclare the event as soon as it is determined that the condition has exceeded, or will likelyexceed, the applicable time. If an ongoing radiological release is detected and the releasestart time is unknown, it should be assumed that the release duration specified in the IC/EALhas been exceeded, absent data to the contrary.3.1.4 Planned vs. Unplanned EventsA planned work activity that results in an expected event or condition which meets or exceedsan EAL does not warrant an emergency declaration provided that: 1) the activity proceeds asplanned, and 2) the plant remains within the limits imposed by the operating license. Suchactivities include planned work to test, manipulate, repair, maintain or modify a system orcomponent. In these cases, the controls associated with the planning, preparation andexecution of the work will ensure that compliance is maintained with all aspects of theoperating license provided that the activity proceeds and concludes as expected. Events orconditions of this type may be subject to the reporting requirements of 10 § CFR 50.72 (ref.4.1.4).Page 9 of 276 3.1.5 Classification Based on AnalysisThe assessment of some EALs is based on the results of analyses that are necessary toascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments,chemistry sampling, RCS leak rate calculation, etc.). For these EALs, the EAL wording or theassociated basis discussion will identify the necessary analysis. In these cases, the 15-minutedeclaration period starts with the availability of the analysis results that show the threshold tobe exceeded (i.e., this is the time that the EAL information is first available). The NRC expectslicensees to establish the capability to initiate and complete EAL-related analyses within areasonable period of time (e.g., maintain the necessary expertise on-shift).3.1.6 Emergency Coordinator JudgmentWhile the EALs have been developed to address a full spectrum of possible events andconditions which may warrant emergency classification, a provision for classification based onoperator/management experience and judgment is still necessary. The NEI 99-01 EALscheme provides the Emergency Coordinator with the ability to classify events and conditionsbased upon judgment using EALs that are consistent with the Emergency Classification Level(ECL) definitions (refer to Category H). The Emergency Coordinator will need to determine ifthe effects or consequences of the event or condition reasonably meet or exceed a particularECL definition. A similar provision is incorporated in the Fission Product Barrier Tables;judgment may be used to determine the status of a fission product barrier.3.2 Classification MethodologyTo make an emergency classification, the user will compare an event or condition (i.e., therelevant plant indications and reports) to an EAL(s) and determine if the EAL has been met orexceeded. The evaluation of an EAL must be consistent with the related Operating ModeApplicability and Notes. If an EAL has been met or exceeded, the associated IC is likewisemet, the emergency classification process "clock" starts, and the ECL must be declared inaccordance with plant procedures no later than fifteen minutes after the process "clock"started.When assessing an EAL that specifies a time duration for the off-normal condition, the "clock"for the EAL time duration runs concurrently with the emergency classification process "clock."For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01 (ref. 4.1.11).3.2.1 Classification of Multiple Events and ConditionsWhen multiple emergency events or conditions are present, the user will identify all met orexceeded EALs. The highest applicable ECL identified during this review is declared. Forexample:* If an Alert EAL and a Site Area Emergency EAL are met, whether at one unit or at twodifferent units, a Site Area Emergency should be declared.There is no "additive" effect from multiple EALs meeting the same ECL. For example:* If two Alert EALs are met, whether at one unit or at two different units, an Alert shouldbe declared.Related guidance concerning classification of rapidly escalating events or conditions isprovided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance forEmergency Notifications During Quickly Changing Events (ref. 4.1.2).Page 10 of 276 3.2.2 Consideration of Mode Changes During ClassificationThe mode in effect at the time that an event or condition occurred, and prior to any plant oroperator response, is the mode that determines whether or not an IC is applicable. If an eventor condition occurs, and results in a mode change before the emergency is declared, theemergency classification level is still based on the mode that existed at the time that the eventor condition was initiated (and not when it was declared). Once a different mode is reached,any new event or condition, not related to the original event or condition, requiring emergencyclassification should be evaluated against the ICs and EALs applicable to the operating modeat the time of the new event or condition.For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicablein the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is enteredduring the subsequent plant response. In particular, the fission product barrier EALs areapplicable only to events that initiate in the Hot Shutdown mode or higher.3.2.3 Classification of Imminent ConditionsAlthough EALs provide specific thresholds, the Emergency Coordinator must remain alert toevents or conditions that could lead to meeting or exceeding an EAL within a relatively shortperiod of time (i.e., a change in the ECL is IMMINENT). If, in the judgment of the EmergencyCoordinator, meeting an EAL is IMMINENT, the emergency classification should be made as ifthe EAL has been met. While applicable to all emergency classification levels, this approach isparticularly important at the higher emergency classification levels since it provides additionaltime for implementation of protective measures.3.2.4 Emergency Classification Level Upgrading and DowngradingAn ECL may be downgraded when the event or condition that meets the highest IC and EALno longer exists, and other site-specific downgrading requirements are met. If downgradingthe ECL is deemed appropriate, the new ECL would then be based on a lower applicable IC(s)and EAL(s). The ECL may also simply be terminated.As noted above, guidance concerning classification of rapidly escalating events or conditions isprovided in RIS 2007-02 (ref. 4.1.2).3.2.5 Classification of Short-Lived EventsEvent-based ICs and EALs define a variety of specific occurrences that have potential oractual safety significance. By their nature, some of these events may be short-lived and, thus,over before the emergency classification assessment can be completed. If an event occursthat meets or exceeds an EAL, the associated ECL must be declared regardless of itscontinued presence at the time of declaration. Examples of such events include anearthquake or a failure of the reactor protection system to automatically trip the reactorfollowed by a successful manual trip.3.2.6 Classification of Transient ConditionsMany of the ICs and/or EALs employ time-based criteria. These criteria will require that theIC/EAL conditions be present for a defined period of time before an emergency declaration iswarranted. In cases where no time-based criterion is specified, it is recognized that sometransient conditions may cause an EAL to be met for a brief period of time (e.g., a few secondsto a few minutes). The following guidance should be applied to the classification of theseconditions.Page 11 of 276 EAL momentarily met during expected plant response -In instances where an EAL is brieflymet during an expected (normal) plant response, an emergency declaration is not warrantedprovided that associated systems and components are operating as expected, and operatoractions are performed in accordance with procedures.EAL momentarily met but the condition is corrected prior to an emergency declaration -If anoperator takes prompt manual action to address a condition, and the action is successful incorrecting the condition prior to the emergency declaration, then the applicable EAL is notconsidered met and the associated emergency declaration is not required. For illustrativepurposes, consider the following example:An ATWS occurs and the high pressure ECCS systems fail to automatically start. RPVlevel rapidly decreases and the plant enters an inadequate core cooling condition (apotential loss of both the fuel clad and RCS barriers). If an operator manually starts ahigh pressure ECCS system in accordance with an EOP step and clears the inadequatecore cooling condition prior to an emergency declaration, then the classification shouldbe based on the ATWS only.It is important to stress that the 15-minute emergency classification assessment period(process clock) is not a "grace period" during which a classification may be delayed to allowthe performance of a corrective action that would obviate the need to classify the event.Emergency classification assessments must be deliberate and timely, with no undue delays.The provision discussed above addresses only those rapidly evolving situations when anoperator is able to take a successful corrective action prior to the Emergency Coordinatorcompleting the review and steps necessary to make the emergency declaration. Thisprovision is included to ensure that any public protective actions resulting from the emergencyclassification are truly warranted by the plant conditions.3.2.7 After-the-Fact Discovery of an Emergency Event or ConditionIn some cases, an EAL may be met but the emergency classification was not made at the timeof the event or condition. This situation can occur when personnel discover that an event orcondition existed which met an EAL, but no emergency was declared, and the event orcondition no longer exists at the time of discovery. This may be due to the event or conditionnot being recognized at the time or an error that was made in the emergency classificationprocess.In these cases, no emergency declaration is warranted; however, the guidance contained inNUREG-1022 (ref. 4.1.3) is applicable. Specifically, the event should be reported to the NRCin accordance with 10 CFR § 50.72 (ref. 4.1.4) within one hour of the discovery of theundeclared event or condition. The licensee should also notify appropriate State and localagencies in accordance with the agreed upon arrangements.3.2.8 Retraction of an Emergency DeclarationGuidance on the retraction of an emergency declaration reported to the NRC is discussed inNUREG-1022 (ref. 4.1.3).Page 12 of 276 1 | |||
==4.0 REFERENCES== | |||
4.1 Developmental4.1.1 NEI 99-01 Revision 6, Methodology for the Development of Emergency ActionLevels for Non-Passive Reactors, ADAMS Accession Number ML12326A8054.1.2 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications DuringQuickly Changing Events, February 2, 2007.4.1.3 NUREG-1022 Event Reporting Guidelines: 10CFR50.72 and 50.734.1.4 10 § CFR 50.72 Immediate Notification Requirements for Operating NuclearPower Reactors4.1.5 10 § CFR 50.73 License Event Report System4.1.6 CPNPP Emergency Plan Appendix E, Complex and Owner Controlled Area4.1.7 CPNPP FSAR Section 2.1.1 Site Location and Description4.1.8 Technical Specifications Table 1.1-1 Modes4.1.9 OPT-408A/B Refueling Containment Penetration Verification4.1.10 ODA-207 Guidelines on the Preparation and Review of Operations Procedures4.1.11 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for NuclearPower Plants4.1.12 IPO-01OA/B Reactor Coolant System Reduced Inventory Operations4.1.13 Technical Specifications 3.9.44.1.14 CPNPP Offsite Dose Calculation Manual (ODCM)4.2 Implementing4.2.1 EPP-201, Assessment of Emergency Action Levels, Emergency Classificationand Plan Activation4.2.2 NEI 99-01 Rev. 6 to CPNPP EAL Comparison Matrix4.2.3 CPNPP EAL MatrixPage 13 of 276 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS5.1 Definitions (ref. 4.1.1 except as noted)Selected terms used in Initiating Condition and Emergency Action Level statements are set inall capital letters (e.g., ALL CAPS). These words are defined terms that have specificmeanings as used in this document. The definitions of these terms are provided below.AlertEvents are in process, 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 hostile action. Anyreleases are expected to be small fractions of the EPA Protective Action Guideline exposurelevels.Containment ClosureThe procedurally defined actions taken to secure containment and its associated structures,systems, and components as a functional barrier to fission product release under shutdownconditions. Containment closure means that all potential escape paths are closed or capableof being closed (ref.4.1.13).A. All penetrations providing direct access from Containment atmosphere to outsideatmosphere are closed except:" Penetrations with automatic valves capable of being closed by an operable CVI" Penetrations under administrative controls (e.g., Control Room notified and designatedperson to close if required by fuel handling accident)B. Equipment hatch is closed and held in place by 4 bolts, or is capable of being closed andheld in place by 4 boltsC. One emergency airlock door is closedD. One personnel airlock door is capable of being closed (ref. 4.1.9)EPA PAGsEnvironment Protection Agency Protective Action Guidelines. The EPA PAGs are expressed interms of dose commitment: 1 Rem TEDE or 5 Rem CDE Thyroid. Actual or projected offsiteexposures in excess of the EPA PAGs requires CPNPP to recommend protective actions forthe general public to offsite planning agencies.Exclusion Area BoundaryExclusion Area Boundary is a synonymous term for Site Boundary. CPNPP FSAR Section2.1.1.3 and Figure 2.1-2 define the Exclusion Area Boundary. This boundary is used forestablishing effluent release limits with respect to the requirements of IOCFR20 (ref. 4.1.7).See also CPNPP Emergency Plan Appendix E, Complex and Owner Controlled Area (ref.4.1.6) and CCNPP ODCM Section 5.0 Design Features (ref. 4.1.14).ExplosionA rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemicalreaction or overpressurization. A release of steam (from high energy lines or components) oran electrical component failure (caused by short circuits, grounding, arcing, etc.) should notPage 14 of 276 automatically be considered an explosion. Such events require a post-event inspection todetermine if the attributes of an explosion are present.FaultedThe term applied to a steam generator that has a steam leak on the secondary side ofsufficient size to cause an uncontrolled drop in steam generator pressure or the steamgenerator to become completely depressurized.FireCombustion characterized by heat and light. Sources of smoke such as slipping drive belts oroverheated electrical equipment do not constitute fires. Observation of flame is preferred butis NOT required if large quantities of smoke and heat are observed.FloodinaA condition where water is enteringq a room or area faster than installed equipment is capableof removal, resulting in a rise of water level within the room or area.General EmergencyEvents are in process or have occurred which involve actual or imminent substantial coredegradation or melting with potential for loss of containment integrity or hostile actions thatresult in an actual loss of physical control of the facility. Releases can be reasonably expectedto exceed EPA Protective Action Guideline exposure levels offsite for more than the immediatesite area.HostageA person(s) held as leverage against the station to ensure that demands will be met by thestation.Hostile ActionAn act toward CPNPP or its personnel that includes the use of violent force to destroyequipment, take hostages, and/or intimidate the licensee to achieve an end. This includesattack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices usedto deliver destructive force. Other acts that satisfy the overall intent may be included. Hostileaction should not be construed to include acts of civil disobedience or felonious acts that arenot part of a concerted attack on CPNPP. Non-terrorism-based EALs should be used toaddress such activities (i.e., this may include violent acts between individuals in the ownercontrolled area).Hostile ForceOne or more individuals who are engaged in a determined assault, overtly or by stealth anddeception, equipped with suitable weapons capable of killing, maiming, or causing destruction.ImminentThe trajectory of events or conditions is such that an EAL will be met within a relatively shortperiod of time regardless of mitigation or corrective actions.ImPede(d)Personnel access to a room or area is hindered to an extent that extraordinary measures arenecessary to facilitate entry of personnel into the affected room/area (e.g., requiring use ofprotective equipment, such as SCBAs, that is not routinely employed).Page 15 of 276 Independent Spent Fuel Storage Installation (ISFSI)A complex that is designed and constructed for the interim storage of spent nuclear fuel andother radioactive materials associated with spent fuel storage.MaintainTake appropriate action to hold the value of an identified parameter within specified limits.Normal LevelsSappliod to radiological lIdEALs, the highest reading in the past t'A'cty-foue orexcluding th.e current peak value.Owner Controlled AreaAs shown in CPNPP Emergency Plan Appendix E, Complex and Owner Controlled Area.ProjectileAn object directed toward a Nuclear Power Plant that could cause concern for its continuedoperability, reliability, or personnel safety.Protected AreaAn area encompassed by physical barriers and to which access is controlled. The ProtectedArea refers to the designated security area around the process buildings and is depicted inFSAR Figure 1.2-1 Plot Plan (ref. 4.1.7).RCS 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).Reduced InventoryPlant condition when fuel is in the reactor vessel and Reactor Coolant System level is < 80inches above core plate (829'8") (ref. 4.1.12).Refueling PathwayThe reactor refueling cavity, spent fuel pool and fuel transfer canal comprise the refuelingpathway.RupturedThe condition of a steam generator in which primary-to-secondary leakage is of sufficientmagnitude to require a safety injection.RestoreTake the appropriate action required to return the value of an identified parameter to theapplicable limits.Safety SystemA system required for safe plant operation, cooling down the plant and/or placing it in the coldshutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in IOCFR50.2):Those structures, systems and components that are relied upon to remain functional duringand following design basis events to assure:Page 16 of 276 (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 could resultin 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 degradation tothe level of safety of the plant. A security condition does not involve a hostile action.Site Area EmergencyEvents are in process or have occurred which involve actual or likely maior failures of plantfunctions needed for protection of the public or hostile actions that result in intentional damageor malicious acts; (1) toward site personnel or equipment that could lead to the likely failure ofor; (2) that prevent effective access to equipment needed for the protection of the public. Anyreleases are not expected to result in exposure levels which exceed EPA Protective ActionGuidelines exposure levels beyond the site boundary.Site BoundarySee EXCLUSION AREA BOUNDARYUnisolableAn open or breached system line that cannot be isolated, remotely or locally.UnplannedA parameter change or an event that is not 1) the result of an intended evolution or 2) anexpected plant response to a transient. The cause of the parameter change or event may beknown or unknown.Unusual EventEvents are in process or have occurred which indicate a potential degradation in the level ofsafety of the plant or indicate a security threat to facility protection has been initiated. Noreleases of radioactive material requiring offsite response or monitoring are expected unlessfurther degradation of safety systems occurs.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) by directobservation by plant personnel, such that doubt related to the indicator's operability, thecondition's existence, or the report's accuracy is removed. Implicit in this definition is the needfor timely assessment.Visible DamageDamage to a component or structure that is readily observable without measurements, testing,or analysis. The visual impact of the damage is sufficient to cause concern regarding theoperability or reliability of the affected component or structure.Page 17 of 276 5.2 Abbreviations/AcronymsOF ....................................................................................................... Degrees Fahrenheit0 .......................................................................................................................... D e g re e sAC ....................................................................................................... Alternating CurrentAPDG .......................................................................... Alternate Power Diesel GeneratorATW S ...................................................................... Anticipated Transient W ithout ScramCPNPP .................................................................. Com anche Peak Nuclear Power PlantCDE ....................................................................................... Com m itted Dose EquivalentCFR ..................................................................................... Code of Federal RegulationsCNTM T .......................................................................................................... Containm entCSFST ....................................................................... Critical Safety Function Status TreeDBA ............................................................................................... Design Basis AccidentDC ............................................................................................................... Direct CurrentEAL ............................................................................................. Em ergency Action LevelECCS ............................................................................ Em ergency Core Cooling SystemECL ................................................................................. Em ergency Classification LevelEO F .................................................................................. Em ergency O perations FacilityEO P ............................................................................... Em ergency Operating ProcedureEPA .............................................................................. Environm ental Protection AgencyERG ................................................................................ Em ergency Response G uidelineEPIP ................................................................ Em ergency Plan Im plem enting ProcedureESF ........................................................................................ Engineered Safety FeatureESW ........................................................................................ Em ergency Service W aterFAA ................................................................................. Federal Aviation Adm inistrationFBI ................................................................................... Federal Bureau of InvestigationFEMA ............................................................... Federal Em ergency M anagem ent AgencyFSAR .................................................................................... Final Safety Analysis ReportG E ..................................................................................................... General Em ergencyIC ......................................................................................................... Initiating ConditionIPEEE ................. Individual Plant Examination of External Events (Generic Letter 88-20)Keff ......................................................................... Effective Neutron M ultiplication FactorLCO .................................................................................. Lim iting Condition of OperationLER ............................................................................................... Licensee Event ReportLOCA ......................................................................................... Loss of Coolant AccidentLW R ................................................................................................... Light W ater ReactorM PC ................................... M axim um Perm issible Concentration/M ulti-Purpose Canisterm R, m Rem , m rem , m REM .............................................. m illi-Roentgen Equivalent M anM SL ........................................................................................................ M ain Steam LineM W .................................................................................................................... M egawattNEI ........................................................ ..................................... Nuclear Energy InstitutePage 18 of 276 NESP ................................................................... National Environm ental Studies ProjectNPP .................................................................................................. Nuclear Power PlantNRC ................................................................................ Nuclear Regulatory Com m issionNSSS ................................................................................ Nuclear Steam Supply SystemNO RAD ................................................... North Am erican Aerospace Defense Com m and(NO )UE ................................................................................ Notification of Unusual EventO BE ...................................................................................... O perating Basis EarthquakeO CA ............................................................................................... Owner Controlled AreaO DCM ............................................................................ Off-site Dose Calculation M anualO RO ................................................................................. Offsite Response OrganizationOTO ............................................................................... Off-Norm al O perating ProcedurePA .............................................................................................................. Protected AreaPAG ........................................................................................ Protective Action GuidelinePRA/PSA ..................... Probabilistic Risk Assessment / Probabilistic Safety AssessmentPW R ....................................................................................... Pressurized W ater ReactorPSIG ................................................................................ Pounds per Square Inch GaugeR ........................................................................................................................ RoentgenRCC ............................................................................................ Reactor Control ConsoleRCS ............................................................................................ Reactor Coolant SystemRem , rem , REM ....................................................................... Roentgen Equivalent M anRETS ......................................................... Radiological Effluent Technical SpecificationsRPS ....................................................................................... Reactor Protection SystemR(P)V ....................................................................................... Reactor (Pressure) VesselRVLIS ................................................................. Reactor Vessel Level Indicating SystemSAR ............................................................................................... Safety Analysis ReportSBO ......................................................................................................... Station BlackoutSCBA ....................................................................... Self-Contained Breathing ApparatusSG ......................................................................................................... Steam GeneratorSI .............................................................................................................. Safety InjectionO DCM ............................................................................. Offsite Dose Calculation M anualSPDS ........................................................................... Safety Param eter Display SystemSRO ............................................................................................ Senior Reactor O peratorTEDE ............................................................................... Total Effective Dose EquivalentTOAF .................................................................................................... Top of Active FuelTSC ... ...................................................................................... Technical Support CenterW O G ................................................................................... W estinghouse Ow ners GroupI Page 19 of 276 6.0 CPNPP-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCEThis cross-reference is provided to facilitate association and location of a CPNPP EAL withinthe NEI 99-01 IC/EAL identification scheme. Further information regarding the development ofthe CPNPP EALs based on the NEI guidance can be found in the EAL Comparison Matrix.CPNPP NEI 99-01 Rev. 6EAL IC ExampleEALRU1.1 AU1 1,2RU1.2 AU1 3RU2.1 AU2 IRA1.1 AA1 1RA1.2 AA1 2RAU.3 AA1 3RA1.4 AA1 4RA2.1 AA2 1RA2.2 AA2 2RA2.3 AA2 3RA3.1 AA3 1RA3.2 AA3 2RS1.1 AS1 1RS1.2 AS1 2RS1.3 AS1 3RS2.1 AS2 1RGI.1 AG1 1RG1.2 AG1 2RG1.3 AG1 3RG2.1 AG2 1CU1.1 CU1 1Page 20 of 276 CPNPP NEI 99-01 Rev. 6EAL IC ExampleEALCU1.2 CUI 2CU2.1 CU2 1CU3.1 CU3 1CU3.2 CU3 2CU4.1 CU4 1CU5.1 CU5 1,2,3CAI.1 CA1 1CAI.2 CA1 2CA2.1 CA2 1CA3.1 CA3 1,2CA6.1 CA6 1CS1.1 CS1 1CS1.2 CS1 2CS1.3 CS1 3CGI.1 CG1 2FAI.1 FA1 IFSI.1 FS1 1FG1.1 FG1 1HU1.1 HU1 1HU1.2 HUI 2HU1.3 HUI 3HU2.1 HU2 IHU3.1 HU3 1HU3.2 HU3 2HU3.3 HU3 3Page 21 of 276 CPNPP NEI 99-01 Rev. 6EAL IC ExampleEALHU3.4 HU3 4HU4.1 HU4 1HU4.2 HU4 2HU4.3 HU4 3HU4.4 HU4 4HU7.1 HU7 1HA1.1 HAl 1HA1.2 HAl 2HA5.1 HA5 IHA6.1 HA6 1HA7.1 HA7 1HSl.1 HS1 IHS6.1 HS6 1HS7.1 HS7 1HG1.1 HG1 1HG7.1 HG7 1su1.1 Sul 1SU3.1 SU2 1SU4.1 SU3 1SU4.2 SU3 2SU5.1 SU4 1,2,3SU6.1 SU5 ISU6.2 SU5 2SU7.1 SU6 1,2,3SU8.1 SU7 1Page 22 of 276 CPNPP NEI 99-01 Rev. 6EAL IC ExampleEALSA1.1 SA1 1SA3.1 SA2 1SA6.1 SA5 1SA9.1 SA9 1SS1.1 SS1 1SS2.1 SS8 1SS6.1 SS5 1SGl.1 SG1 1SG1.2 SG8 1EUI.1 E-HU1 1Page 23 of 276 7.0 ATTACHMENTS7.1 Attachment 1, Emergency Action Level Technical Bases7.2 Attachment 2, Fission Product Barrier Matrix and BasisPage 24 of 276 ATTACHMENT 1EAL BasesCategory 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 barriers becauseof the elevated potential for offsite radioactivity release. Degradation of fission productbarriers though is not always apparent via non-radiological symptoms. Therefore, directindication of elevated radiological effluents or area radiation levels are appropriate symptomsfor emergency classification.At lower levels, abnormal radioactivity releases may be indicative of a failure of containmentsystems or precursors to more significant releases. At higher release rates, offsite radiologicalconditions may result which require offsite protective actions. Elevated area radiation levels inplant may also be indicative of the failure of containment systems or preclude access to plantvital equipment necessary to ensure plant safety.Events of this category pertain to the following subcategories:1. Radiological EffluentDirect indication of effluent radiation monitoring systems provides a rapid assessmentmechanism to determine releases in excess of classifiable limits. Projected offsite doses,actual offsite field measurements or measured release rates via sampling indicate dosesor dose rates above classifiable limits.2. Irradiated Fuel EventConditions indicative of a loss of adequate shielding or damage to irradiated fuel maypreclude access to vital plant areas or result in radiological releases that warrantemergency classification.3. Area Radiation LevelsSustained general area radiation levels which may preclude access to areas requiringcontinuous occupancy also warrant emergency classification.I Page 25 of 276 1 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:R -Abnormal Rad Levels / Rad Effluent1 -Radiological EffluentRelease of gaseous or liquid radioactivity greater than 2 times theODCM limits for 60 minutes or longerEAL:RUI.1 Unusual EventReading on any Table R-1 effluent radiation monitor greater than column "UE" for greaterthan or equal to 60 min.(Notes 1, 2, 3)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the releaseduration has exceeded the specified time limit.Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path isisolated, the effluent monitor reading is no longer VALID for classification purposes.Table R-1 Effluent Monitor Classification ThresholdsRelease Point Monitor GE SAE Alert UEPlant Vent X-RE-5567............6.52E-4 1Cic/,mlPVG384 + PVG385 A + BPlant Vent(WRGM)X-RE-5570(WRGM) A + B 4.0E+7 ý.Ci/sec 4.OE+6 p.Ci/sec 4.0E+5 pCi/sec 4.OE+4 iiCi/sec0PVF684 + PVF685S Main SteamMSLu78 u-RE-2325MSLu79 u-RE-2326 90 pici/ml* 9.0 p.Cil/ml* 0.9 t4ci/ml* 2 x high alarmMSLu80 u-RE-2327 setpoint*MSLu81 u-RE-2328Liquid Waste X-RE-5253 2 x high alarmLWE-076 setpointService Water"J SSei65 u-RE-4269 2 x high alarmu-RE-4270 setpointSSWu66* with reactor shutdownMode Applicability:AllI Page 26 of 276 ATTACHMENT 1EAL BasesDefinition(s):NoneCPNPP Basis:The column "UE" gaseous and liquid release values in Table R-1 represent two times thealarm setpoint of the specified monitors. The setpoints are established to ensure the ODCMrelease limits are not exceeded. (ref. 1)Plant Vent Monitors sample both plant vent stacks prior to discharge to the environment. Theydetect normal operational levels of noble gases. The noble gas detectors (X-RE-5567A, B) canbe used as backups to the wide range gas monitors (X-RE-5570A, B). These monitorscommunicate with the RM-23s in the Control Room. Indication and annunciation are providedin the Control Room for alert and high radiation levels and monitor failure. (ref. 2)The WRGM system is a gaseous effluent monitoring system composed of two identicalmonitors used for detection of noble gas releases through the two plant vent stacks. Exhaustfrom the main turbine gland steam condenser exhauster is routed to the vent stacks formonitoring prior to release. Particulate and iodine grab samples may also be obtained from theWRGM. These monitors also initiate the automatic closure of the gas release valve in thewaste gas processing system on detection of high radiation. Indication and annunciation areprovided in the Control Room for alert and high radiation levels and monitor failure. (ref. 2)There are four online Main Steam Line Monitors (MSL) for each steam generator. Each oneconsists of a shielded, Category II seismic detector mounted adjacent to a main steam line, aremote RM-80 microprocessor and a remote customer interface junction box. The RM-80associated with the MSL monitor communicates with the PC-1I1 CRT console computer. (ref.2).Plant Liquid Waste Processing System (LWPS) discharge is continuously monitored by ashielded gamma sensitive (Nal(T1)) scintillation detector. When a LWPS discharge is required,normally locked-closed control valves can be opened directing flow through a path containing aradiation monitor (X-RE-5253) and a control valve which discharges waste to the circulatingwater discharge tunnel. The control valves are administratively controlled with a key-operatedswitch selectable to closed, automatic, or "key-held" open modes. In the automatic position,the valve will close on monitor high radiation alarm or monitor failure signals. Indication andannunciation are provided on the Waste Processing System (WPS) control panel for alert ormonitor failure alarm and in the Control Room for alert, high, and monitor failure alarms. (ref. 2)Service Water monitors are provided to monitor the Service Water System for radiation sinceleakage from radioactive fluid systems could cause potential radioactive leakage to theenvironment. A shielded gamma sensitive scintillation (Nal(T1)) detector is located in an off-line sample assembly downstream of each component cooling water heat exchanger tomonitor service water being discharged. Indication and annunciation are provided at theControl Room RMS console. (ref. 2)NEI 99-01 Basis:This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time(e.g., an uncontrolled release). It includes any gaseous or liquid radiological release,Page 27 of 276 ATTACHMENT IEAL Basesmonitored or un-monitored, including those for which a radioactivity discharge permit isnormally prepared.Nuclear power plants incorporate design features intended to control the release of radioactiveeffluents to the environment. Further, there are administrative controls established to preventunintentional releases, and to control and monitor intentional releases. The occurrence of anextended, uncontrolled radioactive release to the environment is indicative of degradation inthese features and/or controls.Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plant conditionsalone. The inclusion of both plant condition and radiological effluent EALs more fully addressesthe spectrum of possible accident events and conditions.Classification based on effluent monitor readings assumes that a release path to theenvironment is established. If the effluent flow past an effluent monitor is known to havestopped due to actions to isolate the release path, then the effluent monitor reading is nolonger valid for classification purposes.Releases should not be prorated or averaged. For example, a release exceeding 4 timesrelease limits for 30 minutes does not meet the EAL.I EAL--#-I-This EAL addresses normally occurring continuous radioactivity releases frommonitored gaseous or liquid effluent pathways.EAL #t2 This EAL addergenscs radioactioity leases that cause c-lucRt radiation monitoreadings to oxceed 2 times, the limit established by a radioactivity discharge permit. This EALWill typically be associated with planned batch rcleases fErom non continuous release pathways(e.g., r-adwaste, waste gas).EAL= 13 This E=AL addresses uncontrollcd gaseous or liquid relcascs that are detected bysample analyses or enviMromental sur~eys, particularly on unmonitored pathways (e.g., spoillsof radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.)-.Escalation of the emergency classification level would be via IC AAI-RPAI.CPNPP Basis Reference(s):1. CPNPP ODCM Unit 1 and 22. DBD-EE-023 Radiation Monitoring System3. EAL Section R Revision 6 Table R-1 Effluent Monitor Classification Thresholds Review4. NEI 99-01 AU1Page 28 of 276 1 ATTACHMENT 1EAL BasesCategory: R -Abnormal Rad Levels / Rad EffluentSubcategory: 1 -Radiological EffluentInitiating Condition: Release of gaseous or liquid radioactivity greater than 2 times theODCM limits for 60 minutes or longer.EAL:RUI.2 Unusual EventSample analysis for a gaseous or liquid release indicates a concentration or release rate> 2 x ODCM limits for greater than or equal to 60 min. (Notes 1, 2)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the releaseduration has exceeded the specified time limit.Mode Applicability:AllDefinition(s):NoneCPNPP Basis:NoneNEI 99-01 Basis:This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time(e.g., an uncontrolled release). It includes any gaseous or liquid radiological release,monitored or un-monitored, including those for which a radioactivity discharge permit isnormally prepared.Nuclear power plants incorporate design features intended to control the release of radioactiveeffluents to the environment. Further, there are administrative controls established to preventunintentional releases, and to control and monitor intentional releases. The occurrence of anextended, uncontrolled radioactive release to the environment is indicative of degradation inthese features and/or controls.Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plant conditionsalone. The inclusion of both plant condition and radiological effluent EALs more fully addressesthe spectrum of possible accident events and conditions.Cla~ssfication basecd on cfflucnt mon~itor readings assumes that a release path to henvionmet is established. if the effluent flo)w past an effluent monitor is known to havostopped due to actions to isolatc the release path, thcn the effluent moniRtorF reading is nolonger valid fra f purposes.Page 29 of 276 ATTACHMENT 1EAL BasesReleases should not be prorated or averaged. For example, a release exceeding 4 timesrelease limits for 30 minutes does not meet the EAL.EAL #1 This E.A.L , Addresses normally occurring continuous radioactiVity .eleas' .frommonioredgaseous or liquid effluent pathways.EAL-- -2 This EAL addresses radioactiVity releases that c-ause -ffluent rditomntrreadings to exceed 2 times the li~mit esta-bli-shed by a radioactiv~ity discharge permit. This ,A~LWil typically be asSocGiate ithpland- bhatch releases from non continuous release pathways(e.g., radwaste, waste gas).EAL-#3 --This EAL addresses uncontrolled gaseous or liquid releases that are detected bysample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spillsof radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.).Escalation of the emergency classification level would be via IC AA4-RA1.CPNPP Basis Reference(s):1. CPNPP ODCM Unit 1 and 22. NEI 99-01 AU1Page 30 of 276 ATTACHMENT 1EAL BasesCategory: R -Abnormal Rad Levels / Rad EffluentSubcategory: 1 -Radiological EffluentInitiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dosegreater than 10 mrem TEDE or 50 mrem thyroid CDEEAL:RA1.1 AlertReading on any Table R-1 effluent radiation monitor greater than column "ALERT" forgreater than or equal to 15 min. (Notes 1, 2, 3, 4)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the releaseduration has exceeded the specified time limit.Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path isisolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4 The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be usedfor emergency classification assessments until the results from a dose assessment using actualmeteorology are available.Table R-1 Effluent Monitor Classification ThresholdsRelease Point Monitor GE SAE Alert UEPlant Vent X-RE-5567.............6.52E-4 pCi/mlPVG384 + PVG385 A + BPlant Vent(WRGM)X-RE-5570(WRGM) X- 7 A + B 4.OE+7 ýCi/sec 4.OE+6 4.OE+5 pci/sec 4.OE+4 pCi/sec0 PVF684 + PVF685' Main SteamMSLu78 u-RE-2325MSLu79 u-RE-2326 90 pci/ml* 9.0 pci/ml* 0.9 j, ci/ml* 2xhigh alarmMSLu8O u-RE-2327MSLu81 u-RE-2328Liquid Waste X-RE-5253 2 x high alarmLWE-076 setpointM Servic Water u-RE-4269 2 x high alarmSSWu65u-RE-4270 setpointSSWu66* with reactor shutdownMode Applicability:AllI Page 31 of 276 ATTACHMENT 1EAL BasesDefinition(s):NoneCPNPP Basis:This EAL address gaseous radioactivity releases, that for whatever reason, cause effluentradiation monitor readings corresponding to site boundary doses that exceed either:0 10 mRem TEDE* 50 mRem CDE ThyroidThe column "ALERT" gaseous effluent release values in Table R-1 correspond to calculateddoses of 1 % (10% of the SAE thresholds) of the EPA Protective Action Guidelines (TEDE orCDE Thyroid) (ref. 1).NEI 99-01 Basis:This IC addresses a release of gaseous or liquid radioactivity that results in projected or actualoffsite doses greater than or equal to 1 % of the EPA Protective Action Guides (PAGs). Itincludes both monitored and un-monitored releases. Releases of this magnitude represent anactual or potential substantial degradation of the level of safety of the plant as indicated by aradiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolledrelease).Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plant conditionsalone. The inclusion of both plant condition and radiological effluent EALs more fully addressesthe spectrum of possible accident events and conditions.The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDEwas established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to theenvironment is established. If the effluent flow past an effluent monitor is known to havestopped due to actions to isolate the release path, then the effluent monitor reading is nolonger valid for classification purposes.Escalation of the emergency classification level would be via IC ASI-RSI.CPNPP Basis Reference(s):1. EAL Section R Revision 6 Table R-1 Effluent Monitor Classification Thresholds Review2. NEI 99-01 AA1Page 32 of 276 ATTACHMENT IEAL BasesCategory: R -Abnormal Rad Levels / Rad EffluentSubcategory: 1 -Radiological EffluentInitiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dosegreater than 10 mrem TEDE or 50 mrem thyroid CDEEAL:RA1.2 AlertDose assessment using actual meteorology indicates doses greater than 10 mrem TEDEor 50 mrem thyroid CDE at or beyond the EXCLUSION AREA BOUNDARY (Notes 3, 4)Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path isisolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: The pre-calculated effluent monitor values presented in EALs RAI.1, RS1.1 and RGI.1 should be usedfor emergency classification assessments until the results from a dose assessment using actualmeteorology are available.Mode Applicability:AllDefinition(s):EXCLUSION AREA BOUNDARY- Exclusion Area Boundary is a synonymous term for SiteBoundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion AreaBoundary. This boundary is used for establishing effluent release limits with respect to therequirements of 1OCFR20. See also CPNPP Emergency Plan Appendix E, Complex andOwner Controlled Area and CCNPP ODCM Section 5.0 Design Features.CPNPP Basis:Dose assessments are performed by computer-based method (ref. 1)NEI 99-01 Basis:This IC addresses a release of gaseous or liquid radioactivity that results in projected or actualoffsite doses greater than or equal to 1 % of the EPA Protective Action Guides (PAGs). Itincludes both monitored and un-monitored releases. Releases of this magnitude represent anactual or potential substantial degradation of the level of safety of the plant as indicated by aradiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolledrelease).Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plant conditionsalone. The inclusion of both plant condition and radiological effluent EALs more fully addressesthe spectrum of possible accident events and conditions.The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDEwas established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to theenvironment is established. If the effluent flow past an effluent monitor is known to havePage 33 of 276 ATTACHMENT 1EAL Basesstopped due to actions to isolate the release path, then the effluent monitor reading is nolonger valid for classification purposes.Escalation of the emergency classification level would be via IC AS-14RSI.CPNPP Basis Reference(s):1. EPP-303 Operation of Computer Based Dose Assessment System2. NEI 99-01 AA1Page 34 of 276 ATTACHMENT 1EAL BasesCategory: R -Abnormal Rad Levels / Rad EffluentSubcategory: I -Radiological EffluentInitiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dosegreater than 10 mrem TEDE or 50 mrem thyroid CDEEAL:RA1.3 AlertAnalysis of a liquid effluent sample indicates a concentration or release rate that wouldresult in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond theEXCLUSION AREA BOUNDARY for 60 min. of exposure (Notes 1, 2)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the releaseduration has exceeded the specified time limit.Mode Applicability:AllDefinition(s):EXCLUSION AREA BOUNDARY -Exclusion Area Boundary is a synonymous term for SiteBoundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion AreaBoundary. This boundary is used for establishing effluent release limits with respect to therequirements of 10CFR20. See also CPNPP Emergency Plan Appendix E, Complex andOwner Controlled Area and CCNPP ODCM Section 5.0 Design Features.CPNPP Basis:Dose assessments based on liquid releases are performed per Offsite Dose CalculationManual (ref. 1).NEI 99-01 Basis:This IC addresses a release of gaseous or liquid radioactivity that results in projected or actualoffsite doses greater than or equal to 1 % of the EPA Protective Action Guides (PAGs). Itincludes both monitored and un-monitored releases. Releases of this magnitude represent anactual or potential substantial degradation of the level of safety of the plant as indicated by aradiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolledrelease).Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plant conditionsalone. The inclusion of both plant condition and radiological effluent EALs more fully addressesthe spectrum of possible accident events and conditions.The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDEwas established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Page 35 of 276 ATTACHMENT 1EAL BasesClassification based on effluent monitor readings assumes that a release path to theenvironment is established. If the effluent flow past an effluent monitor is known to havestopped due to actions to isolate the release path, then the effluent monitor reading is nolonger valid for classification purposes.Escalation of the emergency classification level would be via IC AS4RS1.CPNPP Basis Reference(s):1. CPNPP Offsite Dose Calculation Manual2. NEI 99-01 AA1Page 36 of 276 ATTACHMENT 1EAL BasesCategory: R -Abnormal Rad Levels I Rad EffluentSubcategory: 1 -Radiological EffluentInitiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dosegreater than 10 mrem TEDE or 50 mrem thyroid CDEEAL:RA1.4 AlertField survey results indicate EITHER of the following at or beyond the EXCLUSION AREABOUNDARY:" Closed window dose rates greater than 10 mR/hr expected to continue for greaterthan or equal to 60 min." Analyses of field survey samples indicate thyroid CDE greater than 50 mrem for 60min. of inhalation.(Notes 1,2)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the releaseduration has exceeded the specified time limit.Mode Applicability:AllDefinition(s):EXCLUSION AREA BOUNDARY- Exclusion Area Boundary is a synonymous term for SiteBoundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion AreaBoundary. This boundary is used for establishing effluent release limits with respect to therequirements of 10CFR20. See also CPNPP Emergency Plan Appendix E, Complex andOwner Controlled Area and CCNPP ODCM Section 5.0 Design Features.CPNPP Basis:EPP-309 Onsite/In-Plant Radiological Surveys and Offsite Radiological Monitoring providesguidance for emergency or post-accident radiological environmental monitoring (ref. 1).NEI 99-01 Basis:This IC addresses a release of gaseous or liquid radioactivity that results in projected or actualoffsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). Itincludes both monitored and un-monitored releases. Releases of this magnitude represent anactual or potential substantial degradation of the level of safety of the plant as indicated by aradiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolledrelease).Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plant conditionsalone. The inclusion of both plant condition and radiological effluent EALs more fully addressesPage 37 of 276 ATTACHMENT 1EAL Basesthe spectrum of possible accident events and conditions.The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDEwas established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to theenvironment is established. If the effluent flow past an effluent monitor is known to havestopped due to actions to isolate the release path, then the effluent monitor reading is nolonger valid for classification purposes.Escalation of the emergency classification level would be via IC AS4RSI.CPNPP Basis Reference(s):1. EPP-309 Onsite/In-Plant Radiological Surveys and Offsite Radiological Monitoring2. NEI 99-01 AA1Page 38 of 276 1 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:R -Abnormal Rad Levels / Rad Effluent1 -Radiological EffluentRelease of gaseous radioactivity resulting in offsite dose greater than100 mrem TEDE or 500 mrem thyroid CDEEAL:RSI.1 Site Area EmergencyReading on any Table R-1 effluent radiation monitor greater than column "SAE" forgreater than or equal to 15 min.(Notes 1, 2, 3, 4)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the releaseduration has exceeded the specified time limit.Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path isisolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: The pre-calculated effluent monitor values presented in EALs RAI.1, RSI.1 and RGI.1 should be usedfor emergency classification assessments until the results from a dose assessment using actualmeteorology are available.Table R-1 Effluent Monitor Classification ThresholdsRelease Point Monitor GE SAE Alert UEPlant Vent X-RE-5567PVG384 + PVG385 A + BPlant Vent(WRGM)X-RE-5570(WRGM) +- 7 A + B 4.OE+7 [.Ci/sec 4.OE+6 ýtCi/sec 4.0E+5 pCi/sec 4.0E+4 p.Ci/sec0 PVF684 + PVF685C Main SteamMSLu78 u-RE-2325MSLu79 u-RE-2326 90 [Ci/ml* 9.0 [.ci/ml* 0.9 j.ci/ml* 2 x high alarmMSLu80 u-RE-2327 setpoint*MSLu81 u-RE-2328Liquid Waste X-RE-5253 ...............-2 x high alarmLWE-076 setpointr" Service Water* Srcu-RE-4269 2 x high alarmS SSWu65 ----.--.-----.2x.hgh.larSSw65 u-RE-4270 setpointSSWu66* with reactor shutdownMode Applicability:AllDefinition(s):Page 39 of 276 ATTACHMENT 1EAL BasesNoneCPNPP Basis:This EAL address gaseous radioactivity releases, that for whatever reason, cause effluentradiation monitor readings corresponding to site boundary doses that exceed either:* 100 mRem TEDE0 500 mRem CDE ThyroidThe column "SAE" gaseous effluent release value in Table R-1 corresponds to calculateddoses of 10% of the EPA Protective Action Guidelines (TEDE or CDE Thyroid) (ref. 1).NEI 99-01 Basis:This IC addresses a release of gaseous radioactivity that results in projected or actual offsitedoses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includesboth monitored and un-monitored releases. Releases of this magnitude are associated withthe failure of plant systems needed for the protection of the public.Radiological effluent EALs are also-included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plant conditionsalone. The inclusion of both plant condition and radiological effluent EALs more fully addressesthe spectrum of possible accident events and conditions.The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDEwas established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to theenvironment is established. If the effluent flow past an effluent monitor is known to havestopped due to actions to isolate the release path, then the effluent monitor reading is nolonger valid for classification purposes.Escalation of the emergency classification level would be via IC AG4RGI.CPNPP Basis Reference(s):1. EAL Section R Revision 6 Table R-1 Effluent Monitor Classification Thresholds Review2. NEI 99-01 AS1I Page 40 of 276 ATTACHMENT IEAL BasesCategory: R -Abnormal Rad Levels I Rad EffluentSubcategory: 1 -Radiological EffluentInitiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than100 mrem TEDE or 500 mrem thyroid CDEEAL:RS1,2 Site Area EmergencyDose assessment using actual meteorology indicates doses greater than 100 mrem TEDEor500 mrem thyroid CDE at or beyond the EXCLUSION AREA BOUNDARY (Notes 3, 4)Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path isisolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: The pre-calculated effluent monitor values presented in EALs RAI.1, RSI.1 and RG1.1 should be usedfor emergency classification assessments until the results from a dose assessment using actualmeteorology are available.Mode Applicability:AllDefinition(s):EXCLUSION AREA BOUNDARY -Exclusion Area Boundary is a synonymous term for SiteBoundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion AreaBoundary. This boundary is used for establishing effluent release limits with respect to therequirements of 10CFR20. See also CPNPP Emergency Plan Appendix E, Complex andOwner Controlled Area and CCNPP ODCM Section 5.0 Design Features.CPNPP Basis:Dose assessments are performed by computer-based method (ref. 1)NEI 99-01 Basis:This IC addresses a release of gaseous radioactivity that results in projected or actual offsitedoses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includesboth monitored and un-monitored releases. Releases of this magnitude are associated withthe failure of plant systems needed for the protection of the public.Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plant conditionsalone. The inclusion of both plant condition and radiological effluent EALs more fully addressesthe spectrum of possible accident events and conditions.The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDEwas established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to theenvironment is established. If the effluent flow past an effluent monitor is known to havestopped due to actions to isolate the release path, then the effluent monitor reading is noI Page 41 of 276 ATTACHMENT 1EAL Baseslonger valid for classification purposes.Escalation of the emergency classification level would be via IC AG-I-RGI.CPNPP Basis Reference(s):1. EPP-303 Operation of Computer Based Dose Assessment System2. NEI 99-01 AS1Page 42 of 276 ATTACHMENT 1EAL BasesCategory: R -Abnormal Rad Levels / Rad EffluentSubcategory: 1 -Radiological EffluentInitiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than100 mrem TEDE or 500 mrem thyroid CDEEAL:RS1.3 Site Area EmergencyField survey results indicate EITHER of the following at or beyond the EXCLUSION AREABOUNDARY:" Closed window dose rates greater than 100 mR/hr expected to continue for greaterthan or equal to 60 min." Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for 60min. of inhalation.(Notes 1, 2)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the releaseduration has exceeded the specified time limit.Mode Applicability:AllDefinition(s):EXCLUSION AREA BOUNDARY -Exclusion Area Boundary is a synonymous term for SiteBoundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion AreaBoundary. This boundary is used for establishing effluent release limits with respect to therequirements of 10CFR20. See also CPNPP Emergency Plan Appendix E, Complex andOwner Controlled Area and CCNPP ODCM Section 5.0 Design Features.CPNPP Basis:EPP-309 Onsite/In-Plant Radiological Surveys and Offsite Radiological Monitoring providesguidance for emergency or post-accident radiological environmental monitoring (ref. 1).NEI 99-01 Basis:This IC addresses a release of gaseous radioactivity that results in projected or actual offsitedoses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includesboth monitored and un-monitored releases. Releases of this magnitude are associated withthe failure of plant systems needed for the protection of the public.Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plant conditionsalone. The inclusion of both plant condition and radiological effluent EALs more fully addressesthe spectrum of possible accident events and conditions.Page 43 of 276 ATTACHMENT 1EAL BasesThe TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDEwas established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.ClassificAtion on effluent readings as.sum.es. that a releaso path to theenViGronment is established. if the cifluent flo)W past an effluent moenitor is kno)Wn to haVostepped due to- ac-tions to isolate the relcasc path, then the enffluent monitor rea;ding is nolonger valid for classification purposes.Escalation of the emergency classification level would be via IC AG4-RGI.CPNPP Basis Reference(s):1. EPP-309 Onsite/In-Plant Radiological Surveys and Offsite Radiological Monitoring2. NEI 99-01 AS1Page 44 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:R -Abnormal Rad Levels / Rad Effluent1 -Radiological EffluentRelease of gaseous radioactivity resulting in offsite dose greater than1,000 mrem TEDE or 5,000 mrem thyroid CDEEAL:RGI.1 General EmergencyReading on any Table R-1 effluent radiation monitor greater than column "GE" for greaterthan or equal to 15 min.(Notes 1,2, 3, 4)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the releaseduration has exceeded the specified time limit.Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path isisolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RSI.1 and RGI.1 should be usedfor emergency classification assessments until the results from a dose assessment using actualmeteorology are available.Table R-1 Effluent Monitor Classification ThresholdsRelease Point Monitor GE SAE Alert UEPlant Vent X-RE-5567.............6.52E-4 pCi/mlPVG384 + PVG385 A + BPlant Vent(WRGM)X-RE-5570S(WRGM) X-E-7 A + B 4.OE+7 luCi/sec 4.OE+6 liCi/sec 4.OE+5 [iCi/sec 4.OE+4 pCi/sec0 PVF684 + PVF685Main SteamMSLu78 u-RE-2325MSLu79 u-RE-2326 90 p.ci/ml* 9.0 ILci/ml* 0.9 pLCi/ml* 2 x high alarmMSLu80 u-RE-2327 setpoint*MSLu81 u-RE-2328Liquid Waste X-RE-5253 ...............-2 x high alarmLWE-076 setpoint"7 Service Wateru-RE-4269 2 x high alarm----..-------..x.hgh.larSSu65 u-RE-4270 setpointSSWu66* with reactor shutdownMode Applicability:AllPage 45 of 276 ATTACHMENT 1EAL BasesDefinition(s):NoneCPNPP Basis:This EAL address gaseous radioactivity releases, that for whatever reason, cause effluentradiation monitor readings corresponding to site boundary doses that exceed either:* 1000 mRemTEDE* 5000 mRem CDE ThyroidThe column "GE" gaseous effluent release values in Table R-1 correspond to calculated dosesof 100% of the EPA Protective Action Guidelines (TEDE or CDE Thyroid) (ref. 1).NEI 99-01 Basis:This IC addresses a release of gaseous radioactivity that results in projected or actual offsitedoses greater than or equal to the EPA Protective Action Guides (PAGs). It includes bothmonitored and un-monitored releases. Releases of this magnitude will require implementationof protective actions for the public.Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plant conditionsalone. The inclusion of both plant condition and radiological effluent EALs more fully addressesthe spectrum of possible accident events and conditions.The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE wasestablished in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to theenvironment is established. If the effluent flow past an effluent monitor is known to havestopped due to actions to isolate the release path, then the effluent monitor reading is nolonger valid for classification purposes.CPNPP Basis Reference(s):1. EAL Section R Revision 6 Table R-1 Effluent Monitor Classification Thresholds Review2. NEI 99-01 AG1I Page 46 of 276 ATTACHMENT 1EAL BasesCategory: R -Abnormal Rad Levels / Rad EffluentSubcategory: 1 -Radiological EffluentInitiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than1,000 mrem TEDE or 5,000 mrem thyroid CDEEAL:RGI.2 General EmergencyDose assessment using actual meteorology indicates doses greater than 1,000 mremTEDE or5,000 mrem thyroid CDE at or beyond the EXCLUSION AREA BOUNDARY (Notes 3, 4)Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path isisolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: The pre-calculated effluent monitor values presented in EALs RAI.1, RS1.1 and RGI.1 should be usedfor emergency classification assessments until the results from a dose assessment using actualmeteorology are available.Mode Applicability:AllDefinition(s):EXCLUSION AREA BOUNDARY- Exclusion Area Boundary is a synonymous term for SiteBoundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion AreaBoundary. This boundary is used for establishing effluent release limits with respect to therequirements of 10CFR20. See also CPNPP Emergency Plan Appendix E, Complex andOwner Controlled Area and CCNPP ODCM Section 5.0 Design Features.CPNPP Basis:Dose assessments are performed by computer-based method (ref. 1)NEI 99-01 Basis:This IC addresses a release of gaseous radioactivity that results in projected or actual offsitedoses greater than or equal to the EPA Protective Action Guides (PAGs). It includes bothmonitored and un-monitored releases. Releases of this magnitude will require implementationof protective actions for the public.Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plant conditionsalone. The inclusion of both plant condition and radiological effluent EALs more fully addressesthe spectrum of possible accident events and conditions.The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE wasestablished in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to theenvironment is established. If the effluent flow past an effluent monitor is known to havePage 47 of 276 ATTACHMENT 1EAL Basesstopped due to actions to isolate the release path, then the effluent monitor reading is nolonger valid for classification purposes.CPNPP Basis Reference(s):1. EPP-303 Operation of Computer Based Dose Assessment System2. NEI 99-01 AG1Page 48 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:R -Abnormal Rad Levels / Rad Effluent1 -Radiological EffluentRelease of gaseous radioactivity resulting in offsite dose greater than1,000 mrem TEDE or 5,000 mrem thyroid CDEEAL:RG1.3 General EmergencyField survey results indicate EITHER of the following at or beyond the EXCLUSION AREABOUNDARY:" Closed window dose rates greater than 1,000 mR/hr expected to continue for greaterthan or equal to 60 min." Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrem for60 min. of inhalation.(Notes 1, 2)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the releaseduration has exceeded the specified time limit.Mode Applicability:AllDefinition(s):EXCLUSION AREA BOUNDARY- Exclusion Area Boundary is a synonymous term for SiteBoundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion AreaBoundary. This boundary is used for establishing effluent release limits with respect to therequirements of 10CFR20. See also CPNPP Emergency Plan Appendix E, Complex andOwner Controlled Area and CCNPP ODCM Section 5.0 Design Features.CPNPP Basis:EPP-309 Onsite/In-Plant Radiological Surveys and Offsite Radiological Monitoring providesguidance for emergency or post-accident radiological environmental monitoring (ref. 1).NEI 99-01 Basis:This IC addresses a release of gaseous radioactivity that results in projected or actual offsitedoses greater than or equal to the EPA Protective Action Guides (PAGs). It includes bothmonitored and un-monitored releases. Releases of this magnitude will require implementationof protective actions for the public.Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plant conditionsalone. The inclusion of both plant condition and radiological effluent EALs more fully addressesthe spectrum of possible accident events and conditions.Page 49 of 276 ATTACHMENT 1EAL BasesThe TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE wasestablished in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a rolcase path to theenvionmet is established. if the effluent flow past an cifluent monitor is known to aestopped duo to cin to- isolate the release path, then the effluont moenitor reading is nolo~nger valid for clasSiffication purposes.CPNPP Basis Reference(s):1. EPP-309 Onsite/In-Plant Radiological Surveys and Offsite Radiological Monitoring2. NEI 99-01 AG1Page 50 of 276 1 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:EAL:R -Abnormal Rad Levels / Rad Effluent2 -Irradiated Fuel EventUnplanned loss of water level above irradiated fuelRU2.1 Unusual EventUNPLANNED water level drop in the REFUELING PATHWAY as indicated by low waterlevel alarm or indicationANDUNPLANNED rise in corresponding area radiation levels as indicated by any Table R-2area radiation monitorsTable R-2 SFP & Refueling Cavity Area Radiation MonitorsSFP:* SFP001, LRAM SFP 2 E WALL (X-RE-6272)" SFP002, LRAM SFP 2 N WALL (X-RE-6273)" SFP003, LRAM SFP 1 E WALL (X-RE-6274)* SFP004, LRAM SFP 1 S WALL (X-RE-6275)Refueling Cavity:* RFCulO, LRAM W REFUEL CAV860 (u-RE-6251)* RFCu12, LRAM E REFUEL CAV 860 (u-RE-6253)Mode Applicability:AllDefinition(s):UNPLANNED-. A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient. The cause of the parameter changeor event may be known or unknown.REFUELING PATHWAY-. The reactor refueling cavity, spent fuel pool and fuel transfer canalcomprise the refueling pathway.CPNPP Basis:Indication of decreasing level includes Spent Fuel Pool Panel (FB 810) (ref. 2):" RFL CAVITY 1 LEVEL LO (upender area) (SFP-3.3)* RFL CAVITY 2 LEVEL LO (upender area) (SFP-3.7)" SFP 1 LEVEL LO (SFP-3.9)* SFP 2 LEVEL LO (SFP-3.10)Page 51 of 276 ATTACHMENT 1EAL Bases* RFL CAVITY 1 LEVEL LO (vessel area) (SFP-4.3)* RFL CAVITY 2 LEVEL LO (vessel area) (SFP-4.7)* SFP 1 TRANSFER CANAL LEVEL LO (SFP-4.9)* SFP 2 TRANSFER CANAL LEVEL LO (SFP-4.10)Allowing level to decrease could result in spent fuel being uncovered, reducing spent fueldecay heat removal and creating an extremely hazardous radiation environment. TechnicalSpecification Section 3.7.15 (ref. 4) requires at least 23 ft of water above the Spent Fuel Poolstorage racks (857' 31/2") (ref. 2). Technical Specification Section 3.9.7 (ref. 5) requires at least23 ft of water above the Reactor Vessel flange in the refueling cavity (856' 11" in refuelingcavity or 407" above core plate) (ref. 6). During refueling, this maintains sufficient water level inthe fuel transfer canal, refueling cavity, and SFP to retain iodine fission product activity in thewater in the event of a fuel handling accident. ABN-909, Spent Fuel Pool/Refueling CavityMalfunctions, provides appropriate guidance to restore and maintain normal water levels in thefuel transfer canal, refueling cavity, and SFP, and to determine if water levels have droppedbelow the Technical Specification LCOs (ref. 2). The fuel transfer canal is only of concern inassessing this EAL when irradiated fuel transfer is in progress, in which case the spent fuelpool gates are open and connected to the fuel transfer canal.NEI 99-01 Basis:This IC addresses a decrease in water level above irradiated fuel sufficient to cause elevatedradiation levels. This condition could be a precursor to a more serious event and is alsoindicative of a minor loss in the ability to control radiation levels within the plant. It is thereforea potential degradation in the level of safety of the plant.A water level decrease will be primarily determined by indications from available levelinstrumentation. Other sources of level indications may include reports from plant personnel(e.g., from a refueling crew) or video camera observations (if available). A significant drop inthe water level may also cause an increase in the radiation levels of adjacent areas that can bedetected by monitors in those locations.The effects of planned evolutions should be considered. For example, a refueling bridge arearadiation monitor reading may increase due to planned evolutions such as lifting of the reactorvessel head or movement of a fuel assembly. Note that this EAL is applicable only in caseswhere the elevated reading is due to an unplanned loss of water level.A drop in water level above irradiated fuel within the reactor vessel may be classified inaccordance Recognition Category C during the Cold Shutdown and Refueling modes.Escalation of the emergency classification level would be via IC AA2RA2.CPNPP Basis Reference(s):1. ABN-908 Fuel Handling Accident2. ABN-909 Spent Fuel Pool/Refueling Cavity Malfunctions3. 1-ALB-6B SFPCS TRBL4. Technical Specifications 3.7.15 Fuel Storage Area Water Level5. Technical Specifications 3.9.7 Refueling Cavity Water Level6. RFO-102A/B Refueling OperationsPage 52 of 276 ATTACHMENT IEAL Bases7. NEI 99-01 AU2Category: R -Abnormal Rad Levels / Rad EffluentSubcategory: 2 -Irradiated Fuel EventInitiating Condition: Significant lowering of water level above, or damage to, irradiated fuelEAL:RA2.1 Unusual EventUncovery of irradiated fuel in the REFUELING PATHWAYMode Applicability:AllDefinition(s):REFUELING PATHWAY-. The reactor refueling cavity, spent fuel pool and fuel transfer canalcomprise the refueling pathway.CPNPP Basis:None.NEI 99-01 Basis:This IC addresses events that have caused imminent or actual damage to an irradiated fuelassembly, or a significant lowering of water level within the spent fuel pool" see- .evek,.eFNetes). These events present radiological safety challenges to plant personnel and areprecursors to a release of radioactivity to the environment. As such, they represent an actualor potential substantial degradation of the level of safety of the plant.This IC applies to irradiated fuel that is ic~ensed for dry s~torage up to the point that theloaded storage cask is sealed. Once sealed, damage to a loaded cask causiRg loss of theCONFINEMIIENT BOUNDARY is classified in .c-, ... r ccord .A. ;ith IC E HUl.Esc-alation of the emergency would be bhasead on either Recognition Categor,' A Or CEAL-#-#This EAL escalates from AUJ2-RU2.1 in that the loss of level, in the affected portion of theREFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiatedfuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation(e.g., reports from personnel or camera images), as well as significant changes in water andradiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of availableindications, reports and observations.While an area radiation monitor could detect an increase in a dose rate due to a lowering ofwater level in some portion of the REFUELING PATHWAY, the reading may not be a reliableindication of whether or not the fuel is actually uncovered. To the degree possible, readingsshould be considered in combination with other available indications of inventory loss.A drop in water level above irradiated fuel within the reactor vessel may be classified inPage 53 of 276 ATTACHMENT 1EAL Basesaccordance Recognition Category C during the Cold Shutdown and Refueling modes.EAL4#2ThN" F.AI .. .nf rirnch'.. miftr~rn h,' ..tnirraiatd fel, Damaging eventS May include thc dropping, bumping or binding of aassembly, Or dro)pping a heavy load onto an assembly. A risc OR readingG on radiatoM9111101 -SHRUIG 08 GGR1S.-P-FP-Wyn W i0Iii tf vdd o'i"1bi-~tW1EAL.Hut t vn e.#3aue adigaý -trF= .f1.43 wt~ ~i tth""i r iti h n~ n n h ~~ in~n n~~t --.--. ....--. ..-.-. -prevent significafnt; do;se con Rsequences from direct gamma radiation to personnel performingemeratinn5G in the 'Anlnit'i of the speRnt fuel nee[ ThkS GE)RditiGR refkr~ts a ".innffictnt less of-. .....-J- .----.v..................... .... ...............ent Tuel pool water and thlus it is Also a PrecursorF to a less T me anniity toIn4 I I fI Idaacquately cool the irradiated Tuei assemoigis stored in the pool.Escalation of the emergency classification level would be via ICs AS-1-RS1 9F .AS2 (see.A,.DevýeFp-ee)CPNPP Basis Reference(s):1. ABN-909 Spent Fuel Pool/Refueling Cavity Malfunctions2. NEI 99-01 AA2Category: R -Abnormal Rad Levels / Rad EffluentSubcategory- 2 -Irradiated Fuel EventInitiating Condition: Significant lowering of water level above, or damage to, irradiated fuelEAL:RA2.2 AlertDamage to irradiated fuel resulting in a release of radioactivityANDHigh alarm on any of the following:" Any Table R-2 area radiation monitors* CAGu97, CNTMT AIR PIG GAS (u-RE-5503)* CAPu98, CNTMT AIR PIG PART (u-RE-5502)* CAIu99, CNTMT AIR PIG IODINE (u-RE-5566)" FBV088, FB VENT EXH (X-RE-5700)Table R-2 SFP & Refueling Cavity Area Radiation MonitorsSFP:* SFP001, LRAM SFP 2 E WALL (X-RE-6272)* SFP002, LRAM SFP 2 N WALL (X-RE-6273)" SFP003, LRAM SFP I E WALL (X-RE-6274)" SFP004, LRAM SFP 1 S WALL (X-RE-6275)Page 54 of 276 ATTACHMENT 1EAL BasesRefueling Cavity:* RFCulO, LRAM W REFUEL CAV860 (u-RE-6251)" RFCu12, LRAM E REFUEL CAV 860 (u-RE-6253)Mode Applicability:AllDefinition(s):NoneCPNPP Basis:The specified radiation monitors are those expected to see increase area radiation levels as aresult of damage to irradiated fuel (ref. 1, 2).The bases for the SFP ventilation radiation High alarm and the SFP and containment arearadiation readings are a spent fuel handling accident (ref. 1). In the Fuel Handling Building, afuel assembly could be dropped in the fuel transfer canal or in the SFP. Should a fuelassembly be dropped in the fuel transfer canal or in the SFP and release radioactivity above aprescribed level, the Fuel Handling Building ventilation monitors sound an alarm, alertingpersonnel to the problem.NEI 99-01 Basis:This IC addresses events that have caused imminent or actual damage to an irradiated fuelassembly, or a significant lowering of water level within the spent fuel pool(See -Deve..,..Netes). These events present radiological safety challenges to plant personnel and areprecursors to a release of radioactivity to the environment. As such, they represent an actualor potential substantial degradation of the level of safety of the plant.This IC applies to irradiated fuel that is lic-ensed fo-r dry storage up to the point that theloaded- storFage cask is sealed. Once sealed, damage to a loaded cask causing los6 of theCONFINEMI1ENT BOUNDARYis classified in ac.r.d. ,....with ICE HU.Escalation of the emergency would be based on either Recognition Category A-Ror C ICs.EAl lt--hs E,ATL ealates from in that the loss .. of level, inR the affected portion REFUELING P'HWY is f sufficin mgiuethaeruldinuncoYer' of irdaefuel. IndicatioNs of irradiated fuel uncover' may include direct or indirect visual l bselVati1(e.g., reports from personn~el or caer imaes), as well as significant changes in water andradiation levels, or other plant parametes.. o.m.putationa aids may also be used (e.g., a boiloff cu.e). Classification .f aR event using this EA.L should be based n the totality of availableindications, reports and obsevations.While an area radiation monitor could detect an increase in a dose rate due to alowering of water level in som e portion of the REF UELING PIADTHWAIAV the Feading may not be-A reliable indication of whetheror not t~he fu-'el is actually uncovered. To the degree possible,readings should bhe cosidered in combination With o-th-ravailable indicationS of invento' losAdrop in water level above irradiated fuel within the reactor vessel ma be classifiediaccordance Recognition Categor,' C duFrig the Co-ld Shutdo'A' and Refueling modes.Page 55 of 276 ATTACHMENT 1EAL BasesThis EAL addresses a release of radioactive material caused by mechanical damage toirradiated fuel. Damaging events may include the dropping, bumping or binding of anassembly, or dropping a heavy load onto an assembly. A rise in readings on radiationmonitors should be considered in conjunction with in-plant reports or observations of apotential fuel damaging event (e.g., a fuel handling accident).EAL i,,Spent fuel poel waterlevel at this value is within the lower end of the level range necessar,. to prevent S do6e conSequences from direct gamma radiation to perSonnel p.. rrming operations in thevicinity of the spent fuel pool. This condition reflects a signific-ant loss of spent fuel pool wtei nventor,' and thus it is also a prec-ursor to a loss of the ability to adequately cool the irdaefuel assembles stored in the pool.Escalation of the emergency classification level would be via ICes ASI--RS1 9FAS2-(seee-AS22D'evelepeF-Notes).CPNPP Basis Reference(s):1. ABN-908 Fuel Handling Accident2. ABN-909 Spent Fuel Pool/Refueling Cavity Malfunctions3. NEI 99-01 AA2I Page 56 of 276 ATTACHMENT IEAL BasesCategory:Subcategory:R -Abnormal Rad Levels / Rad Effluent2 -Irradiated Fuel EventInitiating Condition: Significant lowering of water level above, or damage to, irradiated fuelEAL:RA2.3 AlertLowering of spent fuel pool level to El. 844.3' (Level 2)Mode Applicability:AllDefinition(s):NoneCPNPP Basis:Post-Fukushima order EA-12-051 (ref. 1) required the installation of reliable SFP levelindication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of thefuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3) (ref. 1).Level 2 is the level that is adequate to provide substantial radiation shielding for a personstanding on the spent fuel pool operating deck. It represents the range of water level whereany necessary operations in the vicinity of the spent fuel pool can be completed withoutsignificant dose consequences from direct gamma radiation from the stored spent fuel.Comanche Peak designated as Level 2 the water level 10 feet (+/- 1.0 foot) above the top of thefuel racks (El 844' -2.75" rounded to 844.3' indicated) (ref. 2).The enhanced SFP level instruments (X-LI-4876, 4878, 4877, 4879) do not have indicationavailable in the control room and must be read remotely outside of the control room.NEI 99-01 Basis:This IC addresses events that have caused imminent or actual damage to an irradiated fuelassembly, or a significant lowering of water level within the spent fuel pool- (see Nte4. These events present radiological safety challenges to plant personnel and areprecursors to a release of radioactivity to the environment. As such, they represent an actualor potential substantial degradation of the level of safety of the plant.This IC applies to irradiated fuel that is licensed for storage up to the point that theloaded storage cask is sealed. Oncc sealed, damage to a loaded cask causing loss of theCONFINEMENT BOUINDARY is classified on with IC E HUE--Escalation of the emergency would be based on either Recognition Category A-Ror C1('q~ PA L44T1,kQ P~A L imi==t= m Ul 2 on- th-n+ +j-P Aflq Af I'PUP' on th'-' 'ff4PL~tp -~.nndon nf ih,-nR=FUELIN 1F=1 HAY, is of sufficient magnitude to hav.e resulted in uncoVery of irrFadiatedfuel. Indications of iRradiated fuel uncovery may include direct or indirect visual observation(e.g., reports from personnel or caer iags), as well as significant changes in water andr-adiatien levels, or other plant par-amieteirs. Coamputational aids may also be used (e.g., a boilindicationS, repo-ts and ,bseR,'atkiN4.Page 57 of 276 ATTACHMENT 1EAL BasesWhile an area radiatienR uld detect aninreae a dose rate due to al-w-oing ef water level in some portion of the REFUELING PATHlw Y, the reading may not bea reliable indicatio enf whether r icnt the fuel is acually urcgvered. To the degree possible,readings should be onsidered ini cofbiatioe with other available indieation9res of i gniVefto loA drfp in water level above inradiated fuel within the reacutor el ay be clasified inaccordance Recognition Categor,' C during the Cold Shu~tdoWn and Refueling moedes.This EAL addrpesrspes a- rpelpease o-f radioactive maeral cased by mnechan~ical damnage toirradiated fuel. Damaging evenss may include the dropping, bLmping or binding oassembly, or dropping a heavy load onRto an assembly. A rise in readings on radiatomonitors should be considered in conjunction With in plant reports or o~bservations of apotential fuel damaging event (e.g., a fuel handling accident).AL#Spent fuel pool water level at this value is within the lower end of the level rangenecessary to prevent significant dose consequences from direct gamma radiation to personnelperforming operations in the vicinity of the spent fuel pool. This condition reflects a significantloss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability toadequately cool the irradiated fuel assembles stored in the pool.Escalation of the emergency classification level would be via ICs AS4-RS1-, A$2 A9/- p A,,., -I -Page 58 of 276 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):1. NRC EA-12-51 Issuance of Order to Modify Licenses with Regard to Reliable Spent FuelPool Instrumentation2. TXX-13103 Overall Integrated Plan in Response to March 12,2012 Commission OrderModifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (OrderNumber EA-12-051), Response to Request for Additional Information3 NEI 99-01 AA2Page 59 of 276 ATTACHMENT 1EAL BasesCategory: R -Abnormal Rad Levels / Rad EffluentSubcategory: 2- Irradiated Fuel EventInitiating Condition: Spent fuel pool level at the top of the fuel racksEAL:RS2.1 Site Area EmergencyLowering of spent fuel pool level to El. 835.3' (Level 3)Mode Applicability:AllDefinition(s):NoneCPNPP Basis:Post-Fukushima order EA-12-051 (ref. 1) required the installation of reliable SFP levelindication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of thefuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3) (ref. 1).Level 3 is the level where fuel remains covered and actions to implement make-up wateraddition should no longer be deferred. Level 3 corresponds nominally (i.e., +/- 1 foot) to thehighest point of any fuel rack seated in the spent fuel pool. Level 3 is defined in this manner toprovide the maximum range of information to operators, decision makers and emergencyresponse personnel. Comanche Peak designated as Level 3 the water level greater than 1 footabove the top of the fuel storage racks plus the accuracy of the SFP level instrument channel(El. 835' -2.75" rounded to 835.3' indicated). Designation of this level as Level 3 isconservative; its selection assures that the fuel will remain covered, and at that point therewould be no functional or operational reason to defer action to implement the addition of make-up water to the pool (ref. 2).The enhanced SFP level instruments (X-LI-4876, 4878, 4877, 4879) do not have indicationavailable in the control room and must be read remotely outside of the control room.NEI 99-01 Basis:This 1C-EAL addresses a significant loss of spent fuel pool inventory control and makeupcapability leading to IMMINENT fuel damage. This condition entails major failures of plantfunctions needed for protection of the public and thus warrant a Site Area Emergencydeclaration.It is recognized that this IC would likely not be met until well after another Site Area EmergencyIC was met; however, it is included to provide classification diversity.Escalation of the emergency classification level would be via IC AG1 or AG2RG2.I Page 60 of 276 1 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):1. NRC EA-12-51 Issuance of Order to Modify Licenses with Regard to Reliable Spent FuelPool Instrumentation2. TXX-13103 Overall Integrated Plan in Response to March 12,2012 Commission OrderModifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (OrderNumber EA-12-051), Response to Request for Additional Information3. NEI 99-01 AS2I Page 61 of 276 ATTACHMENT 1EAL BasesCategory: R -Abnormal Rad Levels / Rad EffluentSubcategory: 2 -Irradiated Fuel EventInitiating Condition: Spent fuel pool level cannot be restored to at least the top of the fuelracks for 60 minutes or longerEAL:RG2.1 General EmergencySpent fuel pool level cannot be restored to at least El. 835.3' (Level 3) for greater than orequal to 60 min. (Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Mode Applicability:AllDefinition(s):NoneCPNPP Basis:Post-Fukushima order EA-12-051 (ref. 1) required the installation of reliable SFP levelindication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of thefuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3) (ref. 1).Level 3 is the level where fuel remains covered and actions to implement make-up wateraddition should no longer be deferred. Level 3 corresponds nominally (i.e., +/- 1 foot) to thehighest point of any fuel rack seated in the spent fuel pool. Level 3 is defined in this manner toprovide the maximum range of information to operators, decision makers and emergencyresponse personnel. Comanche Peak designated as Level 3 the water level greater than 1 footabove the top of the fuel storage racks plus the accuracy of the SFP level instrument channel(El. 835' -2.75" rounded to 835.3' indicated). Designation of this level as Level 3 isconservative; its selection assures that the fuel will remain covered, and at that point therewould be no functional or operational reason to defer action to implement the addition of make-up water to the pool (ref. 2).The enhanced SFP level instruments (X-LI-4876, 4878, 4877, 4879) do not have indicationavailable in the control room and must be read remotely outside of the control room.NEI 99-01 Basis:This I,-EAL addresses a significant loss of spent fuel pool inventory control and makeupcapability leading to a prolonged uncovery of spent fuel. This condition will lead to fueldamage and a radiological release to the environment.It is recognized that this IC would likely not be met until well after another General EmergencyIC was met; however, it is included to provide classification diversity.Page 62 of 276 ATTACHMENT IEAL BasesCPNPP Basis Reference(s):1. NRC EA-12-51 Issuance of Order to Modify Licenses with Regard to Reliable Spent FuelPool Instrumentation2. TXX-13103 Overall Integrated Plan in Response to March 12,2012 Commission OrderModifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (OrderNumber EA-12-051), Response to Request for Additional Information3. NEI 99-01 AG2Page 63 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:EAL:R -Abnormal Rad Levels / Rad Effluent3 -Area Radiation LevelsRadiation levels that IMPEDE access to equipment necessary fornormal plant operations, cooldown or shutdownRA3.1 AlertDose rates greater than 15 mR/hr in EITHER of the following areas:Control RoomCRM048 (X-RE-6281) or CRM049 (X-RE-6282)ORCentral Alarm Station (by survey)Mode Applicability:AllDefinition(s):NoneCPNPP Basis:X-RE-6281 and X-RE-6282 are the installed Control Room area radiation monitors and may beused to assess this EAL threshold (range of 1 E-4 to 1 E+5 mR/hr). However, no permanentlyinstalled area radiation monitoring is installed in the CAS and therefore this threshold must beassessed via local radiation survey (ref. 1).NEI 99-01 Basis:This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to precludeor impede personnel from performing actions necessary to maintain normal plant operation, orto perform a normal plant cooldown and shutdown. As such, it represents an actual orpotential substantial degradation of the level of safety of the plant. The ...Emergqency Coordinator should consider the cause of the increased radiation levels anddetermine if another IC may be applicable.For EAr. #2, an Alert delara-onj, is warranted ifoperatinG mode in effect at the time of the elevated radiation levels. The emerenc'-S;~ ~atcnnir'i~ntinn ~''hrh~r ~nt" kint ini' nr'~rv t th~ tmr~nf hr~ ----....- .J-increased radiation levels. Access should be considered as impeded if extraordinary.- aa , .a ara.,aa co .,+a fn H;+ I+a .+ -Fr, ~r arrr .nna Ltnia l, a +k^ ad raa rnIa -a Ia ----. .IVinstalling temporary shielding, reqIu"irig Use of non rou1tine protective equipment, requesting an4extension in dose limits beyond nremal administrative lmt)An emergency declaration is not wratdif any of the folloWing conditions apply-.aThe plant is in an operating mode differet than the mode specified for the affectedroom/area (i.e., entry is not required during the operating mode in effect at the time otPage 64 of 276 ATTACHMENT 1EAL Basesthe elevated radiation levels). For example, the plant is in Mode 1 whcn the radiationincrease occurFs, and the proGedures used for normnal opcration, cooldoWn andshutdW-n dGo nt require ent' ; into the affected room until Mode 4.alThe incfeased radiation levels are a result of a planed acgtivity that incRlucomnpensator,' measures which addrcss the temporar,' inaGeeSSibility of a room or area(e.g., radiography, spcnt filter or re-sin transfer, etc.).*The action for which room/area entr,'is required is of an administrative or recorFdkeeping nature (e.g., nr9mal rounds or routine inspections).-*The access control measures are of a conservative or precauitionar,' natur~e, and would1not actually prevent or impede a required action-.Escalation of the emergency classification level would be via Recognition Category AýR, C or FICs.CPNPP Basis Reference(s):1. DBD-EE-023 Radiation Monitoring System2. NEI 99-01 AA3Page 65 of 276 ATTACHMENT IEAL BasesCategory: R -Abnormal Rad Levels / Rad EffluentSubcategory: 3- Area Radiation LevelsInitiating Condition: Radiation levels that IMPEDE access to equipment necessary fornormal plant operations, cooldown or shutdownEAL:RA3.2 AlertAn UNPLANNED event results in radiation levels that prohibit or IMPEDE access to anyTable R-3 rooms or areas (Note 5)Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, thenno emergency classification is warranted.Table R-3 Safe Operation & Shutdown Rooms/AreasRoom/Area Mode ApplicabilityCharging Pump Rooms 1, 2, 3, 4, 5, 6CVCS Valve Rooms 1, 2, 3, 4, 5, 61 E Switchgear Rooms AllRHR Pump Rooms 4, 5, 6Mode Applicability:AllDefinition(s):IMPEDE(D) -Personnel access to a room or area is hindered to an extent that extraordinarymeasures are necessary to facilitate entry of personnel into the affected room/area(e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).UNPLANNED-. A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient. The cause of the parameter changeor event may be known or unknown.CPNPP Basis:For this EAL, area or room access is considered impeded if radiation levels require locked highradiation controls to to be imposed.If the equipment in the listed room or area was already inoperable, or out-of-service, before theevent occurred, then no emergency should be declared since the event will have no adverseimpact beyond that already allowed by Technical Specifications at the time of the event.The list of plant rooms or areas with entry-related mode applicability identified specify thoserooms or areas that contain equipment which require a manual/local action as specified inoperating procedures used for normal plant operation, cooldown and shutdown. Rooms orareas in which actions of a contingent or emergency nature would be performed (e.g., anaction to address an off-normal or emergency condition such as emergency repairs, correctivePage 66 of 276 ATTACHMENT 1EAL Basesmeasures or emergency operations) are not included. In addition, the list specifies the plantmode(s) during which entry would be required for each room or area (ref. 1).NEI 99-01 Basis:This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to precludeor impede personnel from performing actions necessary to maintain normal plant operation, orto perform a normal plant cooldown and shutdown. As such, it represents an actual orpotential substantial degradation of the level of safety of the plant. The ergiRyDirecteFEmergency Coordinator should consider the cause of the increased radiation levelsand determine if another IC may be applicable.I For EAL- #2RA3.2, an Alert declaration is warranted if entry into the affected room/area is, ormay be, procedurally required during the plant operating mode in effect at the time of theelevated radiation levels. The emergency classification is not contingent upon whether entry isactually necessary at the time of the increased radiation levels. Access should be consideredas impeded if extraordinary measures are necessary to facilitate entry of personnel into theaffected room/area (e.g., installing temporary shielding, requiring use of non-routine protectiveequipment, requesting an extension in dose limits beyond normal administrative limits).An emergency declaration is not warranted if any of the following conditions apply:" The plant is in an operating mode different than the mode specified for the affectedroom/area (i.e., entry is not required during the operating mode in effect at the time ofthe elevated radiation levels). For example, the plant is in Mode 1 when the radiationincrease occurs, and the procedures used for normal operation, cooldown andshutdown do not require entry into the affected room until Mode 4." The increased radiation levels are a result of a planned activity that includescompensatory measures which address the temporary inaccessibility of a room or area(e.g., radiography, spent filter or resin transfer, etc.)." The action for which room/area entry is required is of an administrative or recordkeeping nature (e.g., normal rounds or routine inspections)." The access control measures are of a conservative or precautionary nature, and wouldnot actually prevent or impede a required action.Escalation of the emergency classification level would be via Recognition Category AR, C or FICs.CPNPP Basis Reference(s):1. Attachment 3 Safe Operation & Shutdown Areas Tables R-3 & H-2 Bases2. NEI 99-01 AA3Page 67 of 276 ATTACHMENT IEAL BasesCategory E -Independent Spent Fuel Storage Installation (ISFSI)EAL Group: Any (EALs in this category are applicable to anyplant condition, hot or cold.)An independent spent fuel storage installation (ISFSI) is a complex that is designed andconstructed for the interim storage of spent nuclear fuel and other radioactive materialsassociated with spent fuel storage. A significant amount of the radioactive material containedwithin a canister must escape its packaging and enter the biosphere for there to be asignificant environmental effect resulting from an accident involving the dry storage of spentnuclear fuel.An Unusual Event is declared on the basis of the occurrence of an event of sufficientmagnitude that a loaded cask confinement boundary is damaged or violated.IPage 68 of 276 ATTACHMENT 1EAL BasesCategory: ISFSISubcategory: Confinement BoundaryInitiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARYEAL:EUl.1 Unusual EventDamage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contactradiation reading greater than EITHER:* 60 mrem/hr (-r + rq) on the top of the overpack* 600 mrem/hr (T" + rj on the side of the overpack (excluding inlet and outlet ducts)Mode Applicability:AllDefinition(s):CONFINEMENT BOUNDARY-. The barrier(s) between spent fuel and the environment oncethe spent fuel is processed for dry storage. As applied to the CPNPP ISFSI, theCONFINEMENT BOUNDARY is defined to be the Multi-Purpose Canister (MPC).CPNPP Basis:The ISFSI includes the dry-cask storage system, the cask transfer facility, onsite transporter,and the storage pads. The dry-cask storage system is the HI-STORM 100 System. This is acanister-based storage system that stores spent nuclear fuel in a vertical orientation. Itconsists of three discrete components: the MPC, the HI-TRAC 125 Transfer Cask, and the HI-STORM 100 System Overpack. The MPC provides the confinement boundary for the storedfuel. The HI-TRAC 125 Transfer Cask provides radiation shielding and structural protection ofthe MPC during transfer operations, while the storage overpack provides radiation shieldingand structural protection of the MPC during storage (ref. 1).The value shown represents 2 times the limits specified in the ISFSI Certificate of ComplianceTechnical Specification section 5.7.4 for radiation external to a loaded MPC overpack (ref. 1).NEI 99-01 Basis:This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of astorage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storagebeginning at the point that the loaded storage cask is sealed. The issues of concern are thecreation of a potential or actual release path to the environment, degradation of one or morefuel assemblies due to environmental factors, and configuration changes which could causechallenges in removing the cask or fuel from storage.The existence of "damage" is determined by radiological survey. The technical specificationI multiple of "2 times", which is also used in Recognition Category A-RIC RAU1, is used here toI Page 69 of 276 ATTACHMENT IEAL Basesdistinguish between non-emergency and emergency conditions. The emphasis for thisclassification is the degradation in the level of safety of the spent fuel cask and not themagnitude of the associated dose or dose rate. It is recognized that in the case of extremedamage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may bedetermined based on measurement of a dose rate at some distance from the cask.Security-related events for ISFSIs are covered under ICs HU1 and HAl.CPNPP Basis Reference(s):1. HI-2104635, HI-STORM Certificate of Compliance Appendix A Technical SpecificationSection 5.7.42. RPI-792 HI-STORM Overpack Surface Dose Rates3. NEI 99-01 E-HU1Page 70 of 276 ATTACHMENT 1EAL BasesCategory C -Cold Shutdown / Refueling System MalfunctionEAL Group: Cold Conditions (RCS temperature < 2000F); EALsin this category are applicable only in one or morecold 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) duringthese periods, the consequences of any given initiating event can vary greatly. For example, aloss of decay heat removal capability that occurs at the end of an extended outage has lesssignificance than a similar loss occurring during the first week after shutdown. Compoundingthese events is the likelihood that instrumentation necessary for assessment may also beinoperable. The cold shutdown and refueling system malfunction EALs are based onperformance capability to the extent possible with consideration given to RCS integrity,containment closure, and fuel clad integrity for the applicable operating modes (5 -ColdShutdown, 6 -Refueling, D -Defueled).The events of this category pertain to the following subcategories:1. RCS LevelRCS water level is directly related to the status of adequate core cooling and, therefore,fuel clad integrity.2. Loss of Emerqency AC PowerLoss of emergency plant electrical power can compromise plant safety system operabilityincluding decay heat removal and emergency core cooling systems which may benecessary to ensure fission product barrier integrity. This category includes loss of onsiteand offsite power sources for 6.9 KV AC emergency buses.3. RCS TemperatureUncontrolled or inadvertent temperature or pressure increases are indicative of a potentialloss of safety functions.4. Loss of Vital DC PowerLoss of emergency plant electrical power can compromise plant safety system operabilityincluding decay heat removal and emergency core cooling systems which may benecessary to ensure fission product barrier integrity. This category includes loss of power toor degraded voltage on the 125V DC vital buses.5. Loss of CommunicationsCertain events that degrade plant operator ability to effectively communicate with essentialpersonnel within or external to the plant warrant emergency classification.6. Hazardous Event Affecting Safety SystemsCertain hazardous natural and technological events may result in visible damage to ordegraded performance of safety systems warranting classification.Page 71 of 276 1 ATTACHMENT 1EAL BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 1 -RCS LevelInitiating Condition: UNPLANNED loss of RCS inventory for 15 minutes or longerEAL:CUI.1 Unusual EventUNPLANNED loss of reactor coolant results in RCS water level less than a required lowerlimit for greater than or equal to 15 min. (Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Mode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):UNPLANNED-. A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient. The cause of the parameter changeor event may be known or unknown.CPNPP Basis:With the plant in Cold Shutdown, RCS water level is normally maintained above thepressurizer low level setpoint of 17% (ref. 1). However, if RCS level is being controlled belowthe pressurizer low level setpoint, or if level is being maintained in a designated band in thereactor vessel it is the inability to maintain level above the low end of the designated controlband due to a loss of inventory resulting from a leak in the RCS that is the concern.With the plant in Refueling mode, RCS water level is normally maintained at or above thereactor vessel flange (Technical Specification LCO 3.9.7 requires at least 23 ft. of water abovethe top of the reactor vessel flange in the refueling cavity during refueling operations) (ref. 2).The Reactor Vessel flange level is 834' 1/2" elevation or 132.5 in. above the upper core plate(top) (ref. 3).NEI 99-01 Basis:This IC addresses the inability to restore and maintain water level to a required minimum level(or the lower limit of a level band), or a loss of the ability to monitor .. .....-v-..-,IRCS {RPWR],r RPV- r[{BVR)-level concurrent with indications of coolant leakage. Either of these conditionsis considered to be a potential degradation of the level of safety of the plant.Refueling evolutions that decrease RCS-water inventory are carefully planned and controlled.An UNPLANNED event that results in water level decreasing below a procedurally requiredlimit warrants the declaration of an Unusual Event due to the reduced water inventory that isavailable to keep the core covered.This EAL-#4 recognizes that the minimum required [P.4'.] or RPV [BW,4..)level can change several times during the course of a refueling outage as different plantconfigurations and system lineups are implemented. This EAL is met if the minimum level,-Page 72 of 276 ATTACHMENT IEAL Basesspecified for the current plant conditions, cannot be maintained for 15 minutes or longer. Theminimum level is typically specified in the applicable operating procedure but may be specifiedin another controlling document.The 15-minute threshold duration allows sufficient time for prompt operator actions to restoreand maintain the expected water level. This criterion excludes transient conditions causing abrief lowering of water level.EAL o #2 addresses a condition ,,,here all means to determ.ine (reactoF vesse.R.S[PI/lRI or RPV [B3VARj) level have been lost. In tthis, cond-iqtion, operators may determinle that aninventory loss is occurring by ebseiR'ig changes1 inSu1mp anmd/ortank levels. Sump and/eFtank level chng Smut bhe evaluated against other potential sources of water flow to ensur~eth e" a re n di r-t:-; Zt ofp l11e.a ka ge fro m th e (re acGtor '.GeS P-VRCGS2 [P- tIP] orF RPV [BWR]).Continued loss of RCS inventory may result in escalation to the Alert emergency classificationlevel via either IC CAl or CA3.CPNPP Basis Reference(s):1. ALM-0052A/B Alarm Procedure u-ALB-5B3 (513-3.6)2. Technical Specification Section 3.9.7 Refueling Cavity Water Level3. IPO-Ol OAIB Reactor Coolant System Reduced Inventory Operations4. NEI 99-01 CUII Page 73 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:C -Cold Shutdown / Refueling System Malfunction1 -RCS LevelInitiating Condition: UNPLANNED loss of RCS inventory for 15 minutes or longerEAL:CUI.2 Unusual EventRCS water level cannot be monitoredAND EITHER* UNPLANNED increase in any Table C-1 sump/tank level due to loss of RCSinventory* Visual observation of UNISOLABLE RCS leakageTable C-I Sumps / Tanks* Containment Sump 1* Containment Sump 2* Reactor Cavity Sump* CCW Surge Tank A* CCW Surge Tank B* PRT.RCDTMode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally.UNPLANNED-. A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient. The cause of the parameter changeor event may be known or unknown.CPNPP Basis:In Cold Shutdown mode, the RCS will normally be intact and standard RCS level monitoringmeans are available.In this EAL, all water level indication is unavailable and the RCS inventory loss must bedetected by indirect leakage indications (Table C-1). Level increases must be evaluatedagainst other potential sources of leakage such as cooling water sources inside thecontainment to ensure they are indicative of RCS leakage. If the make-up rate to the RCSunexplainably rises above the pre-established rate, a loss of RCS inventory may be occurringeven if the source of the leakage cannot be immediately identified. Visual observation ofPage 74 of 276 ATTACHMENT 1EAL Basesleakage from systems connected to the RCS that cannot be isolated could also be indicative ofa loss of RCS inventory (ref. 1, 2, 3, 4).NEI 99-01 Basis:This IC addresses the inability to restore and maintain water level to a required minimum level(or the lower limit of a level band), or a loss of the ability to monitor (FeaGter-veseRCS [PWR]or RP, rl3WRYlevel concurrent with indications of coolant leakage. Either of these conditionsis considered to be a potential degradation of the level of safety of the plant.Refueling evolutions that decrease RCS water inventory are carefully planned and controlled.An UNPLANNED event that results in water level decreasing below a procedurally requiredlimit warrants the declaration of an Unusual Event due to the reduced water inventory that isavailable to keep the core covered.E-=ALI #1 recognizes that the mnumrquired (reactor vessel/RCS [PWRq or RPV[914qR) level can change several times during the course of a refueling outage as difflerent.plant configurations and system lineups are implemented. This EAL is met if the minimumlevel, specified forF the- current plant conditions, cannot be maint-ained- for 15 minutes Or longer.-The minimum lev~,el is typically specified in the apibloperating procedure but ma" bespecified in another controlling document.The 15 minute threshold duration allows sufficient time8 for prompt operator actions to restoreand maintain the-expected water level. Thi-s criterion excludes transient conditinsausn abrief lowering of water level.This EAL-#-2 addresses a condition where all means to determine (reactor vessel!RCS [PWR9F RPV {BWRI.) level have been lost. In this condition, operators may determine that aninventory loss is occurring by observing changes in sump and/or tank levels (Table C-1).Sump and/or tank level changes must be evaluated against other potential sources of waterflow to ensure they are indicative of leakage from the (.eat.. .ess RCS R F R[BWR]).Continued loss of RCS inventory may result in escalation to the Alert emergency classificationlevel via either IC CAl or CA3.CPNPP Basis Reference(s):1. 1PO-01OA/B Reactor Coolant System Reduced Inventory Operations2. SOP-I1NB/1 Reactor Coolant System3. ABN-103 Excessive Reactor Coolant Leakage4. ABN-108 Shutdown Loss of Coolant5. NEI 99-01 CUIPage 75 of 276 ATTACHMENT 1EAL BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 1 -RCS LevelInitiating Condition: Loss of RCS inventoryEAL:CA1.1 AlertLoss of RCS inventory as indicated by RCS level less than 48 in. above upper core plate(top)Mode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):NoneCPNPP Basis:When Reactor Vessel water level decreases to 48 in. above the upper core plate (top) (EL 827'0"), RHR pump cavitation may occur. RCS level can be monitored by one or more of thefollowing (ref. 1, 2, 3, 4, 5, 6):* RCS Level Wide Range u-LI-3615B" RCS Level Narrow Range u-LI-3615A* RCS Extended Wide Range u-LI-3615C* Mansell Level Monitor System u-LT-3619AIB/C-1, -2" Plant Computer* RVLIS* Ultrasonic Level monitoring (optional)NEI 99-01 Basis:This IC addresses conditions that are precursors to a loss of the ability to adequately coolirradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This conditionrepresents a potential substantial reduction in the level of plant safety.For this EAL-#-I-, a lowering of RCS water level below 48 in. above the upper core plate (top)(site specific level) ft. indicates that operator actions have not been successful in restoring andmaintaining RCS (reactor vesse!IRCS [PWrIRI or RPV,4' water level. The heat-up rate ofthe coolant will increase as the available water inventory is reduced. A continuing decrease inwater level will lead to core uncovery.Although related, this EAL-#4 is concerned with the loss of RCS inventory and not the potentialconcurrent effects on systems needed for decay heat removal (e.g., loss of a Residul- DecayHeat Removal suction point). An increase in RGS-RCS temperature caused by a loss of decayheat removal capability is evaluated under IC CA3.Page 76 of 276 ATTACHMENT 1EAL BasesFor EAL #2, the inability to mon.itor (reactor V.SSel.R. S [PV.R] or RPV [B'A'R]) levelmay be caused by instrumentation and/orF power failures, Or water level dropping below therange of available instFrumentation. If water level c3-annot bhe monitored, operators m~aydetermne that an invento' los is occurri. by obsering c.hanges in sump and/or tank levels.Sump and/or tank level changes mus t be evaluated against other potenti-al sou1-rces of waterflow to ensure they are iniaieof leakage from the (reactor vcssel/RCS [PWVRI or RPVThe 15 mninute duration for the loss of level indication was chosen because it is half of the EALduration specified in IC CSIIf RCS the (reactor vcssel/RCS [PWVR] or RPV~ [BVVR]) inventory water level continues tolower, then escalation to Site Area Emergency would be via IC OSi.CPNPP Basis Reference(s):1 .IPO-Ol OA/B Reactor Coolant System Reduced Inventory Operations2. INC-6269 Calibration of the Mansell RCS Measurement System3. SOP-l0lA/B Reactor Coolant System4. ABN-1 03 Excessive Reactor Coolant Leakage5. ABN-104 Residual Heat Removal System Malfunction6. ABN-108 Shutdown Loss of Coolant7. NEI 99-01 CAlPage 77 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:EAL:C -Cold Shutdown / Refueling System Malfunction1 -RCS LevelLoss of RCS inventoryCA1.2 AlertRCS water level cannot be monitored for greater than or equal to 15 min. (Note 1)AND EITHER" UNPLANNED increase in any Table C-1 sump/tank level due to loss of RCSinventory" Visual observation of UNISOLABLE RCS leakageNote 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has beenexceeded, or will likely be exceeded.Table C-1 Sumps / Tanks* Containment Sump 1* Containment Sump 2* Reactor Cavity Sump* CCW Surge TankA* CCW Surge Tank B* PRT* RCDTMode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally.UNPLANNED-. A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient. The cause of the parameter changeor event may be known or unknown.CPNPP Basis:In Cold Shutdown mode, the RCS will normally be intact and standard RCS level monitoringmeans are available.In the Refuel mode, the RCS is not intact and RPV level may be monitored by different means,including the ability to monitor level visually.In this EAL, all RCS water level indication would be unavailable for greater than 15 minutes,and the RCS inventory loss must be detected by indirect leakage indications (Table C-1).Page 78 of 276 ATTACHMENT 1EAL BasesSurveillance procedures provide instructions for calculating primary system leak rate bymanual or computer-based water inventory balances. Level increases must be evaluatedagainst other potential sources of leakage such as cooling water sources inside thecontainment to ensure they are indicative of RCS leakage. If the make-up rate to the RCSunexplainably rises above the pre-established rate, a loss of RCS inventory may be occurringeven if the source of the leakage cannot be immediately identified. Visual observation ofleakage from systems connected to the RCS that cannot be isolated could also be indicative ofa loss of RCS inventory (ref. 1, 2, 3).NEI 99-01 Basis:This IC addresses conditions that are precursors to a loss of the ability to adequately coolirradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This conditionrepresents a potential substantial reduction in the level of plant safety.PFo EAL #1, a lowering of water level below (site specific level) indicates, that operatorFactions have not been successful in restGorig and maintainfin~g (reactor vcsscl!RCS [PL4'R oeRPV wate level. The heat up rate Of the coolant will increase as the available winventory is reduced. A continuing decrease in water leve! will lead to core uncover;.Alth.ugh related, #1 is concerned with the loss of RCS inventor.' and not thepotential concurrent effects On systems needed for decay heat removal (e.g., loss of aResidual Heat Removal suction point). An inrGease in ROS temperature caused by a loss ofdecay heat removal capability is evaluated under IC CA33.For this EAL-#2-, the inability to monitor RCS (reactor ..essel.RS [PWý eor RPV [BWR]) levelmay be caused by instrumentation and/or power failures, or water level dropping below therange of available instrumentation. If water level cannot be monitored, operators maydetermine that an inventory loss is occurring by observing changes in sump and/or tank levels.Sump and/or tank level changes must be evaluated against other potential sources of waterflow to ensure they are indicative of leakage from the (eaeter-vesSeRCS [PWR] or RPVThe 15-minute duration for the loss of level indication was chosen because it is half of the EALduration specified in IC CS1_If the (reaGt O.ves....RCS [PW.r or RPV [B\R]) inventory level continues to lower, thenescalation to Site Area Emergency would be via IC CS1.CPNPP Basis Reference(s):1. ABN-1 03 Excessive Reactor Coolant Leakage2. ABN-108 Shutdown Loss of Coolant3. FSAR 5.2.5.24. NEI 99-01 CA1Page 79 of 276 ATTACHMENT 1EAL BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 1 -RCS LevelInitiating Condition: Loss of RCS inventory affecting core decay heat removal capabilityEAL:CS1.1 Site Area EmergencyWith CONTAINMENT CLOSURE not established, RCS level less than 27.3 in. aboveupper core plate (top)Mode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):CONTAINMENT CLOSURE -The procedurally defined actions taken to secure containmentand its associated structures, systems, and components as a functional barrier to fissionproduct release under shutdown conditions. Containment closure means that all potentialescape paths are closed or capable of being closed:A. All penetrations providing direct access from Containment atmosphere to outsideatmosphere are closed except:" Penetrations with automatic valves capable of being closed by an operable CVI" Penetrations under administrative controls (e.g., Control Room notified and designatedperson to close if required by fuel handling accident)B. Equipment hatch is closed and held in place by 4 bolts, or is capable of being closed andheld in place by 4 boltsC. One emergency airlock door is closedD. One personnel airlock door is capable of being closedCPNPP Basis:When Reactor Vessel water level decreases to 27.25 in. (rounded to 27.3 in. for instrumentreadability), 825'-3 11/4" elevation (ref. 1), water level is six inches below the elevation of thebottom of the RCS hot leg penetration. When Reactor Vessel water level drops significantlybelow the elevation of the bottom of the RCS hot leg penetration, all sources of RCS injectionhave failed or are incapable of making up for the inventory loss. RCS elevations are illustratedin Figure C-3. RCS level can be monitored by one or more of the following (ref. 1, 2, 3):* RCS Level Wide Range u-LI-3615B* RCS Level Narrow Range u-LI-3615A" RCS Extended Wide Range u-LI-3615C* Mansell Level Monitor System u-LT-3619A/B/C-1, -2" Plant Computer" RVLIS* Ultrasonic Level monitoring (optional)Page 80 of 276 ATTACHMENT 1EAL BasesIn Refueling mode, Reactor Vessel water level indication from RVLIS is likely unavailable butalternate means of level indication are normally installed (including visual observation) toassure that the ability to monitor water level will not be interrupted.The status of Containment closure is tracked if plant conditions change that could raise the riskof a fission product release as a result of a loss of decay heat removal (ref. 4, 5).NEI 99-01 Basis:This IC addresses a significant and prolonged loss of (reactor vessel/RGS-RCS_[PWRe] r RPVrf^W])- inventory control and makeup capability leading to IMMINENT fuel damage. The lostinventory may be due to a RCS component failure, a loss of configuration control or prolongedboiling of reactor coolant. These conditions entail major failures of plant functions needed forprotection of the public and thus warrant a Site Area Emergency declaration.Following an extended loss of core decay heat removal and inventory makeup, decay heat willcause reactor coolant boiling and a further reduction in reactor vessel level. If RGS-eaeteivessel RCS level cannot be restored, fuel damage is probable.Outage/shutdown contingency plans typically provide for re-establishing or verifyingCONTAINMENT CLOSURE following a loss of heat removal or RCS inventory controlfunctions. The difference in the specified RCS/reactor vessel levels of EALs 4-1CS1.1 and2-bCS2.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lowerprobability of a fission product release to the environment.in EAL 3.a, the 30 minute criterion is tied to a readily reco~gnizable event start time (i.e., thetetal loss of ability to monitor level), and allows sGfficient time to monitoF, a2s6se -and c iorrelreactor and plant co~nditions to determnine if core Uncovery has actually occurred (i.e., toacco)unt for vaiu accident progression and instFrumentation uncertainties). it also allowssufficient time for pcfo~rmancc of actions to terminate leakage, recover inventor;control/makeup equipment and~or restore leVel monitoring-.The inability to monitor RCS (reactor vesscl/RGS [PWVR] or RPV [BWVR]) level may be causedby instrumentation and/or power failures, or water level dropping below the range of availab~ins~trumentation. if water level cannot be monitored, operators ma" determine thatainventor; loss is Gccurring by observing chane in sup and/or tank levels. Sump and/oitank level changes must be evaluated against other potential sources of water flow to ensurethey are indicativ~e of leakage from theRGS (reactor vesseliRCS [PVVARI or RP'. [B'NRI).These-ThisEALs addresses concerns raised by Generic Letter 88-17, Loss of Decay HeatRemoval; SECY 91 -283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449,Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the UnitedStates; and NUMARO 91-06, Guidelines for Industry Actions to Assess ShutdownManagement.Escalation of the emergency classification level would be via IC CGI or AG4RG1CPNPP Basis Reference(s):1. IPO-01 OA/B Reactor Coolant System Reduced Inventory Operations2. INC-6269 Calibration of the Mansell RCS Measurement System3. SOP-l0lA/B Reactor Coolant System4. Technical Specifications 3.9.4I Page 81 of 276 ATTACHMENT 1EAL Bases5. OPT-408A/B Refueling Containment Penetration Verification6. NEI 99-01 CS1Page 82 of 276 ATTACHMENT 1EAL BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 1 -RCS LevelInitiating Condition: Loss of RCS inventory affecting core decay heat removal capabilityEAL:CS1.2 Site Area EmergencyWith CONTAINMENT CLOSURE established, RCS level less than or equal to 0 in. aboveupper core plate (top)Mode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):CONTAINMENT CLOSURE -The procedurally defined actions taken to secure containmentand its associated structures, systems, and components as a functional barrier to fissionproduct release under shutdown conditions. Containment closure means that all potentialescape paths are closed or capable of being closed:A. All penetrations providing direct access from Containment atmosphere to outsideatmosphere are closed except:* Penetrations with automatic valves capable of being closed by an operable CVI" Penetrations under administrative controls (e.g., Control Room notified and designatedperson to close if required by fuel handling accident)B. Equipment hatch is closed and held in place by 4 bolts, or is capable of being closed andheld in place by 4 boltsC. One emergency airlock door is closedD. One personnel airlock door is capable of being closedCPNPP Basis:When Reactor Vessel water level drops below 0 in. above upper core plate (top) 823'-0"elevation (ref. 1), core uncovery is about to occur. RCS level can be monitored by one or moreof the following (ref. 1, 2, 3):" RCS Level Wide Range u-LI-3615B* RCS Level Narrow Range uj-LI-3615A" RCS Extended Wide Range u-LI-3615C* Mansell Level Monitor System u-LT-3619A/B/C-1, -2" Plant Computer* RVLIS* Ultrasonic Level monitoring (optional)Under the conditions specified by this EAL, continued lowering of Reactor Vessel water level isindicative of a loss of inventory control. Inventory loss may be due to a vessel breach, RCSpressure boundary leakage or continued boiling in the Reactor Vessel. The magnitude of thisPage 83 of 276 ATTACHMENT 1EAL Basesloss of water indicates that makeup systems have not been effective and may not be capableof preventing further RCS or Reactor Vessel water level drop and potential core uncovery. Theinability to restore and maintain level after reaching this setpoint infers a failure of the RCSbarrier and Potential Loss of the Fuel Clad barrier.The status of Containment closure is tracked if plant conditions change that could raise the riskof a fission product release as a result of a loss of decay heat removal (ref. 4, 5).NEI 99-01 Basis:This IC addresses a significant and prolonged loss of (reactor vessel/RS-&RCS_[PWR]-r RPV[-3,WRI}-inventory control and makeup capability leading to IMMINENT fuel damage. The lostinventory may be due to a RCS component failure, a loss of configuration control or prolongedboiling of reactor coolant. These conditions entail major failures of plant functions needed forprotection of the public and thus warrant a Site Area Emergency declaration.Following an extended loss of core decay heat removal and inventory makeup, decay heat willcause reactor coolant boiling and a further reduction in reactor vessel level. If RGS~eaetorYesse4RCS level cannot be restored, fuel damage is probable.Outage/shutdown contingency plans typically provide for re-establishing or verifyingCONTAINMENT CLOSURE following a loss of heat removal or RCS inventory controlfunctions. The difference in the specified RCS/reactor vessel levels of EALs bCS1.1 and2-.bCS1.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lowerprobability of a fission product release to the environment.1in EAL 3.a, the 30 m~inu-te c-riterion is tied to a readily recognizable event start time (i.e., thetotal less of ability to m onGitor level), and .....uffici, rntime to monitor, assess and corelatereoac-tor and plant conditions to determine if coro uncover,' has actually occurred (i.e., tosuffiient time for perfoermance of ac-tions; to terminate leakage, recover inventor!control/makeup equipment and/or restore level moenitoring-.The inability to mon~itor RCS (reactorF vesscl/RGS [PWVR] or RPV~ [BWR]) level may be causedby instrumentation and/or power failures, or water level dropping belo thag F availableinstrum~entation. if water level cannot be monRitored, operatorns may dote rmine that aninventor,' losi ocuring by obser~ing changes in sump and/or tank levels. Sump and/ortank level changes must be evaluated against other potential sources of water flow to ensurethey are indicative of leakage fromn the (reactor vessel/RCS [PVWR] or RPV [BWRJ).These-This EALs addresses concerns raised by Generic Letter 88-17, Loss of Decay HeatRemoval; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449,Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the UnitedStates; and NUMARC 91-06, Guidelines for Industry Actions to Assess ShutdownManagement.Escalation of the emergency classification level would be via IC CG1 or AG--RG1CPNPP Basis Reference(s):1. IPO-01OA/B Reactor Coolant System Reduced Inventory Operations2. INC-6269 Calibration of the Mansell RCS Measurement System3. SOP-101A/B Reactor Coolant SystemPage 84 of 276 1 ATTACHMENT 1EAL Bases4. Technical Specifications 3.9.45. OPT-408A/B Refueling Containment Penetration Verification6. NEI 99-01 CS1Page 85 of 276 1 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:EAL:C -Cold Shutdown / Refueling System Malfunction1 -RCS LevelLoss of RCS inventory affecting core decay heat removal capabilityCS1.3 Site Area EmergencyRCS water level cannot be monitored for greater than or equal to 30 min. (Note 1)ANDCore uncovery is indicated by any of the following:" UNPLANNED increase in any Table C-1 sump/tank level of sufficient magnitude toindicate core uncovery* Erratic Source Range Monitor indication* greater than 20,000 R/hr on any of the following:-CTEu16, Containment HRRM (u-RE-6290A)-CTWu17, Containment HRRM (u-RE-6290B)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has beenexceeded, or will likely be exceeded.Table C-1 Sumps / Tanks* Containment Sump 1* Containment Sump 2* Reactor Cavity Sump* CCW Surge TankA* CCW Surge Tank B* PRT* RCDTMode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally.UNPLANNED-. A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient. The cause of the parameter changeor event may be known or unknown.CPNPP Basis:In Cold Shutdown mode, the RCS will normally be intact and standard RCS level monitoringmeans are available.Page 86 of 276 ATTACHMENT IEAL BasesIn the Refueling mode, the RCS is not intact and RCS level may be monitored by differentmeans, including the ability to monitor level visually.In this EAL, all RCS water level indication would be unavailable for greater than 30 minutes,and the RCS inventory loss must be detected by indirect leakage indications (Table C-1).Surveillance procedures provide instructions for calculating primary system leak rate bymanual or computer-based water inventory balances. Level increases must be evaluatedagainst other potential sources of leakage such as cooling water sources inside thecontainment to ensure they are indicative of RCS leakage. If the make-up rate to the RCSunexplainably rises above the pre-established rate, a loss of RCS inventory may be occurringeven if the source of the leakage cannot be immediately identified (ref. 1, 2).The RCS inventory loss may be detected by the Containment High Range Radiation Monitor(HRRM) or erratic Source Range Monitor indication. As water level in the Reactor Vessellowers, the dose rate above the core will rise. The dose rate due to this core shine shouldresult in Containment High Range Radiation Monitor indication greater than 20,000 R/hr (ref.3). The Containment HRRMs have a range of IE-1 to 1E+8 R/hr.Post-TMI accident studies indicated that the installed PWR nuclear instrumentation will operateerratically when the core is uncovered and that this should be used as a tool for making suchdeterminations (ref. 4, 5).NEI 99-01 Basis:This IC addresses a significant and prolonged loss of (reater- ves, .RCSfPWR-,-RPV.)-inventory control and makeup capability leading to IMMINENT fuel damage. The lostinventory may be due to a RCS component failure, a loss of configuration control or prolongedboiling of reactor coolant. These conditions entail major failures of plant functions needed forprotection of the public and thus warrant a Site Area Emergency declaration.Following an extended loss of core decay heat removal and inventory makeup, decay heat willcause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactervessel level cannot be restored, fuel damage is probable.CONTAINMENT CLOSURE follGoI g a loss of heat removal or RCS , controlfunctionS. The diffeMrene in the specified RGS/reactor vessel levels of EALS 1 .b and 2.b reflectthe fact that with CONTAINIMlENIT CGLOSURE established, ther Is a lower pro.bability of afission product release to the envieroment.In EAL 3.aa-The 30-minute criterion is tied to a readily recognizable event start time (i.e., thetotal loss of ability to monitor level), and allows sufficient time to monitor, assess and correlatereactor and plant conditions to determine if core uncovery has actually occurred (i.e., toaccount for various accident progression and instrumentation uncertainties). It also allowssufficient time for performance of actions to terminate leakage, recover inventorycontrol/makeup equipment and/or restore level monitoring.The inability to monitor RCS (reactor vossel/ROS [PWR] or RPV [BDD 'R]) level may be causedby instrumentation and/or power failures, or water level dropping below the range of availableinstrumentation. If water level cannot be monitored, operators may determine that aninventory loss is occurring by observing changes in sump and/or tank levels. Sump and/orPage 87 of 276 ATTACHMENT IEAL Basestank level changes must be evaluated against other potential sources of water flow to ensurethey are indicative of leakage from the RCS (rca.te. v...e...lRCS [PWR] or RPV [BWVVR]).These-This EALs addresses concerns raised by Generic Letter 88-17, Loss of Decay HeatRemoval; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449,Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the UnitedStates; and NUMARC 91-06, Guidelines for Industry Actions to Assess ShutdownManagement.Escalation of the emergency classification level would be via IC CG1 or AG-1-RG1CPNPP Basis Reference(s):1. ABN-103 Excessive Reactor Coolant Leakage2. ABN-108 Shutdown Loss of Coolant3. Engineering Handbook, Guidelines for Events Beyond Design Basis: Spent Fuel Pools,Figure D "Dose Rate at Elevation 860' above Stored Fuel vs. Water Level Depth in SFP"4. Severe Accident Management Guidance Technical Basis Report, Volume 1: CandidateHigh-Level Actions and Their Effects, pgs 2-18, 2-195. Nuclear Safety Analysis Center (NSAC), 1980, "Analysis of Three Mile Island -Unit 2Accident," NSAC-16. NEI 99-01 CS1I Page 88 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:C -Cold Shutdown / Refueling System MalfunctionI -RCS LevelLoss of RCS inventory affecting fuel clad integrity with containmentchallengedEAL:CGI.1 General EmergencyRCS level less than or equal to 0 in. above upper core plate (top) for _> 30 min. (Note 1)ANDAny Containment Challenge indication, Table C-2Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declarationof a General Emergency is not required.Table C-2 Containment Challenge Indications* CONTAINMENT CLOSURE not established(Note 6)* Containment hydrogen concentration greaterthan 4%* Unplanned rise greater than 1 psig inContainment pressureMode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):CONTAINMENT CLOSURE -The procedurally defined actions taken to secure containmentand its associated structures, systems, and components as a functional barrier to fissionproduct release under shutdown conditions. Containment closure means that all potentialescape paths are closed or capable of being closed:A. All penetrations providing direct access from Containment atmosphere to outsideatmosphere are closed except:" Penetrations with automatic valves capable of being closed by an operable CVI" Penetrations under administrative controls (e.g., Control Room notified and designatedperson to close if required by fuel handling accident)B. Equipment hatch is closed and held in place by 4 bolts, or is capable of being closed andheld in place by 4 boltsC. One emergency airlock door is closedD. One personnel airlock door is capable of being closedPage 89 of 276 ATTACHMENT 1EAL BasesCPNPP Basis:When Reactor Vessel water level drops below 0 in. above upper core plate (top) 823'-0"elevation (ref. 1), core uncovery is about to occur. RCS level can be monitored by one or moreof the following (ref. 1, 2, 3):" RCS Level Wide Range u-LI-3615B" RCS Level Narrow Range u-LI-3615A* RCS Extended Wide Range u-LI-3615C* Mansell Level Monitor System u-LT-3619A/B/C-1, -2" Plant Computer* RVLIS* Ultrasonic Level monitoring (optional)Under the conditions specified by this EAL, continued lowering of Reactor Vessel water level isindicative of a loss of inventory control. Inventory loss may be due to a vessel breach, RCSpressure boundary leakage or continued boiling in the Reactor Vessel. The magnitude of thisloss of water indicates that makeup systems have not been effective and may not be capableof preventing further RCS or Reactor Vessel water level drop and potential core uncovery. Theinability to restore and maintain level after reaching this setpoint infers a failure of the RCSbarrier and Potential Loss of the Fuel Clad barrier.Three conditions are associated with a challenge to Containment integrity:1. CONTAINMENT COSURE not established -The status of Containment closure istracked if plant conditions change that could raise the risk of a fission product release asa result of a loss of decay heat removal (ref. 4, 5). If containment closure is re-established prior to exceeding the 30 minute core uncovery time limit then escalation toGE would not occur.2. Containment hydrogen greater than 4% -The 4% hydrogen concentration threshold isgenerally considered the lower limit for hydrogen deflagrations. CPNPP is equippedwith a Hydrogen Control System (HCS) which serves to limit or reduce combustible gasconcentrations in the Containment. The plant has two hydrogen monitoring systems.Each monitoring system consists of four sensor modules and one microprocessoranalyzer. Two sensors from each Containment are coupled to one of the two hydrogenmicroprocessors located in the Control Room. Thus each microprocessor analyzer isshared by Units 1 and 2. The analyzer system has a range of 0-10% hydrogen byvolume. The detector modules are located on the 905', 873', and 860' elevations inContainment. A fourth detector is located on 832' level across from the loop roomentrance for loops 1 and 4. Hydrogen concentration is displayed in the Control Room onu-AI-5506A/B and u-AI-5506C/D. Hydrogen concentration can also be displayed on thePlant Computer. Alarms at -3% are provided for high hydrogen concentration,u-ALB-3A, window 3.7. If a hydrogen concentration value can not be obtained from thehydrogen monitoring system, a grab sample from the containment PIG radiation monitormay be used to determine the hydrogen concentration (ref. 6, 7, 8, 9).3. UNPLANNED rise in Containment pressure -An unplanned pressure rise incontainment while in cold Shutdown or Refueling modes can threaten ContainmentPage 90 of 276 ATTACHMENT 1EAL BasesClosure capability and thus Containment potentially cannot be relied upon as a barrierto fission product release.NEI 99-01 Basis:This IC addresses the inability to restore and maintain reactor vessel level above the top ofactive fuel with containment challenged. This condition represents actual or IMMINENTsubstantial core degradation or melting with potential for loss of containment integrity.Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for morethan the immediate site area.Following an extended loss of core decay heat removal and inventory makeup, decay heat willcause reactor coolant boiling and a further reduction in reactor vessel level. If RCSRCS/reactor vessel level cannot be restored, fuel damage is probable.With CONTAINMENT CLOSURE not established, there is a high potential for a direct andunmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a GeneralEmergency is not required.The existence of an explosive mixture means, at a minimum, that the containment atmospherichydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagrationlimit). A hydrogen burn will raise containment pressure and could result in collateral equipmentdamage leading to a loss of containment integrity. It therefore represents a challenge toContainment integrity.In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a coreuncovery could result in an explosive gas mixture in containment. If all installed hydrogen gasmonitors are out-of-service during an event leading to fuel cladding damage, it may not bepossible to obtain a containment hydrogen gas concentration reading as ambient conditionswithin the containment will preclude personnel access. During periods when installedcontainment hydrogen gas monitors are out-of-service, operators may use the other listedindications to assess whether or not containment is challenged.In 30-minute criterion is tied to a readily recognizable event start time (i.e., thetotal loss of ability to monitor level), and allows sufficient time to monitor, assess and correlatereactor and plant conditions to determine if core uncovery has actually occurred (i.e., toaccount for various accident progression and instrumentation uncertainties). It also allowssufficient time for performance of actions to terminate leakage, recover inventorycontrol/makeup equipment and/or restore level monitoring.The inability ton monitor (reactor [PAqrI or RPV RCSJr'I'RJ) level may be causedby itrmn atinad/or power failures, or water level dropping below the range of availab-leinstrum~entation. if water level cannot be monflitored, operators ma" determine that aninventor,' loss is occurring by obseR'ing changsinsm and/or tank levels. Sump and/ottank level changes must be evaluatedagaistothetl I al sources of water floewto the" are indicativ~e of leakage from the (reactor ':ossel! RCS [PW4R] or RPV [BIA'R]).Thisese EALs addresses concerns raised by Generic Letter 88-17, Loss of Decay HeatRemoval; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449,Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the UnitedStates; and NUMARC 91-06, Guidelines for Industry Actions to Assess ShutdownPage 91 of 276 ATTACHMENT IEAL BasesManagement.IPage 92 of 276 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):1. IPO-010A/B Reactor Coolant System Reduced Inventory Operations2. INC-6269 Calibration of the Mansell RCS Measurement System3. SOP-101A/B Reactor Coolant System4. Technical Specifications 3.9.45. OPT-408A/B Refueling Containment Penetration Verification6. FRC-0.1A/B Response to Inadequate Core Cooling, Attachment 57. FSAR Section 6.2.58. FSAR Table 7.5-7A9. CHM-1 11 Primary Chemistry Accident Assessment Sampling Program10. NEI 99-01 CS1I Page 93 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:C -Cold Shutdown / Refueling System Malfunction1 -RCS LevelLoss of RCS inventory affecting fuel clad integrity with containmentchallengedEAL:CG1.2 General EmergencyRCS level cannot be monitored for greater than or equal to 30 min. (Note 1)ANDCore uncovery is indicated by any of the following:" UNPLANNED increase in any Table C-1 sump/tank level of sufficient magnitude toindicate core uncovery* Erratic Source Range Monitor indication" Greater than 20,000 R/hr on any of the following:-CTEu16, Containment HRRM (u-RE-6290A)-CTWu17, Containment HRRM (u-RE-6290B)ANDAny Containment Challenge indication, Table C-2Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration ofa General Emergency is not required.Table C-1 Sumps / Tanks* Containment Sump I* Containment Sump 2* Reactor Cavity Sump* CCW Surge TankA* CCW Surge Tank B* PRT* RCDTPage 94 of 276 ATTACHMENT 1EAL BasesTable C-2 Containment Challenge Indications* CONTAINMENT CLOSURE not established(Note 6)" Containment hydrogen concentration greaterthan 4%* Unplanned rise greater than 1 psig inContainment pressureMode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):CONTAINMENT CLOSURE -The procedurally defined actions taken to secure containmentand its associated structures, systems, and components as a functional barrier to fissionproduct release under shutdown conditions. Containment closure means that all potentialescape paths are closed or capable of being closed:A. All penetrations providing direct access from Containment atmosphere to outsideatmosphere are closed except:" Penetrations with automatic valves capable of being closed by an operable CVI" Penetrations under administrative controls (e.g., Control Room notified and designatedperson to close if required by fuel handling accident)B. Equipment hatch is closed and held in place by 4 bolts, or is capable of being closed andheld in place by 4 boltsC. One emergency airlock door is closedD. One personnel airlock door is capable of being closedUNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally.UNPLANNED-. A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient. The cause of the parameter changeor event may be known or unknown.CPNPP Basis:In Cold Shutdown mode, the RCS will normally be intact and standard RCS level monitoringmeans are available.In the Refueling mode, the RCS is not intact and RCS level may be monitored by differentmeans, including the ability to monitor level visually.In this EAL, all RCS water level indication would be unavailable for greater than 30 minutes,and the RCS inventory loss must be detected by indirect leakage indications (Table C-1).Surveillance procedures provide instructions for calculating primary system leak rate bymanual or computer-based water inventory balances. Level increases must be evaluatedagainst other potential sources of leakage such as cooling water sources inside thecontainment to ensure they are indicative of RCS leakage. If the make-up rate to the RCSPage 95 of 276 ATTACHMENT 1EAL Basesunexplainably rises above the pre-established rate, a loss of RCS inventory may be occurringeven if the source of the leakage cannot be immediately identified (ref. 1, 2).The RCS inventory loss may be detected by the Containment High Range Radiation Monitor(HRRM) or erratic Source Range Monitor indication. As water level in the Reactor Vessellowers, the dose rate above the core will rise. The dose rate due to this core shine shouldresult in Containment High Range Radiation Monitor indication greater than 20,000 R/hr (ref.3). The Containment HRRMs have a range of 1 E-1 to 1 E+8 R/hr.Post-TMI accident studies indicated that the installed PWR nuclear instrumentation will operateerratically when the core is uncovered and that this should be used as a tool for making suchdeterminations (ref. 4, 5).Three conditions are associated with a challenge to Containment integrity:1. CONTAINMENT COSURE not established -The status of Containment closure istracked if plant conditions change that could raise the risk of a fission product release asa result of a loss of decay heat removal (ref. 6, 7). If containment closure is re-established prior to exceeding the 30 minute core uncovery time limit then escalation toGE would not occur.2. Containment hydrogen greater than 4% -The 4% hydrogen concentration threshold isgenerally considered the lower limit for hydrogen deflagrations. CPNPP is equippedwith a Hydrogen Control System (HCS) which serves to limit or reduce combustible gasconcentrations in the Containment. The plant has two hydrogen monitoring systems.Each monitoring system consists of four sensor modules and one microprocessoranalyzer. Two sensors from each Containment are coupled to one of the two hydrogenmicroprocessors located in the Control Room. Thus each microprocessor analyzer isshared by Units 1 and 2. The analyzer system has a range of 0-10% hydrogen byvolume. The detector modules are located on the 905', 873', and 860' elevations inContainment. A fourth detector is located on 832' level across from the loop roomentrance for loops 1 and 4. Hydrogen concentration is displayed in the Control Room onu-AI-5506A/B and u-AI-5506C/D. Hydrogen concentration can also be displayed on thePlant Computer. Alarms at -3% are provided for high hydrogen concentration,u-ALB-3A, window 3.7. If a hydrogen concentration value can not be obtained from thehydrogen monitoring system, a grab sample from the containment PIG radiation monitormay be used to determine the hydrogen concentration (ref. 8, 9, 10, 11).3. UNPLANNED rise in Containment pressure -An unplanned pressure rise incontainment while in cold Shutdown or Refueling modes can threaten ContainmentClosure capability and thus Containment potentially cannot be relied upon as a barrierto fission product release.NEI 99-01 Basis:This IC addresses the inability to restore and maintain reactor vessel level above the top ofactive fuel with containment challenged. This condition represents actual or IMMINENTsubstantial core degradation or melting with potential for loss of containment integrity.Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for morethan the immediate site area.Following an extended loss of core decay heat removal and inventory makeup, decay heat willcause reactor coolant boiling and a further reduction in reactor vessel level. IfRCSPage 96 of 276 ATTACHMENT 1EAL Bases9RCS/Fractor Vosel level cannot be restored, fuel damage is probable.With CONTAINMENT CLOSURE not established, there is a high potential for a direct andunmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a GeneralEmergency is not required.The existence of an explosive mixture means, at a minimum, that the containment atmospherichydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagrationlimit). A hydrogen burn will raise containment pressure and could result in collateral equipmentdamage leading to a loss of containment integrity. It therefore represents a challenge toContainment integrity.In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a coreuncovery could result in an explosive gas mixture in containment. If all installed hydrogen gasmonitors are out-of-service during an event leading to fuel cladding damage, it may not bepossible to obtain a containment hydrogen gas concentration reading as ambient conditionswithin the containment will preclude personnel access. During periods when installedcontainment hydrogen gas monitors are out-of-service, operators may use the other listedindications to assess whether or not containment is challenged.in -=n-lA 2.bt-The 30-minute criterion is tied to a readily recognizable event start time (i.e., thetotal loss of ability to monitor level), and allows sufficient time to monitor, assess and correlatereactor and plant conditions to determine if core uncovery has actually occurred (i.e., toaccount for various accident progression and instrumentation uncertainties). It also allowssufficient time for performance of actions to terminate leakage, recover inventorycontrol/makeup equipment and/or restore level monitoring.The inability to monitor (reactor ..esscl,.RGS [Pr4'R] or RPV-RCS.BrK,) level may be causedby instrumentation and/or power failures, or water level dropping below the range of availableinstrumentation. If water level cannot be monitored, operators may determine that aninventory loss is occurring by observing changes in sump and/or tank levels. Sump and/ortank level changes must be evaluated against other potential sources of water flow to ensurethey are indicative of leakage from the (reaGte .vesse./RNCS [P-K/ or RPV [BI\'R,).Thisese EALs addresses concerns raised by Generic Letter 88-17, Loss of Decay HeatRemoval; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1 449,Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the UnitedStates; and NUMARC 91-06, Guidelines for Industry Actions to Assess ShutdownManagement.Page 97 of 276 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):1. ABN-1 03 Excessive Reactor Coolant Leakage2. ABN-108 Shutdown Loss of Coolant3. Engineering Handbook, Guidelines for Events Beyond Design Basis: Spent Fuel Pools,Figure D "Dose Rate at Elevation 860' above Stored Fuel vs. Water Level Depth in SFP"4. Severe Accident Management Guidance Technical Basis Report, Volume 1: CandidateHigh-Level Actions and Their Effects, pgs 2-18, 2-195. Nuclear Safety Analysis Center (NSAC), 1980, "Analysis of Three Mile Island -Unit 2Accident," NSAC-16. Technical Specifications 3.9.47. OPT-408A/B Refueling Containment Penetration Verification8. FRC-O.1A/B Response to Inadequate Core Cooling, Attachment 59. FSAR Section 6.2.510. FSAR Table 7.5-7A11. CHM-1 II Primary Chemistry Accident Assessment Sampling Program12. NEI 99-01 CG1Page 98 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:C -Cold Shutdown / Refueling System Malfunction2 -Loss of Emergency AC PowerLoss of all but one AC power source to safeguard buses for 15minutes or longerEAL:CU2.1 Unusual EventAC power capability, Table C-3, to 6.9 KV safeguard buses uEA1 and uEA2 reduced to asingle power source for greater than or equal to 15 min. (Note 1)ANDAny additional single Table C-3 power source failure will result in loss of all AC power toSAFETY SYSTEMSNote 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Table C-3 AC Power SourcesOffsite:* 138 KV switchyard circuit* 345 KV switchyard circuitOnsite:" uEG1* uEG2Mode Applicability:5 -Cold Shutdown, 6- Refueling, D -DefueledDefinition(s):SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/orplacing it in the cold shutdown condition, including the ECCS. These are typically systemsclassified as safety-related (as defined in 10CFR50.2):Those structures, systems and components that are relied upon to remain functional duringand 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 could resultin potential offsite exposures.Page 99 of 276 ATTACHMENT 1EAL BasesCPNPP Basis:For emergency classification purposes, "capability" means that an offsite AC power source(s)is available to the emergency buses, whether or not the buses are powered from it.The condition indicated by this EAL is the degradation of the offsite and onsite power sourcessuch that any additional single failure would result in a loss of all AC power to the emergencybuses.The safeguards AC distribution system power sources consist of the preferred and alternateoffsite power sources, and the onsite standby emergency diesel generators uEGI and uEG2.Offsite power is supplied to the plant switchyards from the transmission network by five 345 KVand two 138 KV transmission lines. From the switchyards, two electrically and physicallyseparated circuits provide AC power through step down startup transformers, to the 6.9 kVsafeguard buses. The 138 kV switchyard circuit is the preferred source for Unit 2 and alternatesource for Unit 1. The 345 KV circuit is the preferred source for Unit 1 and alternate source forUnit 2. The onsite AC distribution system is divided into redundant trains so that the loss of anyone load group does not prevent the minimum safety functions from being performed. Eachtrain has connections to two offsite power sources and a dedicated diesel generator. Eachoffsite circuit can supply the Unit 1 and Unit 2 6.9 KV safeguard buses. (ref. 1, 2, 3, 4).This cold condition EAL is equivalent to the hot condition EAL SAI.1.NEI 99-01 Basis:This IC describes a significant degradation of offsite and onsite AC power sources such thatany additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. Inthis condition, the sole AC power source may be powering one, or more than one, train ofsafety-related equipment.When in the cold shutdown, refueling, or defueled mode, this condition is not classified as anAlert because of the increased time available to restore another power source to service.Additional time is available due to the reduced core decay heat load, and the lowertemperatures and pressures in various plant systems. Thus, when in these modes, thiscondition is considered to be a potential degradation of the level of safety of the plant.An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplyingrequired power to an essential bus. Some examples of this condition are presented below.* A loss of all offsite power with a concurrent failure of all but one emergency powersource (e.g., an onsite diesel generator)." A loss of all offsite power and loss of all emergency power sources (e.g., onsite dieselgenerators) with a single train of emergency buses being back-fed from the unit maingenerator.* A loss of emergency power sources (e.g., onsite diesel generators) with a single train ofemergency buses being back-fed from an offsite power source.Fifteen minutes was selected as a threshold to exclude transient or momentary losses ofpower.The subsequent loss of the remaining single power source would escalate the event to an Alertin accordance with IC CA2.Page 100 of 276 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):1.2.3.4.5.6.7.FSAR Figure 8.3-1FSAR Section 8.2FSAR Section 8.3Technical Specifications B3.8.1ABN-601 Response to a 138/345 KV System MalfunctionABN-602 Response to a 6900/480V System MalfunctionNEI 99-01 CU2Page 101 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:C -Cold Shutdown / Refueling System Malfunction2 -Loss of Emergency AC PowerLoss of all offsite and all onsite AC power to safeguard buses for 15minutes or longerEAL:CA2.1 AlertLoss of all offsite and all onsite AC power capability, Table C-3, to 6.9 KV safeguardbuses uEA1 and uEA2 for greater than or equal to 15 min. (Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has beenexceeded, or will likely be exceeded.Table C-3 AC Power SourcesOffsite:* 138 KV switchyard circuit* 345 KV switchyard circuitOnsite:" uEG1* uEG2Mode Applicability:5 -Cold Shutdown, 6 -Refueling, D -DefueledCPNPP Basis:For emergency classification purposes, "capability" means that an offsite AC power source(s)is available to the emergency buses, whether or not the buses are powered from it.The safeguards AC distribution system power sources consist of the preferred and alternateoffsite power sources, and the onsite standby emergency diesel generators uEGI and uEG2.Offsite power is supplied to the plant switchyards from the transmission network by five 345 KVand two 138 KV transmission lines. From the switchyards, two electrically and physicallyseparated circuits provide AC power through step down startup transformers, to the 6.9 kVsafeguard buses. The 138 kV switchyard circuit is the preferred source for Unit 2 and alternatesource for Unit 1. The 345 KV circuit is the preferred source for Unit I and alternate source forUnit 2. The onsite AC distribution system is divided into redundant trains so that the loss of anyone load group does not prevent the minimum safety functions from being performed. Eachtrain has connections to two offsite power sources and a dedicated diesel generator. Eachoffsite circuit can supply the Unit 1 and Unit 2 6.9 KV safeguard buses. (ref. 1, 2, 3, 4)This cold condition EAL is equivalent to the hot condition loss of all offsite AC power EALSS1.1.I Page 102 of 276 1 ATTACHMENT 1EAL BasesNEI 99-01 Basis:This IC addresses a total loss of AC power that compromises the performance of all SAFETYSYSTEMS requiring electric power including those necessary for emergency core cooling,containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.When in the cold shutdown, refueling, or defueled mode, this condition is not classified as aSite Area Emergency because of the increased time available to restore an emergency bus toservice. Additional time is available due to the reduced core decay heat load, and the lowertemperatures and pressures in various plant systems. Thus, when in these modes, thiscondition represents an actual or potential substantial degradation of the level of safety of theplant.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Escalation of the emergency classification level would be via IC CS1 or AS4-RS1.CPNPP Basis Reference(s):1. FSAR Figure 8.3-12. FSAR Section 8.23. FSAR Section 8.34. Technical Specifications B3.8.15. ABN-601 Response to a 138/345 KV System Malfunction6. ABN-602 Response to a 6900/480V System Malfunction7. NEI 99-01 CA2Page 103 of 276 1 ATTACHMENT 1EAL BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 3 -RCS TemperatureInitiating Condition: UNPLANNED increase in RCS temperatureEAL:CU3.1 Unusual EventUNPLANNED increase in RCS temperature to greater than 200OF due to loss of decayheat removal capability (Note 9)Note 9: Begin monitoring hot condition EALs concurrently.Mode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):UNPLANNED-. A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient. The cause of the parameter changeor event may be known or unknown.CPNPP Basis:Several instruments are capable of providing indication of RCS temperature with respect to theTechnical Specification cold shutdown temperature limit (200'F, ref. 1). These include loop Thot(u-TR-413A/23A, u -TR-433A/43A, u -TI-413A, u -TI-423A) and, if no RCPs are operating, theCore Exit Thermocouples (TCs). The most limiting temperature indication should be used. Forexample, during heatup, the highest reading temperature indication should be used; duringcooldown, the lowest (ref. 2, 3, 4, 5).In the absence of reliable RCS temperature indication caused by the loss of decay heatremoval capability, classification should be based on the RCS pressure increase criteria whenthe RCS is intact in Mode 5 or based on time to boil data when in Mode 6 or the RCS is notintact in Mode 5.The note is a reminder that any temperature increase above 2001F is an operating modechange from cold to hot conditions. Since each EAL is associated with operating modeapplicability, the set of EALs that must be monitored must now include EALs associated withhot condition operating modes.NEI 99-01 Basis:This IC addresses an UNPLANNED increase in RCS temperature above the TechnicalSpecification cold shutdown temperature limit.,or the inability to determinc RCS t.mpc.atur.aadlevel-and represents a potential degradation of the level of safety of the plant. If the RGSRCS-is not intact and CONTAINMENT CLOSURE is not established during this event, theEmergency Director Emerqency Coordinator should also refer to IC CA3.A momentary UNPLANNED excursion above the Technical Specification cold shutdowntemperature limit when the heat removal function is available does not warrant a classification.Page 104 of 276 ATTACHMENT 1EAL BasesI EAL-#4This EAL involves a loss of decay heat removal capability, or an addition of heat to theRCS in excess of that which can currently be removed, such that reactor coolant temperaturecannot be maintained below the cold shutdown temperature limit specified in TechnicalSpecifications. During this condition, there is no immediate threat of fuel damage because thecore decay heat load has been reduced since the cessation of power operation.During an outage, the level in the reactor vessel will normally be maintained at or above thereactor vessel flange. Refueling evolutions that lower water level below the reactor vesselflange are carefully planned and controlled. A loss of forced decay heat removal at reducedinventory may result in a rapid increase in reactor coolant temperature depending on the timeafter shutdown.EAL #2 reflects a onldition whore there has been a Significant loss orf instrmentationcapability ecessary t on moritor RCS pcoditions and operators would be unable to moaitor keypa1.reters neessarp' to assure core decay heat removal. During this condition, there is noMimmednate threat of fuel damnage beause the core decay heat load has been reduced incethe cessation of power operatien.Fifteen mninutes was selected as a threshold to exclude tr-anSient or moementar,' loesof IRdeation-Escalation to Alert would be via IC CAl based on an inventory loss or IC CA3 based onexceeding plant configuration-specific time criteria.CPNPP Basis Reference(s):1 .Technical Specifications Table 1.1-12. IPO-OO5N/B Plant Cooldown From Hot Standby To Cold Shutdown3. Technical Specifications 3.4.34. OPT-407 RCS Pressure and Temperature Verification5. IPO-01 OA/B Reactor Coolant System Reduced Inventory Condition6. NEI 99-01 CU3Page 105 of 276 1 ATTACHMENT 1EAL BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 3 -RCS TemperatureInitiating Condition: UNPLANNED increase in RCS temperatureEAL:CU3.2 Unusual EventLoss of all RCS temperature and RCS level indication for greater than or equal to 15 min.(Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Mode Applicability:5 -Cold Shutdown, 6- RefuelingDefinition(s):NoneCPNPP Basis:RCS level can be monitored by one or more of the following (ref. 2, 3, 4):" RCS Level Wide Range u-LI-3615B* RCS Level Narrow Range u-LI-3615A* RCS Extended Wide Range u-LI-3615C* Mansell Level Monitor System u-LT-3619A/B/C-1, -2* Plant Computer* RVLIS* Ultrasonic Level monitoring (optional)Several instruments are capable of providing indication of RCS temperature with respect to theTechnical Specification cold shutdown temperature limit (200'F, ref. 1). These include loop Thot(u-TR-413A/23A, u -TR-433A/43A, u -TI-413A, u -TI-423A) and, if no RCPs are operating, theCore Exit Thermocouples (TCs). The most limiting temperature indication should be used. Forexample, during heatup, the highest reading temperature indication should be used; duringcooldown, the lowest (ref. 5, 6, 7, 8).In the absence of reliable RCS temperature indication caused by the loss of decay heatremoval capability, classification should be based on heat up rate data or additionally if inMode 5 with RCS intact on pressure increase.NEI 99-01 Basis:This IG-EAL addresses an UNPLANNED i en RS temperature abo'... the TechniGalSpecific3tio c.ld. shutdown temperature limit, or the inability to determine RCS temperaturePage 106 of 276 ATTACHMENT 1EAL Basesand level, and represents a potential degradation of the level of safety of the plant. If the RCSis not intact and CONTAINMENT CLOSURE is not established during this event, theEmergency DiroctorEmerqency Coordinator should also refer to IC CA3.A tar,' UNPLANNED ,Xcur"sin above the Rpecification cold shutd-n,+ nnr 4..Ir i n-,. *.,kn 4 k kn 4- .-rn ,-,I.v 4n4.a H-. ,-I.k1 AlC +in i M, +;4E=AL #1 involves a loss of decay heat remoeval capability, or an addition of heat to theRCS in exGess ef that Which can currntly b8 rFemroved, such tha-t coolart temperatuFecannot be mnaintained below the coldG sh-utdown40 temperature limit specified in Technic~alSpecifications. DUFiRg this conditinR, there is no immediate threat of fuel damage because thecore decay heat lead has been reduced s the cessatio of power opeFatinr.Duarig an outage, the levol in the reactor ve-s-sel will nrmally be Maintained above thereactor vessel flange. Refueling evoluitions that lower water level below the reactor vesselflange are carefully planned and controlled. A loss- o-f forc-ed decay heat Frnemval at reducedinventei' may result in a rapid increase ireactor coolant temperature depending en the timeafter shutdown.EAl= #2This EAL reflects a condition where there has been a significant loss of instrumentationcapability necessary to monitor RCS conditions and operators would be unable to monitor keyparameters necessary to assure core decay heat removal. During this condition, there is noimmediate threat of fuel damage because the core decay heat load has been reduced sincethe cessation of power operation.Fifteen minutes was selected as a threshold to exclude transient or momentary losses ofindication.Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based onexceeding plant configuration-specific time criteria.CPNPP. Basis Reference(s):1. Technical Specifications Table 1 .1-12. IPO-01OA/B Reactor Coolant System Reduced Inventory Operations3. INC-6269 Calibration of the Mansell RCS Measurement System4. SOP-101A/B Reactor Coolant System5. IPO-005A/B Plant Cooldown From Hot Standby To Cold Shutdown6. Technical Specifications 3.4.37. OPT-407 RCS Pressure and Temperature Verification8. IPO-01OA/B Reactor Coolant System Reduced Inventory Condition9. NEI 99-01 CU3Page 107 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:C -Cold Shutdown / Refueling System Malfunction3 -RCS TemperatureInitiating Condition: Inability to maintain plant in cold shutdownEAL:CA3.1 AlertUNPLANNED increase in RCS temperature to greater than 200OF for greater than TableC-4 duration(Notes 1, 9)ORUNPLANNED RCS pressure increase greater than 10 psig due to a loss of RCS cooling(This EAL does not apply during water-solid plant conditions)Note 1: The Emergency Coordinator should declare the event promptly upon determining that the applicabletime has been exceeded, or will likely be exceeded.Note 9: Begin monitoring hot condition EALs concurrently.Table C-4: RCS Heat-up Duration ThresholdsCONTAINMENTRCS Status CLOSURE Status Heat-up DurationIntact (but notREDUCED N/A 60 min.*INVENTORY)Not intact Established 20 min.*ORREDUCED INVENTORY Not established 0 min.* If an RCS heat removal system is in operation within this time frame and RCS temperature isbeing reduced, the EAL is not applicable.Mode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):CONTAINMENT CLOSURE -The procedurally defined actions taken to secure containmentand its associated structures, systems, and components as a functional barrier to fissionproduct release under shutdown conditions. Containment closure means that all potentialescape paths are closed or capable of being closed:A. All penetrations providing direct access from Containment atmosphere to outsideatmosphere are closed except:" Penetrations with automatic valves capable of being closed by an operable CVI" Penetrations under administrative controls (e.g., Control Room notified and designatedperson to close if required by fuel handling accident)Page 108 of 276 ATTACHMENT 1EAL BasesB. Equipment hatch is closed and held in place by 4 bolts, or is capable of being closed andheld in place by 4 boltsC. One emergency airlock door is closedD. One personnel airlock door is capable of being closedUNPLANNED -. A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient. The cause of the parameter changeor event may be known or unknown.REDUCED INVENTORY -Plant condition when fuel is in the reactor vessel and ReactorCoolant System level is < 80 inches above core plate (829'8").CPNPP Basis:Several instruments are capable of providing indication of RCS temperature with respect to theTechnical Specification cold shutdown temperature limit (200'F, ref. 1). These include loop Thot(u-TR-413A/23A, u -TR-433A/43A, u -TI-413A, u -TI-423A) and, if no RCPs are operating, theCore Exit Thermocouples (TCs). The most limiting temperature indication should be used. Forexample, during heatup, the highest reading temperature indication should be used; duringcooldown, the lowest (ref. 2, 3, 4, 5).In the absence of reliable RCS temperature indication caused by the loss of decay heatremoval capability, classification should be based on heat up rate data or additionally if inMode 5 with RCS intact on pressure increase.A 10 psig RCS pressure increase can be monitored on u-PI-403A and computer pointsP6498A and P6499A (ref. 9, 10).The status of Containment closure is tracked if plant conditions change that could raise the riskof a fission product release as a result of a loss of decay heat removal (ref. 6, 7).The note is a reminder that any temperature increase above 200OF is an operating modechange from cold to hot conditions. Since each EAL is associated with operating modeapplicability, the set of EALs that must be monitored must now include EALs associated withhot condition operating modes.NEI 99-01 Basis:This IC addresses conditions involving a loss of decay heat removal capability or an addition ofheat to the RCS in excess of that which can currently be removed. Either condition representsan actual or potential substantial degradation of the level of safety of the plant.A momentary UNPLANNED excursion above the Technical Specification cold shutdowntemperature limit when the heat removal function is available does not warrant a classification.The RGS-RCSHeat-up Duration Thresholds table addresses an increase in RCS temperaturewhen CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS inventory isreduced (e.g., mid-loop operation i-R PWRs). The 20-minute criterion was included to allowtime for operator action to address the temperature increase.The RCS Heat-up Duration Thresholds table also addresses an increase in RCS temperaturewith the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this conditionsince the intact RCS is providing a high pressure barrier to a fission product release. The 60-Page 109 of 276 ATTACHMENT 1EAL Basesminute time frame should allow sufficient time to address the temperature increase without asubstantial degradation in plant safety.Finally, in the case where there is an increase in RCS temperature, the RCS is not intact or isat reduced inventory-fP-WR4, and CONTAINMENT CLOSURE is not established, no heat-upduration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may bereleased directly into the containment atmosphere and subsequently to the environment, and2) there is reduced reactor coolant inventory above the top of irradiated fuel.EAL4#2The RCS pressure increase threshold provides a pressure-based indication of RCSheat-up in the absence of RCS temperature monitoring capability.Escalation of the emergency classification level would be via IC CS1 or A-S4RS1.CPNPP Basis Reference(s):1. Technical Specifications Table 1.1-12. IPO-005A/B Plant Cooldown From Hot Standby To Cold Shutdown3. Technical Specifications 3.4.34. OPT-407 RCS Pressure and Temperature Verification5. IPO-01OA/B Reactor Coolant System Reduced Inventory Condition6. Technical Specifications 3.9.47. OPT-408A/B Refueling Containment Penetration Verification8. IPO-010A/B Reactor Coolant System Reduced Inventory Operations9. IPO-005AIB Plant Cooldown From Hot Standby To Cold Shutdown10. SOP-101A/B Reactor Coolant System Reduced Inventory11. NEI 99-01 CA3I Page 110 of 276 ATTACHMENT 1EAL BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 4 -Loss of Vital DC PowerInitiating Condition: Loss of vital DC power for 15 minutes or longerEAL:CU4.1 Unusual EventLess than 105 VDC bus voltage indications on Technical Specification required 125 VDCbuses (uED1, uED2, uED3, uED4) for greater than or equal to 15 min. (Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Mode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):NoneCPNPP Basis:The purpose of this EAL is to recognize a loss of DC power compromising the ability to monitorand control the removal of decay heat during cold shutdown or refueling operations. This EALis intended to be anticipatory in as much as the operating crew may not have necessaryindication and control of equipment needed to respond to the loss. The fifteen minute intervalis intended to exclude transient or momentary power losses.The safeguards 125 VDC buses are the Class 1E buses uED1, uED2, uED3 and uED4 (ref. 1).The 125 VDC safeguard distribution system is illustrated in Figure C-2 (ref. 2, 3).Each redundant safeguards 125 VDC system consists of two independent batteries eachhaving one main distribution bus, two static battery chargers (one spare), and local distributionpanels. For Unit 1, batteries BT1 ED1 and BT1 ED3 feed all train A load requirements, whilebatteries BT1 ED2 and BT1 ED4 supply train B load requirements.For Unit 2, batteries BT2ED1 and BT2ED3 feed all train A load requirements, while batteriesBT2ED2 and BT2ED4 supply train B load requirements. There are no bus ties or sharing ofpower supplies between redundant trains (ref. 1).Minimum DC bus voltage is 105 VDC (ref. 4). Bus voltage may be monitored from the followingindications (ref. 6):Control Room Panel CP-10 Annunciator u--ALB-10B Plant ComputerV-1ED1, 125VDC SWITCH PNL 1ED1 VOLT 1.13 V6501A BATT BT1ED1 VOLTV-1 ED2, 125VDC SWITCH PNL 1ED2 VOLT 2.13 V6502A BATT BT1ED2 VOLTV-1ED3, 125VDC SWITCH PNL 1ED3 VOLT 1.9 V6503A BATT BT1ED3 VOLTV-1 ED4, 125VDC SWITCH PNL 1ED4 VOLT 3.9 V6504A BATT BT1ED4 VOLTThis EAL is the cold condition equivalent of the hot condition loss of DC power EAL SS7.1.Page 111 of 276 ATTACHMENT 1EAL BasesNEI 99-01 BasisThis IC addresses a loss of vital DC power which compromises the ability to monitor andcontrol operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode.In these modes, the core decay heat load has been significantly reduced, and coolant systemtemperatures and pressures are lower; these conditions increase the time available to restorea vital DC bus to service. Thus, this condition is considered to be a potential degradation ofthe level of safety of the plant.As used in this EAL, "required" means the vital DC buses necessary to support operation ofthe in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, ifTrain A is out-of-service (inoperable) for scheduled outage maintenance work and Train B isin-service (operable), then a loss of Vital DC power affecting Train B would require thedeclaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant anemergency classification.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Depending upon the event, escalation of the emergency classification level would be via ICCA1 or CA3, or an IC in Recognition Category AR.CPNPP Basis Reference(s):1. FSAR 8.3.22. FSAR Figure 8.3-143. FSAR Figure 8.3-14A4. ECA-O.OA/B Loss of All AC Power5. SOP-605ANB 125 VDC Switchgear and Distribution Systems, Batteries and BatteryChargers6. ALM-0102A/B Alarm Procedures Manual, u-ALB-1OB, nos. 1.9, 1.13, 2.13, 3.87. NEI 99-01 CU4Page 112 of 276 ATTACHMENT IEAL BasesCategory:Subcategory:Initiating Condition:EAL:C -Cold Shutdown / Refueling System Malfunction5 -Loss of CommunicationsLoss of all onsite or offsjte communications capabilitiesCU5.1 Unusual EventLoss of all Table C-5 onsite communication methodsORLoss of all Table C-5 offsite communication methodsORLoss of all Table C-5 NRC communication methodsTable C-5 Communication MethodsSystem Onsite Offsite NRCGai-tronics Page/Party (PA) XPlant Radios XPABX X X XPublic Telephone X X XFederal Telephone System (FTS) X XMode Applicability:5 -Cold Shutdown, 6 -Refueling, D -DefueledDefinition(s):NoneCPNPP Basis:Onsite/offsite communications include one or more of the systems listed in Table C-5 (ref. 1,2).This EAL is the cold condition equivalent of the hot condition EAL SU7.1.NEI 99-01 Basis:This IC addresses a significant loss of on-site or offsite communications capabilities. While nota direct challenge to plant or personnel safety, this event warrants prompt notifications toOROs and the NRC.Page 113 of 276 ATTACHMENT 1EAL BasesThis IC should be assessed only when extraordinary means are being utilized to makecommunications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent tooffsite locations, etc.).I EAL-#4The first EAL condition addresses a total loss of the communications methods used insupport of routine plant operations.EAL #2The second EAL condition addresses a total loss of the communications methods usedto notify all OROs of an emergency declaration. The offsite (OROs. referred to here are-(seeDevelGpe.-N.tee) the State Department of Public Safety, Somervell and Hood County EOCs7EAL4#3The third EAL addresses a total loss of the communications methods used to notify theNRC of an emergency declaration.CPNPP Basis Reference(s):1. FSAR 9.5.22. DBD-EE-048 Communication System3. NEI 99-01 CU5I Page 114 of 276 1 ATTACHMENT IEAL BasesCategory:Subcategory:Initiating Condition:C -Cold Shutdown / Refueling System Malfunction6 -Hazardous Event Affecting Safety SystemsHazardous event affecting a SAFETY SYSTEM needed for the currentoperating modeEAL:CA6.1 AlertThe occurrence of any Table C-6 hazardous eventAND EITHER:" Event damage has caused indications of degraded performance in at least one trainof a SAFETY SYSTEM needed for the current operating mode* The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component orstructure needed for the current operating modeTable C-6 Hazardous Events* Seismic event (earthquake)* Internal or external FLOODING event* High winds or tornado strike" FIRE* EXPLOSION" Other events with similar hazard characteristicsas determined by the Emergency CoordinatorMode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):EXPLOSION -A rapid, violent and catastrophic failure of a piece of equipment due tocombustion, chemical reaction or overpressurization. A release of steam (from high energylines or components) or an electrical component failure (caused by short circuits, grounding,arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drivebelts or overheated electrical equipment do not constitute fires. Observation of flame ispreferred but is NOT required if large quantities of smoke and heat are observed.Page 115 of 276 ATTACHMENT 1EAL BasesFLOODING -A condition where water is entering a room or area faster than installedequipment is capable of removal, resulting in a rise of water level within the room or area.SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/orplacing it in the cold shutdown condition, including the ECCS. These are typically systemsclassified as safety-related (as defined in 10CFR50.2):Those structures, systems and components that are relied upon to remain functional duringand 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 could resultin potential offsite exposures.VISIBLE DAMAGE -Damage to a component or structure that is readily observable withoutmeasurements, testing, or analysis. The visual impact of the damage is sufficient to causeconcern regarding the operability or reliability of the affected component or structure.CPNPP Basis:" The significance of seismic events are discussed under EAL HU2.1 (ref. 1).* Internal FLOODING may be caused by events such as component failures, equipmentmisalignment, or outage activity mishaps (ref. 2).* External flooding may be due to high lake level (ref. 3, 4).* Seismic Category I structures are analyzed to withstand a sustained, design wind velocityof at least 80 mph. (ref. 5).* Areas containing functions and systems required for safe shutdown of the plant areidentified by fire area (ref. 6, 7).* An explosion that degrades the performance of a SAFETY SYSTEM train or visiblydamages a SAFETY SYSTEM component or structure would be classified under this EAL.NEI 99-01 Basis:This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or astructure containing SAFETY SYSTEM components, needed for the current operating mode.This condition significantly reduces the margin to a loss or potential loss of a fission productbarrier, and therefore represents an actual or potential substantial degradation of the level ofsafety of the plant.EAL 4.b.'The first conditional addresses damage to a SAFETY SYSTEM train that is inservice/operation since indications for it will be readily available. The indications of degradedperformance should be significant enough to cause concern regarding the operability orreliability of the SAFETY SYSTEM train.Page 116 of 276 ATTACHMENT 1EAL BasesEAL l1b.2The second conditional addresses damage to a SAFETY SYSTEM component thatis not in service/operation or readily apparent through indications alone, or to a structurecontaining SAFETY SYSTEM components. Operators will make this determination based onthe totality of available event and damage report information. This is intended to be a briefassessment not requiring lengthy analysis or quantification of the damage.Escalation of the emergency classification level would be via IC CS1 or A-S--RSI.CPNPP Basis Reference(s):1. ABN-907 Acts of Nature2. CPNPP PRA Accident Sequence Analysis "Internal Flooding Sequences"3. FSAR Section 2.4.3.7 Flood Evaluations for Safe Shutdown Impoundment4. DBD-CS-071 Maximum Probable Flood5. FSAR Section 3.3.1.1 Wind Loadings6. CPNPP Fire Protection Report, Section 5.0 "Fire Safe Shutdown Equipment List"7. FSAR Section 7.4 Systems Required for Safe Shutdown8. NEI 99-01 CA6Page 117 of 276 ATTACHMENT 1EAL BasesCategory H -Hazards and Other Conditions Affecting Plant SafetyEAL Group: ANY (EALs in this category are applicable to any plantcondition, hot or cold.)Hazards are non-plant, system-related events that can directly or indirectly affect plantoperation, reactor plant safety or personnel safety.1. SecurityUnauthorized entry attempts into the Protected Area, bomb threats, sabotage attempts, andactual security compromises threatening loss of physical control of the plant.2. Seismic EventNatural events such as earthquakes have potential to cause plant structure or equipmentdamage of sufficient magnitude to threaten personnel or plant safety.3. Natural or Technolo-gy HazardOther natural and non-naturally occurring events that can cause damage to plant facilitiesinclude tornados, FLOODING, hazardous material releases and events restricting siteaccess warranting classification.4. FireFires can pose significant hazards to personnel and reactor safety. Appropriate forclassification are fires within the site Protected Area or which may affect operability ofequipment needed for safe shutdown5. Hazardous GasToxic, corrosive, asphyxiant or flammable gas leaks can affect normal plant operations orpreclude access to plant areas required to safely shutdown the plant.6. Control Room EvacuationEvents that are indicative of loss of Control Room habitability. If the Control Room must beevacuated, additional support for monitoring and controlling plant functions is necessarythrough the emergency response facilities.Page 118 of 276 1 ATTACHMENT IEAL Bases7. Emeraencv Coordinator JudQmentThe EALs defined in other categories specify the predetermined symptoms or events thatare 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 emergencies basedon operator/management experience and judgment is still necessary. The EALs of thiscategory provide the Emergency Coordinator the latitude to classify emergency conditionsconsistent with the established classification criteria based upon Emergency Coordinatorjudgment.Page 119 of 276 ATTACHMENT 1EAL BasesCategory: H -HazardsSubcategory: 1 -SecurityInitiating Condition: Confirmed SECURITY CONDITION or threatEAL:HUI.1 Unusual EventA SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by theSecurity Shift SupervisorMode Applicability:AllDefinition(s):SECURITY CONDITION -Any security event as listed in the approved security contingencyplan that constitutes a threat/compromise to site security, threat/risk to site personnel, or apotential degradation to the level of safety of the plant. A security condition does not involve ahostile action.HOSTILE ACTION -An act toward CPNPP or its personnel that includes the use of violentforce to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end.This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or otherdevices used to deliver destructive force. Other acts that satisfy the overall intent may beincluded. Hostile action should not be construed to include acts of civil disobedience orfelonious acts that are not part of a concerted attack on CPNPP. Non-terrorism-based EALsshould be used to address such activities (i.e., this may include violent acts betweenindividuals in the owner controlled area).CPNPP Basis:The security shift supervision is defined as the Security Shift Supervisor.This EAL is based on the CPNPP Safeguards Contingency Plan (ref. 1).NEI 99-01 Basis:This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEMequipment, and thus represent a potential degradation in the level of plant safety. Securityevents which do not meet one of these EALs are adequately addressed by the requirements of10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS areclassifiable under ICs HA1, HS1 and HGI.Timely and accurate communications between Security Shift Supervision and the ControlRoom is essential for proper classification of a security-related event. Classification of theseevents will initiate appropriate threat-related notifications to plant personnel and OffsiteResponse Orqanizations.Security plans and terminology are based on the guidance provided by NEI 03-12, Templatefor the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [aP4ndoondPnt Spent Fuel Storage Installation Se276iy P-rogram.Page 120 of 276 ATTACHMENT 1EAL Bases&AL--#-This EAL references (site speGifiG-the seGU-ity-_shift Shift SecuritySupervisor because these are the individuals trained to confirm that a security event isoccurring or has occurred. Training on security event confirmation and classification iscontrolled due to the nature of Safeguards and 10 CFR § 2.39 information.Emergency plans and implementing procedures are public documents; therefore, EALs shouldnot incorporate Security-sensitive information. This includes information that may beadvantageous to a potential adversary, such as the particulars concerning a specific threat orthreat location. Security-sensitive information should be contained in non-public documentssuch as the CPNPP Safeguards Contingency Plan (ref. 1).Escalation of the emergency classification level would be via IC HAl.CPNPP Basis Reference(s):1. CPNPP Safeguards Contingency Plan (Safeguards)2. NEI 99-01 HU1Page 121 of 276 i ATTACHMENT 1EAL BasesCategory: H -HazardsSubcategory: 1 -SecurityInitiating Condition: Confirmed SECURITY CONDITION or threatEAL:HUI.2 Unusual EventNotification of a credible security threat directed at the siteMode Applicability:AllDefinition(s):SECURITY CONDITION -Any security event as listed in the approved security contingencyplan that constitutes a threat/compromise to site security, threat/risk to site personnel, or apotential degradation to the level of safety of the plant. A security condition does not involve ahostile action.HOSTILE ACTION- An act toward CPNPP or its personnel that includes the use of violentforce to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end.This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or otherdevices used to deliver destructive force. Other acts that satisfy the overall intent may beincluded. Hostile action should not be construed to include acts of civil disobedience orfelonious acts that are not part of a concerted attack on CPNPP. Non-terrorism-based EALsshould be used to address such activities (i.e., this may include violent acts betweenindividuals in the owner controlled area).CPNPP Basis:The security shift supervision is defined as the Security Shift Supervisor.This EAL is based on the CPNPP Safeguards Contingency Plan (ref. 1).NEI 99-01 Basis:This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEMequipment, and thus represent a potential degradation in the level of plant safety. Securityevents which do not meet one of these EALs are adequately addressed by the requirements of10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS areclassifiable under ICs HA1, HS1 and HGI.Timely and accurate communications between Security Shift Supervision and the ControlRoom is essential for proper classification of a security-related event. Classification of theseevents will initiate appropriate threat-related notifications to plant personnel and OffsiteResponse Orqanizations.Security plans and terminology are based on the guidance provided by NEI 03-12, Templatefor the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [aRdndpcndent Spent Fuel Storage installation Seur.ty Po.gra., .EAL #1 references (sitespecific security Shift supeiion) because these are the i d trained to cnR thatsecurity event is ocurner has occGurred. Trainin on security event confirmation andPage 122 of 276 ATTACHMENT 1EAL Basesis cRntrlled due to the nature of Saf.guards and 10 CIFR § 2.39 EAL #2 This EAL addresses the receipt of a credible security threat. The credibility of thethreat is assessed in accordance with (site-specific procedure).EA6L #3 addresses the threat from thoe of an aircraft on the plant. The NRI mIeadguartors Operations Officer (HOO) Will commUnicate to the licensee if the threat innvoVesan aircraft. The status and size of thc plane mnay also be provided by NORA.D through theNRC. Valid-ation of the threat is performed in accordance with (site specific proceduro).Emergency plans and implementing procedures are public documents; therefore, EALs shouldnot incorporate Security-sensitive information. This includes information that may beadvantageous to a potential adversary, such as the particulars concerning a specific threat orthreat location. Security-sensitive information should be contained in non-public documentssuch as the CPNPP Safequards Contingrency Plan (ref. 1).Escalation of the emergency classification level would be via IC HA1.CPNPP Basis Reference(s):1. CPNPP Safeguards Contingency Plan (Safeguards)2. NEI 99-01 HU1IPage 123 of 276 ATTACHMENT 1EAL BasesCategory: H -HazardsSubcategory: I -SecurityInitiating Condition: Confirmed SECURITY CONDITION or threatEAL:HUl.3 Unusual EventA validated notification from the NRC providing information of an aircraft threatMode Applicability:AllDefinition(s):NoneCPNPP Basis:This EAL is based on the CPNPP Safeguards Contingency Plan (ref. 1).NEI 99-01 Basis:This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEMequipment, and thus represent a potential degradation in the level of plant safety. Securityevents which do not meet one of these EALs are adequately addressed by the requirements of10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS areclassifiable under ICs HA1, HS1 and HGI.Timely and accurate communications between Security Shift Supervision and the ControlRoom is essential for proper classification of a security-related event. Classification of theseevents will initiate appropriate threat-related notifications to plant personnel and OffsiteResponse Orqanizations.Security plans and terminology are based on the guidance provided by NEI 03-12, Templatefor the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [a&REAL-#3This EAL addresses the threat from the impact of an aircraft on the plant. The NRCHeadquarters Operations Officer (HOO) will communicate to the licensee if the threat involvesan aircraft. The status and size of the plane may also be provided by NORAD through theNRC. Validation of the threat is performed in accordance with (site-specific procedure).Emergency plans and -implementing procedures are public documents; therefore, EALs shouldnot incorporate Security-sensitive information. This includes information that may beadvantageous to a potential adversary, such as the particulars concerning a specific threat orthreat location. Security-sensitive information should be contained in non-public documentssuch as the CPNPP Safeguards Continqency Plan (ref. 1).Escalation of the emergency classification level would be via IC HAl.CPNPP Basis Reference(s):1. CPNPP Safeguards Contingency Plan (Safeguards)Page 124 of 276 ATTACHMENT 1EAL Bases2. NEI 99-01 HU1Category: H -HazardsSubcategory: 1 -SecurityInitiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA orairborne attack threat within 30 minutesEAL:HAI.1 AlertA HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLEDAREA as reported by the Security Shift SupervisorMode Applicability:AllDefinition(s):HOSTILE ACTION -An act toward CPNPP or its personnel that includes the use of violentforce to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end.This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or otherdevices used to deliver destructive force. Other acts that satisfy the overall intent may beincluded. Hostile action should not be construed to include acts of civil disobedience orfelonious acts that are not part of a concerted attack on CPNPP. Non-terrorism-based EALsshould be used to address such activities (i.e., this may include violent acts betweenindividuals in the owner controlled area).OWNER CONTROLLED AREA -As shown in CPNPP Emergency Plan Appendix E, Complexand Owner Controlled Area.CPNPP Basis:The security shift supervision is defined as the Security Shift Supervisor.NEI 99-01 Basis:This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLEDAREA or notification of an aircraft attack thrcFat. This event will require rapid response andassistance due to the possibility of the attack progressing to the PROTECTED AREA, or theneed to prepare the plant and staff for a potential aircraft impact.Timely and accurate communications between the Security Shift uperiAsioR -Supervisor andthe Control Room is essential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Templatefor the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan-[aiwdIndopondont Spent Fu~el Storage Installation Se G1rFtY P-ro grFam.As time and conditions allow, these events require a heightened state of readiness by the plantstaff and implementation of onsite protective measures (e.g., evacuation, dispersal orsheltering). The Alert declaration will also heighten the awareness of Offsite ResponseI Organizations (OROs), allowing them to be better prepared should it be necessary to considerfurther actions.Page 125 of 276 ATTACHMENT 1EAL BasesThis IC does not apply to incidents that are accidental events, acts of civil disobedience, orotherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples includethe crash of a small aircraft, shots from hunters, physical disputes between employees, etc.Reporting of these types of events is adequately addressed by other EALs, or therequirements of 10 CFR § 73.71 or 10 CFR § 50.72.EAL-#!This EAL is applicable for any HOSTILE ACTION occurring, or that has occurred, in theOWNER CONTROLLED AREA. This includes any action directed against an ISFSI that islocated outside the plant PROTECTED AREA.EAL #taddresses the threat f9Rom the imnpact of an aircraft on the plant, and the anticipatednotific-';ations arc made in a timely mRannr so that plant personnRel and OROsr are inheightened state of readireSS. This EAL is met when the threat related iRnformation has beenva-lid,-afted on aG,,,,G l-Jn,'aR w^ith (site fifa The NRC Headquarters Operations Officer (HO00) Will communicate to the licensee if thethreat involves an aircraft. The status and size of the plane may be provided by NORAIDthr.ugh the NRC.Inn Sme cases, it t be readily apparent if an aircraft impat within the OWNERCONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). it is expected, although notcertain, that notification A ' by an appropriate Federal agency to the site would .larif,' this point.÷ Inthis ease, the appropriate federal agency is intended to be NORAID, FBI, FA. or NRC. Theemnergency declar-ation, including onRe -b-ased on other l~siFEALs, should not be unduly delayewhile awaiting notification by a Federal agcncy.Emergency plans and implementingprocedures are public documents; therefore, EALs should not incorporate Security-sensitiveinformation. This includes information that may be advantageous to a potential adversary,such as the particulars concerning a specific threat or threat location. Security-sensitiveinformation should be contained in non-public documents such as the CPNPP SafeguardsContingiency Plan (ref. 1).CPNPP Basis Reference(s):1. CPNPP Safeguards Contingency Plan (Safeguards)2. NEI 99-01 HA1IPage 126 of 276 ATTACHMENT 1EAL BasesCategory: H -HazardsSubcategory: 1 -SecurityInitiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA orairborne attack threat within 30 minutesEAL:HA1.2 AlertA validated notification from NRC of an aircraft attack threat within 30 min. of the siteMode Applicability:AllDefinition(s):NoneCPNPP Basis:NoneNEI 99-01 Basis:This IC addresses the occurrence of a HOSTILE ACTION withiR the OWNER CONTROLLEDAREA or notification of an aircraft attack threat. This event will require rapid response andassistance due to the possibility of the attack progressing to the PROTECTED AREA, or theneed to prepare the plant and staff for a potential aircraft impact.I Timely and accurate communications between the Security Shift Su pe iSieR-Supervisor andthe Control Room is essential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Templatefor the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan-[andIndependent Spent Fue! Storage Installation SocuritY P-rogram]ý1.As time and conditions allow, these events require a heightened state of readiness by the plantstaff and implementation of onsite protective measures (e.g., evacuation, dispersal orsheltering). The Alert declaration will also heighten the awareness of Offsite ResponseOrganizations (OROs), allowing them to be better prepared should it be necessary to considerfurther actions.This IC does not apply to incidents that are accidental events, acts of civil disobedience, orotherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples includethe crash of a small aircraft, shots from hunters, physical disputes between employees, etc.Reporting of these types of events is adequately addressed by other EALs, or therequirements of 10 CFR § 73.71 or 10 CFR § 50.72.EAL #1 is applicable for any H"OSTILn EAC''TPI"ION oc-curring, or that has -occured, in the OWNEA R N E D AR EA This inql- cludepan',' action against an ISSI- that is loc-Gated outside the plant PROTECTED [AREAEAL-#2This EAL addresses the threat from the impact of an aircraft on the plant, and theanticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in aPage 127 of 276 ATTACHMENT 1EAL Basesheightened state of readiness. This EAL is met when the threat-related information has beenvalidated in accordance with (site-specific security procedures).The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if thethreat involves an aircraft. The status and size of the plane may be provided by NORADthrough the NRC.In some cases, it may not be readily apparent if an aircraft impact within the OWNERCONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although notcertain, that notification by an appropriate Federal agency to the site would clarify this point. Inthis case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. Theemergency declaration, including one based on other ICs/EALs, should not be unduly delayedwhile awaiting notification by a Federal agency.Emergency plans and implementing procedures are public documents; therefore, EALs shouldnot incorporate Security-sensitive information. This includes information that may beadvantageous to a potential adversary, such as the particulars concerning a specific threat orthreat location. Security-sensitive information should be contained in non-public documentssuch as the CPNPP Safeguards Contingency Plan (ref. 1).CPNPP Basis Reference(s):1. CPNPP Safeguards Contingency Plan (Safeguards)2. NEI 99-01 HAIPage 128 of 276 ATTACHMENT 1EAL BasesCategory: H -HazardsSubcategory: 1 -SecurityInitiating Condition: HOSTILE ACTION within the PROTECTED AREAEAL:HSI.1 Site Area EmergencyA HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA asreported by the Security Shift SupervisorMode Applicability:AllDefinition(s):HOSTILE ACTION -,An act toward CPNPP or its personnel that includes the use of violentforce to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end.This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or otherdevices used to deliver destructive force. Other acts that satisfy the overall intent may beincluded. Hostile action should not be construed to include acts of civil disobedience orfelonious acts that are not part of a concerted attack on CPNPP. Non-terrorism-based EALsshould be used to address such activities (i.e., this may include violent acts betweenindividuals in the owner controlled area).PROTECTED AREA -An area encompassed by physical barriers and to which access iscontrolled. The Protected Area refers to the designated security area around the processbuildings and is depicted in FSAR Figure 1.2-1 Plot Plan.CPNPP Basis:The security shift supervision is defined as the Security Shift Supervisor.These individuals are the designated on-site personnel qualified and trained to confirm that asecurity event is occurring or has occurred. Training on security event classificationconfirmation is closely controlled due to the strict secrecy controls placed on the CPNPPSafeguards Contingency Plan (Safeguards) information. (ref. 1)NEI 99-01 Basis:This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA.This event will require rapid response and assistance due to the possibility for damage to plantequipment.Timely and accurate communications between Security Shift Supervision and the ControlRoom is essential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Templatefor the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [a&RIndependent Spent Fue! Storage. n'stallatioAn Security Pregram].As time and conditions allow, these events require a heightened state of readiness by the plantstaff and implementation of onsite protective measures (e.g., evacuation, dispersal orPage 129 of 276 ATTACHMENT 1EAL Basessheltering). The Site Area Emergency declaration will mobilize Offsite Response Organization(ORO) resources and have them available to develop and implement public protective actionsin the unlikely event that the attack is successful in impairing multiple safety functions.This IC does not apply to a HOSTILE ACTION directed at an ISFSI PROTECTED AREAlocated outside the plant PROTECTED AREA; such an attack should be assessed using ICHAl. It also does not apply to incidents that are accidental events, acts of civil disobedience,or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examplesinclude the crash of a small aircraft, shots from hunters, physical disputes between employees,etc. Reporting of these types of events is adequately addressed by other EALs, or therequirements of 10 CFR § 73.71 or 10 CFR § 50.72.Emergency plans and implementing procedures are public documents; therefore, EALs shouldnot incorporate Security-sensitive information. This includes information that may beadvantageous to a potential adversary, such as the particulars concerning a specific threat orthreat location. Security-sensitive information should be contained in non-public documentssuch as the CPNPP Safeguards Contingency Plan (ref. 1).Escalation of the emergency classification level would be via IC HG1.CPNPP Basis Reference(s):1. CPNPP Safeguards Contingency Plan (Safeguards)2. NEI 99-01 HS1Page 130 of 276 ATTACHMENT 1EAL BasesCategory:H -HazardsSubcategory: 1 -SecurityInitiating Condition: HOSTILE ACTION resulting in loss of physical control of the facilityEAL:HGI.1 General EmergencyA HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA asreported by the Security Shift SupervisorAND EITHER of the following has occurred:" One or more of the following safety functions cannot be controlled or maintained-Reactivity control-Core cooling-RCS heat removalOR" Damage to spent fuel has occurred or is IMMINENTMode Applicability:AllDefinition(s):HOSTILE ACTION -An act toward CPNPP or its personnel that includes the use of violentforce to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end.This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or otherdevices used to deliver destructive force. Other acts that satisfy the overall intent may beincluded. Hostile action should not be construed to include acts of civil disobedience orfelonious acts that are not part of a concerted attack on CPNPP. Non-terrorism-based EALsshould be used to address such activities (i.e., this may include violent acts betweenindividuals in the owner controlled area).IMMINENT -The trajectory of events or conditions is such that an EAL will be met within arelatively short period of time regardless of mitigation or corrective actionsPROTECTED AREA -An area encompassed by physical barriers and to which access iscontrolled. The Protected Area refers to the designated security area around the processbuildings and is depicted in FSAR Figure 1.2-1 Plot Plan.CPNPP Basis:The security shift supervision is defined as the Security Shift Supervisor.NEI 99-01 Basis:This IC addresses an event in which a HOSTILE FORCE has taken physical control of thefacility to the extent that the plant staff can no longer operate equipment necessary to maintainkey safety functions. It also addresses a HOSTILE ACTION leading to a loss of physicalcontrol that results in actual or IMMINENT damage to spent fuel due to 1) damage to a spentPage 131 of 276 I ATTACHMENT 1EAL Basesfuel pool cooling system (e.g., pumps, heat exchangers, controls, etc.) or, 2) loss of spent fuelpool integrity such that sufficient water level cannot be maintained.Timely and accurate communications between Security Shift Supervision and the ControlRoom is essential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Templatefor the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [andIndependent Spent Fu4el Storage Installation Socbrity P-rogram].Emergency plans and implementing procedures are public documents; therefore, EALs shouldnot incorporate Security-sensitive information. This includes information that may beadvantageous to a potential adversary, such as the particulars concerning a specific threat orthreat location. Security-sensitive information should be contained in non-public documentssuch as the CPNPP Safeguards Contingency Plan (ref.1).CPNPP Basis Reference(s):1. CPNPP Safeguards Contingency Plan (Safeguards)2. NEI 99-01 HG1Page 132 of 276 ATTACHMENT 1EAL BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 2 -Seismic EventInitiating Condition: Seismic event greater than OBE levelEAL:HU2.1 Unusual EventSeismic event greater than OBE as indicated by annunciator 2A-3.1, OBE EXCEEDED, oryellow OBE light on Seismic Monitoring system panelMode Applicability:AllDefinition(s):NoneCPNPP Basis:Seismic events of this magnitude can result in a areas needed for safe shutdown beingsubjected to forces beyond design limits, and thus damage may be assumed to have occurredto plant safety systems.A conservative Safe Shutdown Earthquake (SSE) having a peak horizontal groundacceleration at the top of bedrock of 0.12 g has been selected for design (FSAR Section2.5.2.6). The Operating Basis Earthquake (OBE) is equal to 1/2 the SSE (ref. 1).When the seismic recorder indicates that the OBE has been exceeded, System Engineeringmust evaluate and determine whether the reactor must be shut down and remain shutdownuntil inspection of the facility shows that no damage has been incurred which would jeopardizesafe operation of the facility or until such damage is repaired. CPNPP was designed such that,for ground motion less than the OBE, the features of the plant necessary for continuedoperation without undue risk to the health and safety of the public will remain functional. Anyground motion in excess of this results in an uncertainty as to the extent of the damage whichmust be resolved before continued operation can be considered safe (ref. 2).The seismic trigger, CP1-SIATAS-03, is set to activate the strong motion recording system atan acceleration level slightly above normal ambient vibrations (0.01g) and well below thepostulated OBE "free field" ground acceleration (0.06g horizontal). This causes an alarm in thecontrol room to alert the operator. (ref. 2, 3) The seismic recorders (strong motionaccelerators) monitor earth vibration and, when triggered, store data in the recorder. TriaxialSMAs are installed at appropriate locations to provide data on the frequency, amplitude, andphase relationship of the seismic response of the containment structure and the seismic inputto other seismic Category I structures, systems, and components. The Seismic Instrumentationconsists of strong motion accelerograph (triaxial time history accelerograph system), triaxialpeak accelerograph recorders, passive response spectrum recorders, a response spectrumswitch, and a seismic switch. Except for sensors for the active instrumentation, all electronicsfor processing and storage of the seismic data are located in the seismic instrumentation panelCPX-ECPRCV-1 1 in the control room. There is no additional seismic instrumentation requiredfor Unit 2. However, alarms from seismic instrumentation in Unit 1 are duplicated in Unit 2. ThePage 133 of 276 ATTACHMENT 1EAL Basestime history accelerograph system is fully operational within 0.1 second after the seismictrigger is actuated.It will operate continuously during the period in which the earthquake exceeds the seismictrigger threshold (0.01g) plus 5 seconds (minimum) beyond the last seismic trigger signal.ABN-907 Acts of Nature provides the guidance for determining if the OBE earthquakethreshold is exceeded and any required response actions. (ref. 2)To avoid inappropriate emergency classification resulting from spurious actuation of theseismic instrumentation or felt motion not attributable to seismic activity, an offsite agency(USGS, National Earthquake Information Center) can confirm that an earthquake has occurredin the area of the plant. Such confirmation should not, however, preclude a timely emergencydeclaration based on receipt of the OBE alarm. The NEIC can be contacted by calling (303)273-8500. Select option #1 and inform the analyst you wish to confirm recent seismic activityin the vicinity of CPNPP. Alternatively, near real-time seismic activity can be accessed via theNEIC website:http://earthquake.usgs.gov/earthquakes/dyfi/archives.phpNEI 99-01 Basis:This IC addresses a seismic event that results in accelerations at the plant site greater thanthose specified for an Operating Basis Earthquake (OBE). An earthquake greater than anOBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact onsafety-related systems, structures and components; however, some time may be required forthe plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections). Given the time necessary to perform walk-downs andinspections, and fully understand any impacts, this event represents a potential degradation ofthe level of safety of the plant.Event verification with external sources should not be necessary during or following an OBE.Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as aseismic event (e.g., lateral accelerations in excess of 0.08g). The Shift Manager orEmergency IDEeeter-Coordinator may seek external verification if deemed appropriate (e.g., acall to the USGS, check internet news sources, etc.); however, the verification action must notpreclude a timely emergency declaration.Depending upon the plant mode at the time of the event, escalation of the emergencyclassification level would be via IC CA6 or SA9.CPNPP Basis Reference(s):1. FSAR Section 2.5.4.9 Earthquake Design Basis2. ABN-907 Acts of Nature3. DBD-EE-077 Seismic Instrumentation4. 1, 2-ALB-2A-3.1 OBE EXCEEDED5. DBD-ME-028 Classification of Structures, Systems and Components6. NEI 99-01 HU2Page 134 of 276 ATTACHMENT 1EAL BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 3 -Natural or Technology HazardInitiating Condition: Hazardous eventEAL:HU3.1 Unusual EventA tornado strike within the PROTECTED AREAMode Applicability:AllDefinition(s):PROTECTED AREA -An area encompassed by physical barriers and to which access iscontrolled. The Protected Area refers to the designated security area around the processbuildings and is depicted in FSAR Figure 1.2-1 Plot Plan.CPNPP Basis:Response actions associated with a tornado onsite is provided in ABN-907 Acts of Nature (ref.1).If damage is confirmed visually or by other in-plant indications, the event may be escalated toan Alert under EAL CA6.1 or SA9.1.A tornado striking (touching down) within the PROTECTED AREA warrants declaration of anUnusual Event regardless of the measured wind speed at the meteorological tower. A tornadois defined as a violently rotating column of air in contact with the ground and extending fromthe base of a thunderstorm.NEI 99-01 Basis:This IC addresses hazardous events that are considered to represent a potential degradationof the level of safety of the plant.I EAL #1EAL HU3.1 addresses a tornado striking (touching down) within the PROTECTEDAREA.EAL #2 addresses flooding of a building roomn or area that results in operators isolating powcrto a SAFETY SYSTEM compnenRGt du-e to wgater level oer o-ther wetting cARonerns. Cla~ssificloatio~n.is also rcguircd if the water level or rclated Wetting causcs an automatic isolation of a SAFEPSYSTEM comnponent from its power source (e.g., a breaker or relay trip). To9 warrnclassification, operability of the affected component must be reguircd by TcehnicalSpecifications for the current operating moede.-EAL #3 addresses a hadosmtrlsevent originating at an eff-sitc loc-ation and ofsufiietmagnitude to impede the moevement Of peF8rRonnl Within t-hc PROTEC=GTEFD AREA.EAL #4 addrcsscs a hazardous event that c~auses an on site impedimcnet to Vehicle moVmentand Significant enough topo ibthe plant staff from accesSing the site usngprsoalvehcis.Exape ofsuch. an. event include site flooding caused by -a huMrricamne, hcaVY rains,Page 135 of 276 ATTACHMENT 1EAL Basesup rier wter releases, damn failure, etc., Or an on site train derailment blocking the accessThis EAL is inteRded apply to routine impediments such as fog, snow, i.e, or Vehi-lebreakdowns or accidents, but rather to mrnee significant conditions such as the HurricaneAndrew strike on Turkey Point in 1992, the flooding around the Cooper Station during theMidw~est floods of 1993, or the flooding aroun~d Ft. Calhoun Station in 2011.EAL #5 addresses (site specific description).Escalation of the emergency classification level would be based on ICs in RecognitionCategories AR, F, S or C.CPNPP Basis Reference(s):1. ABN-907 Acts of Nature2. NEI 99-01 HU3I Page 136 of 276 ATTACHMENT 1EAL BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 3 -Natural or Technology HazardInitiating Condition: Hazardous eventEAL:HU3.2 Unusual EventInternal room or area FLOODING of a magnitude sufficient to require manual or automaticelectrical isolation of a SAFETY SYSTEM component needed for the current operatingmodeMode Applicability:AllDefinition(s):FLOODING -A condition where water is entering a room or area faster than installedequipment is capable of removal, resulting in a rise of water level within the room or area.SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/orplacing it in the cold shutdown condition, including the ECCS. These are typically systemsclassified as safety-related (as defined in 10CFR50.2):Those structures, systems and components that are relied upon to remain functional duringand 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 could resultin potential offsite exposures.CPNPP Basis:The internal flooding areas of concern are the Safeguards Building and Turbine Building(ref.1). Refer to EAL CA6.1 for internal flooding affecting one or more SAFETY SYSTEMtrains.NEI 99-01 Basis:This IC addresses hazardous events that are considered to represent a potential degradationof the level of safety of the plant.EAL #1 addresses a tornado striking (touching doWn) Within tho_ P2ROTECQTED ARE=A.This EAL addresses FLOODING of a building room or area that results in operators isolatingpower to a SAFETY SYSTEM component due to water level or other wetting concerns.Classification is also required if the water level or related wetting causes an automatic isolationof a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). Towarrant classification, operability of the affected component must be required by TechnicalSpecifications for the current operating mode.Page 137 of 276 ATTACHMENT 1EAL BasesEAL #3 address-sc-es aa materials evPent rigiRnating at an f-fsite lcatio;n and ofSUfficieRt magnitude to impede the movement of pcrSOnncl Within thc PROTECTED rAREA.AE:AI #1 addresses a h.zArdou e.....vent that c.au.ses an onsite impediment to vehicle movementand significant cnough to prohibit the plant staff fro~m acc-pessing the site using personRalvehicles. Examples oif such an event include site flooding caused by a heavy rains&,Iup rier wter releases, dam failure, etc., or an on site train derailment blocking the accessThis EAL intended apply to routine impediments .uch as fog, snow, iGe, or vehiclebhre-akdoAwns or accidents, but rather to imopree significant conditions such As6 the- HurricaneAndrew strike on Turkey Point in 1992, the floodin;g around the Cooper Station during theMidwest flo-dis o-f 1Q993, or the flooding around Ft. C-alhoun61 Station in 20-11.EAL #5 addresses (site specific description).Escalation of the emergency classification level would be based on ICs in Recognition Categories AR,F, SorC.CPNPP Basis Reference(s):1. CPNPP PRA Accident Sequence Analysis "Internal Flooding Sequences"2. NEI 99-01 HU3Page 138 of 276 ATTACHMENT 1EAL BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 3 -Natural or Technology HazardInitiating Condition: Hazardous eventEAL:HU3.3 Unusual EventMovement of personnel within the PROTECTED AREA is IMPEDED due to an offsiteevent involving hazardous materials (e.g., an offsite chemical spill or toxic gas release)Mode Applicability:AllDefinition(s):IMPEDE(D) -Personnel access to a room or area is hindered to an extent that extraordinarymeasures are necessary to facilitate entry of personnel into the affected room/area(e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).PROTECTED AREA -An area encompassed by physical barriers and to which access iscontrolled. The Protected Area refers to the designated security area around the processbuildings and is depicted in FSAR Figure 1.2-1 Plot Plan.CPNPP Basis:As used here, the term "offsite" is meant to be areas external to the CPNPP PROTECTEDAREA.NEI 99-01 Basis:This IC addresses hazardous events that are considered to represent a potential degradationof the level of safety of the plant.EAL 11 addresses a terade stFriking (touching dwnr) within the PROTECTED AREA.This EA'l addresses flooding of a buildi'g room or area that .esults in operators isolatirgpower to a SAFETY SYSTEM co4r-mponont due to water level or other wetting conceFrn.Classificatioen is also required if the water level or related wetting causes an automatic isolatioof a SAFETY SYSTEM compoenet froM its power sourcGe (o.g., a breaker or relay trip). ToGwarrant classification, operability of the affected component must be required by Technicalpecifiations for the Gcurent operating mode.EAL-#3This EAL addresses a hazardous materials event originating at an offsite location andof sufficient magnitude to impede the movement of personnel within the PROTECTED AREA.EAL #4 addresses a hazardous event that c~auses an on site impediment to vehicle movemenand significant enough to prohibit the plant staff from accessing the Site using personalvehicloes. Examples Of such an event include site flooding caused by a hurricane, heav' rains&,up rior wter releases, dam failure, etc., or an on site train derailment blocking the accessPage 139 of 276 ATTACHMENT 1EAL BasesThis EAL is notntended apply to routineim;pediments 6uch as fog, snow, ice, or vehiclebre~akdoGwns or accidcnts, but rather to) mor)e Significant conditionRs such as the HurcnAndrew strike.on Turkey Point in 1992, the flooding around the .r. Qtati-"n during theMidwest floods of 1993, Or the flooding aroun~d Ft. Calhoun Station in 2011.E=AL #f5 addresses (site specific description).Escalation of the emergency classification level would be based on ICs in RecognitionCategories AR, F, S or C.CPNPP Basis Reference(s):1. NEI 99-01 HU3Page 140 of 276 ATTACHMENT 1EAL BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 3 -Natural or Technology HazardInitiating Condition: Hazardous eventEAL:HU3.4 Unusual EventA hazardous event that results in on-site conditions sufficient to prohibit the plant staff fromaccessing the site via personal vehicles (Note 7)Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns oraccidents.Mode Applicability:AllDefinition(s):NoneCPNPP Basis:NoneNEI 99-01 Basis:This IC addresses hazardous events that are considered to represent a potential degradationof the level of safety of the plant.EAL #1 -,,e,4 a t .... ,,euh. @ down) ,i;hi14the PROTECTED AREA.This EAL addresses flooding of a building reem o area that results in operatorsi ipovee to a SAFETY SYSTEM comproent due to water level or other wetting concerns.lassificatine is also required if the water level or related wetting causes an automatic isolationef a SAFETY SYSTEM comrponent fom its powf r seouce (e.g., a breaker or relay trip). Towarrant classification, operability of the affected compoenet must be required by TechnicalSpecfcations for the current operating mode.EAL #3 addresses a hazardous materials event originating at an offsite location and ofsufficient magnitude to impede the mo'erncent of personnel w~ithin the PROTECTED ,AREA.ELAL44This EAL addresses a hazardous event that causes an on-site impediment to vehiclemovement and significant enough to prohibit the plant staff from accessing the site usingpersonal vehicles. Examples of such an event include site FLOODING caused by a hurricane,heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blockingthe access road.This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehiclebreakdowns or accidents, but rather to more significant conditions such as the HurricaneAndrew strike on Turkey Point in 1992, the flooding around the Cooper Station during theMidwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011 .EAL #5 addresses(site specific description)Page 141 of 276 ATTACHMENT 1EAL BasesEscalation of the emergency classification level would be based on ICs in RecognitionI Categories AR, F, S or C.CPNPP Basis Reference(s):1. NEI 99-01 HU3Page 142 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:H -Hazards and Other Conditions Affecting Plant Safety4- FireInitiating Condition: FIRE potentially degrading the level of safety of the plantEAL:HU4.1 Unusual EventA FIRE is not extinguished within 15 min. of any of the following FIRE detectionindications (Note 1):" Report from the field (i.e., visual observation)* Receipt of multiple (more than 1) fire alarms or indications" Field verification of a single fire alarmANDThe FIRE is located within any Table H-1 areaNote 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Table H-1 Fire Areas* u-Containment* u-Safeguards Building* X-Auxiliary Building* X-Electrical & Control Building* X-Fuel Building.X-Service Water Intake Structure* u-Diesel Generator Building* u-Normal Switchgear Rooms* u-CST* u-RWSTMode Applicability:AllDefinition(s):FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drivebelts or overheated electrical equipment do not constitute fires. Observation of flame ispreferred but is NOT required if large quantities of smoke and heat are observed.CPNPP Basis:The 15 minute requirement begins with a credible notification that a fire is occurring, or receiptof multiple valid fire detection system alarms or field validation of a single fire alarm. The alarmis to be validated using available Control Room indications or alarms to prove that it is notspurious, or by reports from the field. Actual field reports must be made within the 15 minutePage 143 of 276 ATTACHMENT 1EAL Basestime limit or a classification must be made. If a fire is verified to be occurring by field report, the15 minute time limit is from the original receipt of the fire detection alarm.Table H-1 applies to buildings and areas housing equipment needed for safe shutdown(SAFETY SYSTEMS) (ref. 1, 2).NEI 99-01 Basis:This IC addresses the magnitude and extent of FIRES that may be indicative of a potentialdegradation of the level of safety of the plant.EAL-#4-The-For EAL HU4.1 the intent of the 15-minute duration is to size the FIRE and to discriminateagainst small FIRES that are readily extinguished (e.g., smoldering waste paper basket). Inaddition to alarms, other indications of a FIRE could be a drop in fire main pressure, automaticactivation of a suppression system, etc.Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm,indication, or report. For EAL assessment purposes, the emergency declaration clock starts atthe time that the initial alarm, indication, or report was received, and not the time that asubsequent verification action was performed. Similarly, the fire duration clock also starts atthe time of receipt of the initial alarm, indication or report.EAL #2This EAL addresses rece.pt of a single fire alarm, and thc existence of a FPRE is not verified(i.e., proyed or disproved) within 30 minutes of the alarm. UJpon receipt, operators will takeprompt actions to confirm the validity of a single fire alarm. For E=AL assessment purposes, th~e30 Minute clock starts at the time that the initial -alar~m w~as received, and not the time that asubsequent verification action was performed.A single fire alarm, absent other iniainsef a FIRE, may be indicative of equipment failuror I .a.puius activation, and not an actual FIRE. Fer this reaseo, additional time is allowed toverif;, the validity of the alarm. The 30 minute period is a reasonable amount of time_ todetermiRe if an actual FIRE exists; however, after that time, and absent information to thecontrary, it is assumed that an actual FIRE is in progress.if an a.tual FIRE is verified by a report from the field, then EAL #11 is immediately applicable1and the emergency must be declared if the FIRE is not extinguished within 15-minutes of threporFt. if the alarm is verified to be due to an equipmnent failure or a Gpurious activation, andthis .occur within 30 minutes of the receipt of the alarm, then this EAL is netappicable aRd no emeFrgency declaratin is warranted.in addition to a FIRE addressed by, EAL #i or EAL #t2, a FIRE within the plant PROTECTEDAREA not extinguished within 60 minutes may also potentially degrade the level of plantsafety. This basis extends to a FIRE occurring within the PROTECTED AREA of an ISFSilocated outside the plant PROTEC-TED ,AREA. [Sentence for plants with an ISFSI outside th~eplant Protec-ted Area]P::AI if a-FIRE= within the o IOF I for,, plants with an ISF=I outside the plant Protected Area]PROTECiGTED AREA is of sufficient size to requie a re se by an offeite firefighting agencyPage 144 of 276 [ | |||
ATTACHMENT 1EAL Bases(e.g., a local tovn Fire Department.), then the lev~,el of plant safety is potentially degraded. Theis needed to actively SUPPort firefighting efforts because the fire i beyond the capability of thoFire BrFigade to extinguiSh. Declaration is not neesar if--- th 1 gnc resources arc placed 9Rstand by, or supporting post extinguishment reor;r investigation actiolns.Basis DRel+nA 0 ... ..r.m.. A .n...ixv 0Appendix R to 10 CFR 50, states in part:Criterion 3 of Appendix A to this part specifies that "Structures, systems, andcomFponents important to safety shall be designed and located to minimize, consistentwith other safety requirements, the probability and effect of fires adepoin.WheR considering the effects of fire, those systems associated with achieving aRdmaintaining safe shutdown conditions assume major importance to safety becausedamage to themB can lead to core damage resulting from loss of coolant through boil off.Because fire man" affect safe shutdown systems aRd because the loss of functifn Ofsystems used to mitigate the consequences of design basis accidents under postfrco~nditions does not per se impact public safety, the need to limit fire damage to systemsrequired to -achieve and mnaintain safe shutdown conditions is greater than the need to-limit fire damage to those systems required to mitigate the consequences Of desig.nbasis accidents.In addition, Appendix R to 10 .FR 50, requires, among other considerations, the use of 1 hourfire barriers for the enclosure of cable and equipment and associated non safety cirut of one-redundant train (G.2.G). As used in EAL #2, the 30 minutes to verify a single alarm is wellwithin this wor-st ease 1 hour time period.Depending upon the plant mode at the time of the event, escalation of the emergencyclassification level would be via IC CA6 or SA9.CPNPP Basis Reference(s):1. CPNPP Fire Protection Report, Section 5.0 "Fire Safe Shutdown Equipment List"2. FSAR Section 7.4 Systems Required for Safe Shutdown3. NEI 99-01 HU4Page 145 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:H -Hazards and Other Conditions Affecting Plant Safety4- FireInitiating Condition: FIRE potentially degrading the level of safety of the plantEAL:HU4.2 Unusual EventReceipt of a single fire alarm (i.e., no other indications of a FIRE)ANDThe fire alarm is indicating a FIRE within any Table H-1 areaANDThe existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Table H-1 Fire Areas* u-Containment* u-Safeguards Building* X-Auxiliary Building* X-Electrical & Control Building" X-Fuel Building* X-Service Water Intake Structure" u-Diesel Generator Building* u-Normal Switchgear Rooms" u-CST* u-RWSTMode Applicability:AllDefinition(s):FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drivebelts or overheated electrical equipment do not constitute fires. Observation of flame ispreferred but is NOT required if large quantities of smoke and heat are observed.CPNPP Basis:The 30 minute requirement begins upon receipt of a single valid fire detection system alarm.The alarm is to be validated using available Control Room indications or alarms to prove that itis not spurious, or by reports from the field. Actual field reports must be made within the 30minute time limit or a classification must be made. If a fire is verified to be occurring by fieldreport, classification shall be made based on EAL HU4.1.Page 146 of 276 ATTACHMENT 1EAL BasesTable H-1 applies to buildings and areas housing equipment needed for safe shutdown(SAFETY SYSTEMS) (ref. 1, 2).NEI 99-01 Basis:This IC addresses the magnitude and extent of FIRES that may be indicative of a potentialdegradation of the level of safety of the plant.E-AL-#-The intent of the 15 minute duration is to size the FIRE and to discrimninatc against smallFIRES that are readily extinguished (c.g., smoldering waste papcr basket). In addition toala~rms, other indle~ations of a FIRE could be a drop in fire mnain pressure, automatic activatioof a suppression system, etc.Upon receipt, operators will take prompt actionS to cofirm the validity of an initial fire alarm,indication, or report. For EAL assessment purposes, the emergency declaration clock starts at.the time that the initial alarm, indication, Or report was received, and not the time thasubsequent verification action was performed. Similarly, the fire duration clock alsoe starts at.the time of receipt of the initial alarm, indication or report.EAL-#2-This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified(i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will takeprompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the30-minute clock starts at the time that the initial alarm was received, and not the time that asubsequent verification action was performed.A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failureor a spurious activation, and not an actual FIRE. For this reason, additional time is allowed toverify the validity of the alarm. The 30-minute period is a reasonable amount of time todetermine if an actual FIRE exists; however, after that time, and absent information to thecontrary, it is assumed that an actual FIRE is in progress.IIf an actual FIRE is verified by a report from the field, then EAL #1 HU4.1 is immediatelyapplicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spuriousactivation, and this verification occurs within 30-minutes of the receipt of the alarm, then thisEAL is not applicable and no emergency declaration is warranted.EALI-#-3In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant PROTECTEDAREA not extinguished within 60- minutes may also potentially degrade the level of plantsafety. Thi' s basis extends to a FIRE= oGccurrg within the PROTECTEDý AREA of an !S1located outside the plant P-ROTECTED AREA4. [Senten-Ge for plants with an ISFSI outside theplant Pr-etected Area]EAI-#4Basis-Related Requirements from Appendix RAppendix R to 10 CFR 50, states in part:Page 147 of 276 ATTACHMENT 1EAL BasesCriterion 3 of Appendix A to this part specifies that "Structures, systems, andcomponents important to safety shall be designed and located to minimize, consistentwith other safety requirements, the probability and effect of fires and explosions."When considering the effects of fire, those systems associated with achieving andmaintaining safe shutdown conditions assume major importance to safety becausedamage to them can lead to core damage resulting from loss of coolant through boil-off.Because fire may affect safe shutdown systems and because the loss of function ofsystems used to mitigate the consequences of design basis accidents under post-fireconditions does not per se impact public safety, the need to limit fire damage to systemsrequired to achieve and maintain safe shutdown conditions is greater than the need tolimit fire damage to those systems required to mitigate the consequences of designbasis accidents.In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hourfire barriers for the enclosure of cable and equipment and associated non-safety circuits of oneredundant train (G.2.c). As used in EAL-#2HU4.2, the 30-minutes to verify a single alarm iswell within this worst-case 1-hour time period.Depending upon the plant mode at the time of the event, escalation of the emergencyclassification level would be via IC CA6 or SA9.CPNPP Basis Reference(s):1. CPNPP Fire Protection Report, Section 5.0 "Fire Safe Shutdown Equipment List"2. FSAR Section 7.4 Systems Required for Safe Shutdown3. NEI 99-01 HU4I Page 148 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:EAL:H -Hazards and Other Conditions Affecting Plant Safety4 -FireFIRE potentially degrading the level of safety of the plantHU4.3 Unusual EventA FIRE within the ISFSI or plant PROTECTED AREA not extinguished within 60 min. ofthe initial report, alarm or indication (Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Mode Applicability:AllDefinition(s):FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drivebelts or overheated electrical equipment do not constitute fires. Observation of flame ispreferred but is NOT required if large quantities of smoke and heat are observed.PROTECTED AREA -An area encompassed by physical barriers and to which access iscontrolled. The Protected Area refers to the designated security area around the processbuildings and is depicted in FSAR Figure 1.2-1 Plot Plan.CPNPP Basis:NoneNEI 99-01 Basis:This IC addresses the magnitude and extent of FIRES that may be indicative of a potentialdegradation of the level of safety of the plant.The intent of the 15 minute duration is to size the FIRE and to discrhiminate against smallFIRES that are readily (e.g., smoldering waste papcr basket). in addition toalarms, othc .indications of a FIRE could be a drop infir. main pr e..ure, automatic activationof a suppression system+, et+.UJpon receipt, oper-ators will take prompt actions to co4nfr the validity of an initial fire alarm,indicatiEn, or repeot. For EAL assessment purpoSes, the em.ergency declaration clock starts atthe time that the initial alarm, indicatin, Or report was rece.ived, and not the time that asubsequent verification action was perormed. Similarly, the fire duration clock also starts atthe time of receipt of the initial alarm, indication or report.This EA- addresses receipt of a single fire alarm, and the existenRce of FIRE is net Verified(i.e., proved Or disproved) within 30 of the alarm. Upon receipt, opeFators will takep.ropt to confirm the validity of a single fire alarm. For EAL assessment purposes,hPage 149 of 276 ATTACHMENT 1EAL Bases30 minute clock starts at the time that the initial! alarm was received, and not the time that asubsequent verfification action was performed.A single fire alarmn, absen-t o-t-her indication(s) of a F=IRE, may be indicative of equipment failror a spurio~us activation, and not an actuial FIRE. ForF this reason, additional time is allowed toverif' the validity of the alarm. The 30 minute period is a reasonable amount Of tim~e todetermnine if an actual F=IRE= exists; however, after that time, and absent informnation to thecontrar',, it is assumed that an a-tual IRE is in progress.If an actual FIRE is verified by a repod from the field, then .AL #1 is immediately applicable,and the emergency must be declared if the FIRE is not extinguished within 15 m-iutes ef therepodt. if the alarmn is verified to be due to an equipment failure or a spurious activation, and-this verification occurs Within 30 mFinutes of the receipt of the alarm, then this EAL= is notapplicable and no) emergency declaration is warranted.EAL-#3In addition to a FIRE addressed by EAL #4-HU4.1 or EAL #2HU4.2, a FIRE within the plantPROTECTED AREA not extinguished within 60-minutes may also potentially degrade the levelof plant safety. This basis extends to a FIRE occurring within the PROTECTED AREA of anISFSI located outside the plant PROTECTED AREA. [SentenGe foAplants with an !SFSIout-side the plaRt Pr-tected Ara]eAI"/Aif a FIRE within the [for plants with an ISFS outsido the Prote.etd Are.a]PROTECTED AREA is of sufficient size to require a nse by an offsite firefighting agency(e.g., a localI twn1 Fire Depadtment), then the level of plant safet is potentially degraded. Theis6 needed to actively suppedt firefighting effod-s because the fire is beyond the capability of theFire Brigade to extinguish. Declaration is not necessar,' if the agency resources are placed ostand by, or suppoding post- extinguishment recve ;orivestigation actions.Basis Related 0 Re.. {ir .. ...+ ,.,m A nn.R. ;,d .Appendix R to 10 CF=R 50, states in pad:-Criterion 3 of Appendix A to this padt specifies that "Structures, systems, andcomponents impodtant to safet shall be designed and located to minimize cnIstent4with other safety requirements, the probability and effect of fires and explio,-nos."When considering the effects of fire, those systemns associated with achieving andmaintaining safe shutdown conditionRs assumne mjo impedtance to safet becauseedamage to themR can lead to core damage resulting fromn loss of coolant through boil off.Beause fire may affect safe shutdown systems and because the loSs Of funcItion ofsystemns used to mitigate the consequences of design basis accidents under postfrconditions does not per se impact public safety, the need to limit fire damage to systemnsrequired to achieve and maintain safe shutdown conditions is greater than the need tolimit fire damage to those systems required to mitigate the consequences of designbasis accidents%.in addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1 hourfire barriers for the enclosure of cablean-d e-quipment and associated non safety iruits of. onePage 150 of 276 ATTACHMENT IEAL Basesred-undiant train (G.2.c). As used in EAL #2, the 30 minutes to verif' a single alarm is wellwithin this wor.st case 1 hour time .periOd.Depending upon the plant mode at the time of the event, escalation of the emergencyclassification level would be via IC CA6 or SA9.CPNPP Basis Reference(s):1. NEI 99-01 HU4Page 151 of 276 ATTACHMENT 1EAL BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 4- FireInitiating Condition: FIRE potentially degrading the level of safety of the plantEAL:HU4.4 Unusual EventA FIRE within the ISFSI or plant PROTECTED AREA that requires firefighting support byan offsite fire response agency to extinguishMode Applicability:AllDefinition(s):FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drivebelts or overheated electrical equipment do not constitute fires. Observation of flame ispreferred but is NOT required if large quantities of smoke and heat are observed.PROTECTED AREA -An area encompassed by physical barriers and to which access iscontrolled. The Protected Area refers to the designated security area around the processbuildings and is depicted in FSAR Figure 1.2-1 Plot Plan.CPNPP Basis:NoneNEI 99-01 Basis:This IC addresses the magnitude and extent of FIRES that may be indicative of a potentialdegradation of the level of safety of the plant.EIL-#!The intent of the 15 minute duration is to size the FIRE and to against smallFIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition toalarms, other indc;ations of a FIRE could be a drop in fire main pr essure, automatic activatioof a 6uppression system, etc.Upon receipt, operateos Will take prompt actoRns to- co'Rfirm the validity of an initial fire alarm,'indication, or report. For EAL assessment purposes, the emergency starts at.the time that the initial alarmn, indication, or report was rec~eived, and not the time that akn n,~, + ; fnln +; +; IfArrr Q; 1- InI1 +1,. F; Ar Cl, ,rn !n drr in -,c 4. + 4- -4I a Vtttl'J* VtaI ~*I .A ),lcthe time of receipt of the initial alarm, indication Or reporFt.EIAL-4t2This EAL addresses receipt of a single fire alarmF, and the existence of a FIRE is no~t Verifiedr rnnA rj r rlcý rrnxfpA\ iAcn~n2 oilhjtpC r ri 14c of +kc',nm nc. ranr +nf pnrp+r..rq %.At;! +nLp.promnpt actions to confirmn the validity of a single fire alarm. ForF E=AL assessmnent purposes,th30 _minuteý clock starts at the time that the initial alarm was received, and not the time that asubsequent verific-ation action Was pe~ore~fld.I Page 152 of 276 1 ATTACHMENT 1EAL BasesA single fire alarm, absent other indication(s) of a FIRE, mnay be indicativ~e of equipment failuror sprius activation, and not an actuai PRE~. tr-e 1IS reasGn, auuRIuuai umne is aluoweu toerif, the ,l;4idity of the alaFrm. The 30 minute peFred is a reasonable amou.nt of time todetermie if .an atual FIRE exists; however, after that time, and absent to theGontrar,', it is -assu~med that an actual FIRE is i rgesIf an actual FIRE is erified by a repoFt from the field, then EAL imm ediatel applicable,and- the emergency must be declredAp-- if the- FIRE is not extingui-shed- 1Aithin 15 m-inutes o~f thereport. if the alarm is verified to be due to an equipment failure or a spurious activation, and-this verification occurs- Within 0miue of the receipt of the alaFrm, then this EAL is notapplicable and no emergency declaration is warranted.In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant PROTECTEDAREA not extinguished xvithin 60 minutes may also potentially degrade the love! of plantsafety. This basis extends to a FIRE occUrring within the P-ROTECTEiD AREA of aý IS/Slocated outside the plant P-ROTEC TED9 AREA. [Sentence for plants w.'th anQ 1SFSI obutsd theplant P-rotected Area]E-AL-#4If a FIRE within the plant or ISFSI [forplants with an I-SFS! outSide the plant PDotected, Ar..a]PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency(e.g., a local town Fire Department), then the level of plant safety is potentially degraded. Thedispatch of an offsite firefighting agency to the site requires an emergency declaration only if itis needed to actively support firefighting efforts because the fire is beyond the capability of theFire Brigade to extinguish. Declaration is not necessary if the agency resources are placed onstand-by, or supporting post-extinguishment recovery or investigation actions.Basi Related~~nr Reau'rcrnts afro(mn AooeRnrJdiAppendix R to 10 CFR 50, states n part--.Criterion 3 of Appendix A to this part specifies that "Structures, systems, andiriir uui ^npnr m[ ni ruel n:i II Iii Q:1e' ý-; 1:1 [ me nu- ýu InnL nnn iii nri:i~ Ii 11111 n r-Ti-rt-r I ---J -..- .- .-----with other safety requirements, the probability ard effect of fires When onnsideiRng the effects of fire, those systems ass"oiated with achieving am, ,inrin f ... ..cndtn5 ~ m m iri nr hnet 'ftx ~~~5e.--.-. .----..-.---.--. ..... .. c.....wn .damage to them. can lead to core damage resulting fro. lss of colant through boil o.ff.fire may affect safe shutdcOWRsystems and because the lo9s Of fLuncrtion Ofsystems used to mitigate the consequences of design basis accidents under post frconditions does not per se impact public safety, the need to limnit fire damnage to systemsrequired to achieve and mnaintain safe Shutdown conditions is greater than the need toli.it fire damage to those systems required to .itigate the .cnsequen~esof design_basis accidents.In addition, Appendix R to 10 CPR 50, aFmGog other cnsiderations, the use of 1 haoufire barriers for the encloGsure Of cable and equipment and associated non safetywcircuits; of onePage 153 of 276 ATTACHMENT 1EAL Bases,redundant train (G.2.c;). As used in EAL #2, the 30 to ve,,;' a single alarm i wellwithin this WorSt caSe 1 hour time period.Depending upon the plant mode at the time of the event, escalation of the emergencyclassification level would be via IC CA6 or SA9.CPNPP Basis Reference(s):1. NEI 99-01 HU4Page 154 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:H -Hazards and Other Conditions Affecting Plant Safety5 -Hazardous GasesGaseous release IMPEDING access to equipment necessary fornormal plant operations, cooldown or shutdownEAL:HA5.1 AlertRelease of a toxic, corrosive, asphyxiant or flammable gas into any Table H-2 rooms orareasANDEntry into the room or area is prohibited or IMPEDED (Note 5)Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, thenno emergency classification is warranted.Table H-2 Safe Operation & Shutdown Rooms/AreasRoom/Area Mode ApplicabilityCharging Pump Rooms 1, 2, 3, 4, 5, 6CVCS Valve Rooms 1, 2, 3, 4, 5, 61 E Switchgear Rooms AllRHR Pump Rooms 4, 5, 6Mode Applicability:AllDefinition(s):IMPEDE(D) -Personnel access to a room or area is hindered to an extent that extraordinarymeasures are necessary to facilitate entry of personnel into the affected room/area(e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).CPNPP Basis:If the equipment in the listed room or area was already inoperable, or out-of-service, before theevent occurred, then no emergency should be declared since the event will have no adverseimpact beyond that already allowed by Technical Specifications at the time of the event.The list of plant rooms or areas with entry-related mode applicability identified specify thoserooms or areas that contain equipment which require a manual/local action as specified inoperating procedures used for normal plant operation, cooldown and shutdown. Rooms orareas in which actions of a contingent or emergency nature would be performed (e.g., anaction to address an off-normal or emergency condition such as emergency repairs, correctivemeasures or emergency operations) are not included. In addition, the list specifies the plantmode(s) during which entry would be required for each room or area (ref. 1).I Page 155 of 276 ATTACHMENT 1EAL BasesNEI 99-01 Basis:This IC addresses an event involving a release of a hazardous gas that precludes or impedesaccess to equipment necessary to maintain normal plant operation, or required for a normalplant cooldown and shutdown. This condition represents an actual or potential substantialdegradation of the level of safety of the plant.An Alert declaration is warranted if entry into the affected room/area is, or may be, procedurallyrequired during the plant operating mode in effect at the time of the gaseous release. Theemergency classification is not contingent upon whether entry is actually necessary at the timeof the release.Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the,,nEergencY Emergency Coordinator's judgment that the gas concentration in theaffected room/area is sufficient to preclude or significantly impede procedurally requiredaccess. This judgment may be based on a variety of factors including an existing job hazardanalysis, report of ill effects on personnel, advice from a subject matter expert or operatingexperience with the same or similar hazards. Access should be considered as impeded ifextraordinary measures are necessary to facilitate entry of personnel into the affectedroom/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinelyemployed).An emergency declaration is not warranted if any of the following conditions apply:" The plant is in an operating mode different than the mode specified for the affectedroom/area (i.e., entry is not required during the operating mode in effect at the time of thegaseous release). For example, the plant is in Mode 1 when the gaseous release occurs,and the procedures used for normal operation, cooldown and shutdown do not requireentry into the affected room until Mode 4." The gas release is a planned activity that includes compensatory measures which addressthe temporary inaccessibility of a room or area (e.g., fire suppression system testing)." The action for which room/area entry is required is of an administrative or record keepingnature (e.g., normal rounds or routine inspections).* The access control measures are of a conservative or precautionary nature, and would notactually prevent or impede a required action.An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerouslevels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment.This reduces the concentration of oxygen below the normal level of around 19%, which canlead to breathing difficulties, unconsciousness or even death.This EAL does not apply to firefighting activities that automatically or manually activate a firesuppression system in an area, or to intentional in"rting of etainment.. (BVWR -ony).Escalation of the emergency classification level would be via Recognition Category A R, C or FICs.CPNPP Basis Reference(s):1. Attachment 3 Safe Operation & Shutdown Areas Tables R-3 & H-2 Bases2. NEI 99-01 HA5Page 156 of 276 ATTACHMENT 1EAL BasesI Page 157 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:H -Hazards and Other Conditions Affecting Plant Safety6 -Control Room EvacuationControl Room evacuation resulting in transfer of plant control toalternate locationsEAL:HA6.1 AlertAn event has resulted in plant control being transferred from the Control Room to theRemote Shutdown Panel (RSP)Mode Applicability:AllDefinition(s):NoneCPNPP Basis:Upon evacuation of the Control Room plant control is established at the Remote ShutdownPanel (RSP). ABN-905A/B "Loss of Control Room Habitability" and ABN-803A/B "Response toa Fire in the Control Room or Cable Spreading Room" provide the instructions for tripping theunit, and maintaining RCS inventory and Hot Shutdown conditions from outside the ControlRoom. The Shift Manager (SM) determines if the Control Room is inoperable and requiresevacuation. Control Room inhabitability may be caused by fire, dense smoke, noxious fumes,bomb threat in or adjacent to the Control Room, or other life threatening conditions. (Ref. 1, 2,3, 4, 5).Inability to establish plant control from outside the Control Room escalates this event to a SiteArea Emergency per EAL HS6.1.NEI 99-01 Basis:This IC addresses an evacuation of the Control Room that results in transfer of plant control toalternate locations outside the Control Room. The loss of the ability to control the plant fromthe Control Room is considered to be a potential substantial degradation in the level of plantsafety.Following a Control Room evacuation, control of the plant will be transferred to alternateshutdown locations. The necessity to control a plant shutdown from outside the Control Room,in addition to responding to the event that required the evacuation of the Control Room, willpresent challenges to plant operators and other on-shift personnel. Activation of the ERO andemergency response facilities will assist in respondingto these challenges.Escalation of the emergency classification level would be via IC HS6.Page 158 of 276 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):1. DBD-ME-003 Control Room Habitability2. ABN-905A Loss of Control Room Habitability3. ABN-905B Loss of Control Room Habitability4. ABN-803A Response to a Fire in the Control Room or Cable Spreading Room5. ABN-803B Response to a Fire in the Control Room or Cable Spreading Room6. NEI 99-01 HA6Page 159 of 276 ATTACHMENT 1EAL BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 6 -Control Room EvacuationInitiating Condition: Inability to control a key safety function from outside the Control RoomEAL:HS6.1 Site Area EmergencyAn event has resulted in plant control being transferred from the Control Room to theRemote Shutdown Panel (RSP)ANDControl of any of the following key safety functions is not re-established within 15 min.(Note 1):" Reactivity" Core Cooling" RCS heat removalNote 1: The Emergency Coordinator should declare the event promptly upon determining that.time limit hasbeen exceeded, or will likely be exceeded.Mode Applicability:AllDefinition(s):NoneCPNPP Basis:Upon evacuation of the Control Room plant control is established at the Remote ShutdownPanel (RSP). ABN-905A/B "Loss of Control Room Habitability" and ABN-803A/B "Response toa Fire in the Control Room or Cable Spreading Room" provide the instructions for tripping theunit, and maintaining RCS inventory and Hot Shutdown conditions from outside the ControlRoom. The Shift Manager (SM) determines if the Control Room is inoperable and requiresevacuation. Control Room inhabitability may be caused by fire, dense smoke, noxious fumes,bomb threat in or adjacent to the Control Room, or other life threatening conditions. (Ref. 1, 2,3, 4, 5).The intent of this EAL is to capture events in which control of the plant cannot be reestablishedin a timely manner. The fifteen minute time for transfer starts when the Control Room begins tobe evacuated (not when the ABN is entered). The time interval is based on how quickly controlmust be reestablished without core uncovery and/or core damage. The determination ofwhether or not control is established from outside the Control Room is based on EmergencyCoordinator judgment. The Emergency Coordinator is expected to make a reasonable,informed judgment that control of the plant from outside the Control Room cannot beestablished within the fifteen minute interval.Once the Control Room is evacuated, the objective is to establish control of important plantequipment and maintain knowledge of important plant parameters in a timely manner. PrimaryPage 160 of 276 ATTACHMENT 1EAL Basesemphasis should be placed on components and instruments that supply protection for andinformation about safety functions. Typically, these safety functions are reactivity control(ability to shutdown the reactor and maintain it shutdown), RCS inventory (ability to cool thecore), and secondary heat removal (ability to maintain a heat sink).NEI 99-01 Basis:This IC addresses an evacuation of the Control Room that results in transfer of plant control toalternate locations, and the control of a key safety function cannot be reestablished in a timelymanner. The failure to gain control of a key safety function following a transfer of plant controlto alternate locations is a precursor to a challenge to one or more fission product barrierswithin a relatively short period of time.The determination of whether or not "control" is established at the remote safe shutdownlocation(s) is based on Emergency Directer-Coordinator judgment. The Emergency DiieetOrCoordinator is expected to make a reasonable, informed judgment within (the site ,pecific timefeF1t4ansfeF)15 minutes whether or not the operating staff has control of key safety functionsfrom the remote safe shutdown location(s).Escalation of the emergency classification level would be via IC FG1 or CG1CPNPP Basis Reference(s):1. DBD-ME-003 Control Room Habitability2. ABN-905A Loss of Control Room Habitability3. ABN-905B Loss of Control Room Habitability4. ABN-803A Response to a Fire in the Control Room or Cable Spreading Room5. ABN-803B Response to a Fire in the Control Room or Cable Spreading Room6. NEI 99-01 HS6I Page 161 of 276 ATTACHMENT 1EAL BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 7 -Emergency Coordinator JudgmentInitiating Condition: Other conditions existing that in the judgment of the EmergencyCoordinator warrant declaration of a UEEAL:HU7.1 Unusual EventOther conditions exist which in the judgment of the Emergency Coordinator 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.No releases of radioactive material requiring offsite response or monitoring are expectedunless further degradation of SAFETY SYSTEMS occurs.Mode Applicability:AllDefinition(s):SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/orplacing it in the cold shutdown condition, including the ECCS. These are typically systemsclassified as safety-related (as defined in 10CFR50.2):Those structures, systems and components that are relied upon to remain functional duringand 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 could resultin potential offsite exposures.CPNPP Basis:The Emergency Coordinator is the designated onsite individual having the responsibility andauthority for implementing the CPNPP Radiological Emergency Response Plan. The ShiftManager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions asoutlined in the Emergency Plan implementing procedures. If required by the emergencyclassification or if deemed appropriate by the Emergency Coordinator, emergency responsepersonnel are notified and instructed to report to their emergency response locations. In thismanner, the individual usually in charge of activities in the Control[Room is responsible forinitiating the necessary emergency response, but Plant Management is expected to managethe emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1).NEI 99-01 Basis:This IC addresses unanticipated conditions-not addressed explicitly elsewhere but that warrantdeclaration of an emergency because conditions exist which are believed by the EmergencyPage 162 of 276 ATTACHMENT 1EAL BasesDireeteF-Coordinator to fall under the emergency classification level description for anNQUEUnusual Event.Page 163 of 276 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):1. CPNPP Radiological Emergency Response Plan section 1.1.2 Response2. NEI 99-01 HU7Page 164 of 276 ATTACHMENT 1EAL BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 7 -Emergency Coordinator JudgmentInitiating Condition: Other conditions exist that in the judgment of the EmergencyCoordinator warrant declaration of an AlertEAL:HA7.1 AlertOther conditions exist which, in the judgment of the Emergency Coordinator, 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.Mode Applicability:AllDefinition(s):HOSTILE ACTION -An act toward CPNPP or its personnel that includes the use of violentforce to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end.This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or otherdevices used to deliver destructive force. Other acts that satisfy the overall intent may beincluded. Hostile action should not be construed to include acts of civil disobedience orfelonious acts that are not part of a concerted attack on CPNPP. Non-terrorism-based EALsshould be used to address such activities (i.e., this may include violent acts betweenindividuals in the owner controlled area).CPNPP Basis:The Emergency Coordinator is the designated onsite individual having the responsibility andauthority for implementing the CPNPP Radiological Emergency Response Plan. The ShiftManager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions asoutlined in the Emergency Plan implementing procedures. If required by the emergencyclassification or if deemed appropriate by the Emergency Coordinator, emergency responsepersonnel are notified and instructed to report to their emergency response locations. In thismanner, the individual usually in charge of activities in the Control Room is responsible forinitiating the necessary emergency response, but Plant Management is expected to managethe emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1).NEI 99-01 Basis:This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrantdeclaration of an emergency because conditions exist which are believed by the EmergencyI Direotor-Coordinator to fall under the emergency classification level description for an Alert.I Page 165 of 276 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):1. CPNPP Radiological Emergency Response Plan section 1.1.2 Response2. NEI 99-01 HA7Page 166 of 276 ATTACHMENT 1EAL BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 7 -Emergency Coordinator JudgmentInitiating Condition: Other conditions existing that in the judgment of the EmergencyCoordinator warrant declaration of a Site Area EmergencyEAL:HS7.1 Site Area EmergencyOther conditions exist which in the judgment of the Emergency Coordinator 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 could leadto 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 beyond the EXCLUSION AREABOUNDARYMode Applicability:AllDefinition(s):EXCLUSION AREA BOUNDARY -Exclusion Area Boundary is a synonymous term for SiteBoundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion AreaBoundary. This boundary is used for establishing effluent release limits with respect to therequirements of 10CFR20. See also CPNPP Emergency Plan Appendix E, Complex andOwner Controlled Area and CCNPP ODCM Section 5.0 Design Features.HOSTILE ACTION -An act toward CPNPP or its personnel that includes the use of violentforce to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end.This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or otherdevices used to deliver destructive force. Other acts that satisfy the overall intent may beincluded. Hostile action should not be construed to include acts of civil disobedience orfelonious acts that are not part of a concerted attack on CPNPP. Non-terrorism-based EALsshould be used to address such activities (i.e., this may include violent acts betweenindividuals in the owner controlled area)CPNPP Basis:The Emergency Coordinator is the designated onsite individual having the responsibility andauthority for implementing the CPNPP Radiological Emergency Response Plan. The ShiftManager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions asoutlined in the Emergency Plan implementing procedures. If required by the emergencyclassification or if deemed appropriate by the Emergency Coordinator, emergency responsepersonnel are notified and instructed to report to their emergency response locations. In thismanner, the individual usually in charge of activities in the Control Room is responsible forinitiating the necessary emergency response, but Plant Management is expected to managethe emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1).Page 167 of 276 ATTACHMENT 1EAL BasesNEI 99-01 Basis:This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrantdeclaration of an emergency because conditions exist which are believed by the EmergencyDir-eeter-Coordinator to fall under the emergency classification level description for a Site AreaEmergency.CPNPP Basis Reference(s):1. CPNPP Radiological Emergency Response Plan section 1.1.2 ResponseI 2. NEI 99-01 HS7Page 168 of 276 ATTACHMENT 1EAL BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 7 -Emergency Coordinator JudgmentInitiating Condition: Other conditions exist which in the judgment of the EmergencyCoordinator warrant declaration of a General EmergencyEAL:HG7.1 General EmergencyOther conditions exist which in the judgment of the Emergency Coordinator 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 offsite formore than the immediate site areaMode Applicability:AllDefinition(s):HOSTILE ACTION -An act toward CPNPP or its personnel that includes the use of violentforce to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end.This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or otherdevices used to deliver destructive force. Other acts that satisfy the overall intent may beincluded. Hostile action should not be construed to include acts of civil disobedience orfelonious acts that are not part of a concerted attack on CPNPP. Non-terrorism-based EALsshould be used to address such activities (i.e., this may include violent acts betweenindividuals in the owner controlled area).IMMINENT -The trajectory of events or conditions is such that an EAL will be met within arelatively short period of time regardless of mitigation or corrective actions.CPNPP Basis:The Emergency Coordinator is the designated onsite individual having the responsibility andauthority for implementing the CPNPP Radiological Emergency Response Plan. The ShiftManager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions asoutlined in the Emergency Plan implementing procedures. If required by the emergencyclassification or if deemed appropriate by the Emergency Coordinator, emergency responsepersonnel are notified and instructed to report to their emergency response locations. In thismanner, the individual usually in charge of activities in the Control Room is responsible forinitiating the necessary emergency response, but Plant Management is expected to managethe emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1).Releases can reasonably be expected to exceed EPA PAG plume exposure levels outside theSite Boundary.Page 169 of 276 ATTACHMENT 1EAL BasesNEI 99-01 Basis:This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrantdeclaration of an emergency because conditions exist which are believed by the EmergencyDi-eeter-Coordinator to fall under the emergency classification level description for a GeneralEmergency.CPNPP Basis Reference(s):1. CPNPP Radiological Emergency Response Plan section 1.1.2 Response2. NEI 99-01 HG7L Page 170 of 276 ATTACHMENT 1EAL BasesCategory S -System MalfunctionEAL Group: Hot Conditions (RCS temperature greater than 2001F);EALs in this category are applicable only in one or morehot operating modes.Numerous system-related equipment failure events that warrant emergency classification havebeen identified in this category. They may pose actual or potential threats to plant safety.The events of this category pertain to the following subcategories:1. Loss of Emergency AC PowerLoss of emergency electrical power can compromise plant safety system operabilityincluding decay heat removal and emergency core cooling systems which may benecessary to ensure fission product barrier integrity. This category includes loss of onsiteand offsite sources for 6.9KV AC safeguard buses.2. Loss of Vital DC PowerLoss of emergency electrical power can compromise plant safety system operabilityincluding decay heat removal and emergency core cooling systems which may benecessary to ensure fission product barrier integrity. This category includes loss of vitalplant 125 VDC power sources.3. Loss of Control Room IndicationsCertain events that degrade plant operator ability to effectively assess plant conditionswithin the plant warrant emergency classification. Losses of indicators are in thissubcategory.4. RCS ActivityDuring 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 increase fromthese base-line levels (2% -5% clad failures) is indicative of fuel failures and is coveredunder the Fission Product Barrier Degradation category. However, lesser amounts of claddamage may result in coolant activity exceeding Technical Specification limits. Thesefission products will be circulated with the reactor coolant and can be detected by coolantsampling.5. RCS LeakageThe reactor vessel provides a volume for the coolant that covers the reactor core. Thereactor pressure 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 indicatespotential pipe cracks that may propagate to an extent threatening fuel clad, RCS andcontainment integrity.6. RPS FailureThis subcategory includes events related to failure of the Reactor Protection System (RPS)to initiate and complete reactor trips. In the plant licensing basis, postulated failures of theRPS to complete a reactor trip comprise a specific set of analyzed events referred to asPage 171 of 276 ATTACHMENT IEAL BasesAnticipated Transient Without Scram (ATWS) events. For EAL classification, however,ATWS is intended to mean any trip failure event that does not achieve reactor shutdown. IfRPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk andcould cause a threat to fuel clad, RCS and containment integrity.7. Loss of CommunicationsCertain events that degrade plant operator ability to effectively communicate with essentialpersonnel within or external to the plant warrant emergency classification.8. Containment FailureFailure of containment isolation capability (under conditions in which the containment is notcurrently challenged) warrants emergency classification. Failure of containment pressurecontrol capability also warrants emergency classification.9. Hazardous Event Affectinq Safety SystemsVarious natural and technological events that result in degraded plant safety systemperformance or significant visible damage warrant emergency classification under thissubcategory.Page 172 of 276 1 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:S -System Malfunction1 -Loss of Emergency AC PowerLoss of all offsite AC power capability to safeguard buses for 15minutes or longerEAL:SUI.1 Unusual EventLoss of all offsite AC power capability, Table S-1, to 6.9 KV safeguard buses uEA1 anduEA2 for greater than or equal to 15 min. (Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Table S-1 AC Power SourcesOffsite:* 138 KV switchyard circuit* 345 KV switchyard circuitOnsite:* uEG1" uEG2Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):NoneBasis:CPNPP Basis:For emergency classification purposes, "capability" means that an offsite AC power source(s)is available to the safeguard buses, whether or not the buses are powered from it.The safeguards AC distribution system power sources consist of the preferred and alternateoffsite power sources, and the onsite standby emergency diesel generators uEG1 and uEG2.Offsite power is supplied to the plant switchyards from the transmission network by five 345 KVand two 138 KV transmission lines. From the switchyards, two electrically and physicallyseparated circuits provide AC power through step down startup transformers, to the 6.9 kVsafeguard buses. The 138 kV switchyard circuit is the preferred source for Unit 2 and alternatesource for-Unit 1. The 345 KV circuit is the preferred source for Unit 1 and alternate source forUnit 2. The onsite AC distribution system is divided into redundant trains so that the loss of anyone load group does not prevent the minimum safety functions from being performed. Eachtrain has connections to two offsite power sources and a dedicated diesel generator. Eachoffsite circuit can supply the Unit 1 and Unit 2 6.9 KV safeguard buses. (ref. 1, 2, 3, 4)Page 173 of 276 ATTACHMENT 1EAL BasesThe 15-minute interval was selected as a threshold to exclude transient or momentary powerlosses.NEI 99-01 Basis:This IC addresses a prolonged loss of offsite power. The loss of offsite power sources rendersthe plant more vulnerable to a complete loss of power to AC emergency buses. This conditionrepresents a potential reduction in the level of safety of the plant.For emergency classification purposes, "capability" means that an offsite AC power source(s)is available to the emergency buses, whether or not the buses are powered from it.Fifteen minutes was selected as a threshold to exclude transient or momentary losses ofoffsite power.Escalation of the emergency classification level would be via IC SAI.CPNPP Basis Reference(s):1. FSAR Figure 8.3-12. FSAR Section 8.23. FSAR Section 8.34. Technical Specifications B3.8.15. ABN-601 Response to a 138/345 KV System Malfunction6. ABN-602 Response to a 6900/480V System Malfunction7. NEI 99-01 SUWPage 174 of 276 1 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:S -System Malfunction1 -Loss of Emergency AC PowerLoss of all but one AC power source to safeguard buses for 15minutes or longerEAL:SAI.1 AlertAC power capability, Table S-1, to 6.9 KV safeguard buses uEA1 and uEA2 reduced to asingle power source for greater than or equal to 15 min. (Note 1)ANDAny additional single power source failure will result in loss of all AC power to SAFETYSYSTEMSNote 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Table S-1 AC Power SourcesOffsite:* 138 KV switchyard circuit* 345 KV switchyard circuitOnsite:* uEG1" uEG2Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/orplacing it in the cold shutdown condition, including the ECCS. These are typically systemsclassified as safety-related (as defined in 1OCFR50.2):Those structures, systems and components that are relied upon to remain functional duringand 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.Basis:CPNPP Basis:Page 175 of 276 ATTACHMENT 1EAL BasesFor emergency classification purposes, "capability" means that an offsite AC power source(s)is available to the emergency buses, whether or not the buses are powered from it.The condition indicated by this EAL is the degradation of the offsite and onsite power sourcessuch that any additional single failure would result in a loss of all AC power to the safeguardbuses.The safeguards AC distribution system power sources consist of the preferred and alternateoffsite power sources, and the onsite standby emergency diesel generators uEG1 and uEG2.Offsite power is supplied to the plant switchyards from the transmission network by five 345 KVand two 138 KV transmission lines. From the switchyards, two electrically and physicallyseparated circuits provide AC power through step down startup transformers, to the 6.9 kVsafeguard buses. The 138 kV switchyard circuit is the preferred source for Unit 2 and alternatesource for Unit 1. The 345 KV circuit is the preferred source for Unit 1 and alternate source forUnit 2. The onsite AC distribution system is divided into redundant trains so that the loss of anyone load group does not prevent the minimum safety functions from being performed. Eachtrain has connections to two offsite power sources and a dedicated diesel generator. Eachoffsite circuit can supply the Unit 1 and Unit 2 6.9 KV safeguard buses. (ref. 1, 2, 3, 4).The 15-minute interval was selected as a threshold to exclude transient or momentary powerlosses. If the capability of a second source of emergency bus power is not restored within 15minutes, an Alert is declared under this EAL.This hot condition EAL is equivalent to the cold condition EAL CU2. 1.NEI 99-01 Basis:This IC describes a significant degradation of offsite and onsite AC power sources such thatany additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. Inthis condition, the sole AC power source may be powering one, or more than one, train ofsafety-related equipment. This IC provides an escalation path from IC SUI.An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplyingrequired power to an emergency bus. Some examples of this condition are presented below." A loss of all offsite power with a concurrent failure of all but one emergency powersource (e.g., an onsite diesel generator)." A loss of all offsite power and loss of all emergency power sources (e.g., onsite dieselgenerators) with a single train of emergency buses being back-fed from the unit maingenerator." A loss of emergency power sources (e.g., onsite diesel generators) with a single train ofemergency buses being ba-k-fed from an offsite power source.Fifteen minutes was selected as a threshold to exclude transient or momentary losses ofpower.Escalation of the emergency classification level would be via IC SSI.I Page 176 of 276 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):1. FSAR Figure 8.3-12. FSAR Section 8.23. FSAR Section 8.34. Technical Specifications B3.8.15. ABN-601 Response to a 138/345 KV System Malfunction6. ABN-602 Response to a 6900/480V System Malfunction7. NEI 99-01 SAlI Page 177 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:S -System MalfunctionI -Loss of Emergency AC PowerLoss of all offsite power and all onsite AC power to safeguard busesfor 15 minutes or longerEAL:EAL:SS1.1 Site Area EmergencyLoss of all offsite and all onsite AC power capability, Table S-1, to 6.9 KV safeguard busesuEA1 and uEA2 for greater than or equal to 15 min. (Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Table S-1 AC Power SourcesOffsite:* 138 KV switchyard circuit* 345 KV switchyard circuitOnsite:" uEGI* uEG2Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):NoneCPNPP Basis:For emergency classification purposes, "capability" means that an AC power source isavailable to the safeguard buses, whether or not the buses are powered from it.The safeguards AC distribution system power sources consist of the preferred and alternateoffsite power sources, and the onsite standby emergency diesel generators uEG1 and uEG2.Offsite power is supplied to the plant switchyards from the transmission network by five 345 KVand two 138 KV transmission lines. From the switchyards, two electrically and physicallyseparated circuits provide AC power through step down startup transformers, to the 6.9 kVsafeguard buses. The 138 kV switchyard circuit is the preferred source for Unit 2 and alternatesource for Unit 1. The 345 KV circuit is the preferred source for Unit 1 and alternate source forUnit 2. The onsite AC distribution system is divided into redundant trains so that the loss of anyone load group does not prevent the minimum safety functions from being performed. Eachtrain has connections to two offsite power sources and a dedicated diesel generator. Eachoffsite circuit can supply the Unit I and Unit 2 6.9 KV safeguard buses. (ref. 1, 2, 3, 4).The 15-minute interval was selected as a threshold to exclude transient or momentary powerlosses. The interval begins when both offsite and onsite AC power capability are lost.Page 178 of 276 ATTACHMENT 1EAL BasesNEI 99-01 Basis:This IC addresses a total loss of AC power that compromises the performance of all SAFETYSYSTEMS requiring electric power including those necessary for emergency core cooling,containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.In addition, fission product barrier monitoring capabilities may be degraded under theseconditions. This IC represents a condition that involves actual or likely major failures of plantfunctions needed for the protection of the public.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Escalation of the emergency classification level would be via ICs AG-1-RG1, FG1 or SGI.CPNPP Basis Reference(s):1.2.3.4.5.6.7.FSAR Figure 8.3-1FSAR Section 8.2FSAR Section 8.3Technical Specifications B3.8.1ABN-601 Response to a 138/345 KV System MalfunctionABN-602 Response to a 6900/480V System MalfunctionNEI 99-01 SS1Page 179 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:S -System Malfunction1 -Loss of Emergency AC PowerProlonged loss of all offsite and all onsite AC power to safeguardbusesEAL:SGI.1 General EmergencyLoss of all offsite and all onsite AC power capability, Table S-I, to 6.9 KV safeguard busesuEA1 and uEA2AND EITHER:* Restoration of at least one safeguard bus from a Table S-1 source or APDG in lessthan 4 hours is not likely (Note 1)* CSFST Core Cooling RED Path conditions metNote 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Table S-1 AC Power SourcesOffsite:* 138 KV switchyard circuit* 345 KV switchyard circuitOnsite:* uEG1* uEG2Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):NoneCPNPP Basis:This EAL is indicated by the extended loss of all offsite and onsite AC power capability to 6.9KV safeguard buses uEA1 and uEA2 either for greater then the CPNPP Station Blackout(SBO) coping analysis time (4 hrs.) (ref. 7) or that has resulted in indications of an actual lossof adequate core cooling.Indication of continuing core cooling degradation is manifested by CSFST Core Cooling REDPath conditions being met. (ref. 8).For emergency classification purposes, "capability" means that an AC power source isavailable to the emergency buses, whether or not the buses are powered from it.The safeguards AC distribution system power sources consist of the preferred and alternateoffsite power sources, and the onsite standby emergency diesel generators uEG1 and uEG2.Page 180 of 276 ATTACHMENT 1EAL BasesOffsite power is supplied to the plant switchyards from the transmission network by five 345 KVand two 138 KV transmission lines. From the switchyards, two electrically and physicallyseparated circuits provide AC power through step down startup transformers, to the 6.9 kVsafeguard buses. The 138 kV switchyard circuit is the preferred source for Unit 2 and alternatesource for Unit 1. The 345 KV circuit is the preferred source for Unit 1 and alternate source forUnit 2. The onsite AC distribution system is divided into redundant trains so that the loss of anyone load group does not prevent the minimum safety functions from being performed. Eachtrain has connections to two offsite power sources and a dedicated diesel generator. Eachoffsite circuit can supply the Unit 1 and Unit 2 6.9 KV safeguard buses. (ref. 1, 2, 3, 4).CPNPP has also provided a set of non-safety related Alternate Power Diesel Generators(APDGs) for each unit with the capability to connect to a safeguards bus one at a time toprovide defense-in-depth for safe shutdown of a unit during outages or during extendedduration of an inoperable offsite circuit on occurrence of concurrent loss of offsite power andfailure of EDGs. The APDGs can provide 3450 kVA to provide long term cooling of each unit(ref. 3).Four hours is the station blackout coping time (ref 7).Indication of continuing core cooling degradation must be based on fission product barriermonitoring with particular emphasis on Emergency Coordinator judgment as it relates toimminent loss of fission product barriers and degraded ability to monitor fission productbarriers. Indication of continuing core cooling degradation is manifested by CSFST CoreCooling RED path conditions being met (ref. 8). Critical Safety Function Status Tree (CSFST)Core Cooling-RED path indicates significant core exit superheating and core uncovery. (ref. 3).NEI-9901 Basis:This IC addresses a prolonged loss of all power sources to AC emergency buses. A loss of allAC power compromises the performance of all SAFETY SYSTEMS requiring electric powerincluding those necessary for emergency core cooling, containment heat removal/pressurecontrol, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buseswill lead to a loss of one or more fission product barriers. In addition, fission product barriermonitoring capabilities may be degraded under these conditions.The EAL should require declaration of a General Emergency prior to meeting the thresholdsfor IC FGI. This will allow additional time for implementation of offsite protective actions.Escalation of the emergency classification from Site Area Emergency will occur if it is projectedthat power cannot be restored to at least one AC emergency bus by the end of the analyzedstation blackout coping period. Beyond this time, plant responses and event trajectory aresubject to greater uncertainty, and there is an increased likelihood of challenges to multiplefission product barriers.The estimate for restoring at least one emergency bus should be based on a realistic appraisalof the situation. Mitigation actions with a low probability of success should not be used as abasis for delaying a classification upgrade. The goal is to maximize the time available toprepare for, and implement, protective actions for the public.The EAL will also require a General Emergency declaration if the loss of AC power results inparameters that indicate an inability to adequately remove decay heat from the core.Page 181 of 276 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):1. FSAR Figure 8.3-12. FSAR Section 8.23. FSAR Section 8.34. Technical Specifications B3.8.15. ABN-601 Response to a 138/345 KV System Malfunction6. ABN-602 Response to a 6900/480V System Malfunction7. FSAR Section 8B Station Blackout8. FRC-O.1A/B Response to Inadequate Core Cooling9. NEI 99-01 SG1Page 182 of 276 1 ATTACHMENT 1EAL BasesCategory:Subcategory:S -System Malfunction1 -Loss of Emergency AC PowerInitiating Condition: Loss of all AC and vital DC power sources for 15 minutes or longerEAL:SG1.2 General EmergencyLoss of all offsite and all onsite AC power capability, Table S-1, to 6.9 KV safeguard busesuEA1 and uEA2 for greater than or equal to 15 min.ANDLess than 105 VDC on all 125 VDC safeguard buses uED1, uED2, uED3 and uED4 forgreater than or equal to 15 min.(Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Table S-1 AC Power SourcesOffsite:0 138 KV switchyard circuit* 345 KV switchyard circuitOnsite:" uEG1* uEG2Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):NoneCPNPP Basis:This EAL is indicated by the loss of all offsite and onsite emergency AC power capability to 6.9KV safeguard buses uEA1 and uEA2 for greater than 15 minutes in combination with degradedvital DC power voltage. This EAL addresses operating experience from the March 2011accident at Fukushima Daiichi.For emergency classification purposes, "capability" means that an AC power source isavailable to the emergency buses, whether or not the buses are powered from it.The safeguards AC distribution system power sources consist of the preferred and alternateoffsite power sources, and the onsite standby emergency diesel generators uEG1 and uEG2.Offsite power is supplied to the plant switchyards from the transmission network by five 345 KVand two 138 KV transmission lines. From the switchyards, two electrically and physicallyPage 183 of 276 ATTACHMENT 1EAL Basesseparated circuits provide AC power through step down startup transformers, to the 6.9 kVsafeguard buses. The 138 kV switchyard circuit is the preferred source for Unit 2 and alternatesource for Unit 1. The 345 KV circuit is the preferred source for Unit 1 and alternate source forUnit 2. The onsite AC distribution system is divided into redundant trains so that the loss of anyone load group does not prevent the minimum safety functions from being performed. Eachtrain has connections to two offsite power sources and a dedicated diesel generator. Eachoffsite circuit can supply the Unit 1 and Unit 2 6.9 KV safeguard buses. (ref. 1, 2, 3, 4).The safeguards 125 VDC buses are the Class 1E buses uED1, uED2, uED3 and uED4 (ref. 7,8,9).Each redundant safeguards 125 VDC system consists of two independent batteries eachhaving one main distribution bus, two static battery chargers (one spare), and local distributionpanels. For Unit 1, batteries BT1 ED1 and BT1 ED3 feed all train A load requirements, whilebatteries BT1 ED2 and BT1 ED4 supply train B load requirements.For Unit 2, batteries BT2ED1 and BT2ED3 feed all train A load requirements, while batteriesBT2ED2 and BT2ED4 supply train B load requirements. There are no bus ties or sharing ofpower supplies between redundant trains (ref. 7).Minimum DC bus voltage is 105 VDC (ref. 10). Bus voltage may be monitored from thefollowing indications (ref. 12):Control Room Panel CP-10 Annunciator u--ALB-10B Plant ComputerV-IED1, 125VDC SWITCH PNL lED1 VOLT 1.13 V6501A BATT BT1ED1 VOLTV-1 ED2, 125VDC SWITCH PNL 1ED2 VOLT 2.13 V6502A BATT BT1ED2 VOLTV-I ED3, 125VDC SWITCH PNL IED3 VOLT 1.9 noneV-i ED4, 125VDC SWITCH PNL IED4 VOLT 3.9 V6504A BATT BT1ED4 VOLTNEI-9901 Basis:This IC addresses a concurrent and prolonged loss of both emergency AC and Vital DCpower. A loss of all emergency AC power compromises the performance of all SAFETYSYSTEMS requiring electric power including those necessary for emergency core cooling,containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.A loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS.A sustained loss of both emergency AC and vital DC power will lead to multiple challenges tofission product barriers.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.The 15-minute emergency declaration clock begins at the point when both EAL thresholds aremet.Page 184 of 276 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):1. FSAR Figure 8.3-12. FSAR Section 8.23. PSAR Section 8.34. Technical Specifications B3.8.15. ABN-601 Response to a 138/345 KV System Malfunction6. ABN-602 Response to a 6900/480V System Malfunction7. FSAR 8.3.28. FSAR Figure 8.3-149. PSAR Figure 8.3-14A10. ECA-O.OA/B Loss of All AC Power11. SOP-605A/B 125 VDC Switchgear and Distribution Systems, Batteries and BatteryChargers12.ALM-0102A/B Alarm Procedures Manual, u-ALB-1OB, nos. 1.9, 1.13, 2.13, 3.813.NEI 99-01 SG8Page 185 of 276 ATTACHMENT 1EAL BasesCategory: S -System MalfunctionSubcategory: 2 -Loss of Vital DC PowerInitiating Condition: Loss of all vital DC power for 15 minutes or longerEAL:SS2.1 Site Area EmergencyLess than 105 VDC on all 125 VDC safeguard buses uED1, uED2, uED3 and uED4 forgreater than or equal to 15 min. (Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):NoneCPNPP Basis:The safeguards 125 VDC buses are the Class 1E buses uED1, uED2, uED3 and uED4 (ref. 1,2,3).Each redundant safeguards 125 VDC system consists of two independent batteries eachhaving one main distribution bus, two static battery chargers (one spare), and local distributionpanels. For Unit 1, batteries BT1 ED1 and BT1 ED3 feed all train A load requirements, whilebatteries BT1 ED2 and BT1 ED4 supply train B load requirements.For Unit 2, batteries BT2ED1 and BT2ED3 feed all train A load requirements, while batteriesBT2ED2 and BT2ED4 supply train B load requirements. There are no bus ties or sharing ofpower supplies between redundant trains (ref. 1).Minimum DC bus voltage is 105 VDC (ref. 4). Bus voltage may be monitored from the followingindications (ref. 6):Control Room Panel CP-10 Annunciator u--ALB-1OB Plant ComputerV-IED1, 125VDC SWITCH PNL IED1 VOLT 1.13 V6501A BATT BT1ED1 VOLTV-1ED2, 125VDC SWITCH PNL 1ED2 VOLT 2.13 V6502A BATT BT1ED2 VOLTV-1 ED3, 125VDC SWITCH PNL IED3 VOLT 1.9 V6503A BATT BT1ED3 VOLTV-I ED4, 125VDC SWITCH PNL 1ED4 VOLT 3.9 V6504A BATT BT1ED4 VOLTPage 186 of 276 ATTACHMENT 1EAL BasesNEI 99-01 Basis:This IC addresses a loss of vital DC power which compromises the ability to monitor andcontrol SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a majorfailure of plant functions needed for the protection of the public.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Escalation of the emergency classification level would be via ICs AG-1RGI, FG1 or SG8SGI.CPNPP Basis Reference(s):1. FSAR 8.3.22. FSAR Figure 8.3-143. FSAR Figure 8.3-14A4. ECA-O.OA/B Loss of All AC Power5. SOP-605A/B 125 VDC Switchgear and Distribution Systems, Batteries and BatteryChargers6. ALM-0102A/B Alarm Procedures Manual, u-ALB-1OB, nos. 1.9, 1.13, 2.13, 3.87. NEI 99-01 SS8I Page 187 of 276 1 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:S -System Malfunction3 -Loss of Control Room IndicationsUNPLANNED loss of Control Room indications for 15 minutes orlongerEAL:SU3.1 Unusual EventAn UNPLANNED event results in the inability to monitor one or more Table S-2parameters from within the Control Room for greater than or equal to 15 min. (Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Table S-2 Safety System Parameters" Reactor power" RCS level* RCS pressure" Core Exit T/C temperature" Level in at least one SG" Auxiliary or emergency feed flow inat least one SGMode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):UNPLANNED -A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient. The cause of the parameter changeor event may be known or unknown.CPNPP Basis:SAFETY SYSTEM parameters listed in Table S-2 are monitored in the Control Room througha combination of hard control panel indicators as well as computer based information systems.The Plant Process Computer, which displays SPDS required information, serves as aredundant compensatory indicator which may be utilized in lieu of normal Control Roomindicators (ref. 1, 2, 3, 4).Page 188 of 276 ATTACHMENT 1EAL BasesNEI 99-01 Basis:This IC addresses the difficulty associated with monitoring normal plant conditions without theability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition isa precursor to a more significant event and represents a potential degradation in the level ofsafety of the plant.As used in this EAL, an "inability to monitor" means that values for one or more of the listedparameters cannot be determined from within the Control Room. This situation would requirea loss of all of the Control Room sources for the given parameter(s). For example, the reactorpower level cannot be determined from any analog, digital and recorder source within theControl Room.An event involving a loss of plant indications, annunciators and/or display systems is evaluatedin accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine ifan NRC event report is required. The event would be reported if it significantly impaired thecapability to perform emergency assessments. In particular, emergency assessmentsnecessary to implement abnormal operating procedures, emergency operating procedures,and emergency plan implementing procedures addressing emergency classification, accidentassessment, or protective action decision-making.This EAL is focused on a selected subset of plant parameters associated with the key safetyfunctions of reactivity control, core cooling [PWR] /i RPV level and RCS heat removal.The loss of the ability to determine one or more of these parameters from within the ControlRoom is considered to be more significant than simply a reportable condition. In addition, if allindication sources for one or more of the listed parameters are lost, then the ability todetermine the values of other SAFETY SYSTEM parameters may be impacted as well. Forexample, if the value for reactor vessel level [PWRI / RPV water level [.WR] cannot bedetermined from the indications and recorders on a main control board, the SPDS or the plantcomputer, the availability of other parameter values may be compromised as well.Fifteen minutes was selected as a threshold to exclude transient or momentary losses ofindication.Escalation of the emergency classification level would be via IC SA2SA3.CPNPP Basis Reference(s):1. FSAR Section 7.52. DBD-EE-033 Detailed Control Room Design, 5.1.2, Figure 13. SOP 906 Plant Process Computer System Guidelines4. ABN 906 Plant Process Computer System Malfunction5. NEI 99-01 SU2Page 189 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:EAL:S -System Malfunction3 -Loss of Control Room IndicationsUNPLANNED loss of Control Room indications for 15 minutes orlonger with a significant transient in progressSA3.1 AlertAn UNPLANNED event results in the inability to monitor one or more Table S-2parameters from within the Control Room for greater than or equal to 15 min. (Note 1)ANDAny significant transient is in progress, Table S-3Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Table S-2 Safety System Parameters* Reactor power* RCS level* RCS pressure" Core Exit T/C temperature* Level in at least one SG* Auxiliary or emergency feed flow inat least one SGTable S-3 Significant Transients" Reactor trip* Runback greater than or equal to25% thermal power* Electrical load rejection greater than25% electrical load* ECCS actuationMode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):1JNPLANNED -A parameter change or an event that is not 1) the result of an intendedtolution or 2) an expected plant response to a transient. The cause of the parameter change93vent may be known or unknown.Page 190 of 276 1 ATTACHMENT 1EAL BasesCPNPP Basis:SAFETY SYSTEM parameters listed in Table S-1 are monitored in the Control Room througha combination of hard control panel indicators as well as computer based information systems.The Plant Computer, which displays SPDS required information, serves as a redundantcompensatory indicator which may be utilized in lieu of normal Control Room indicators (ref. 1,2,3,4).Significant transients are listed in Table S-2 and include response to automatic or manuallyinitiated functions such as reactor trips, runbacks involving greater than or equal to 25%thermal power change, electrical load rejections of greater than 25% full electrical load orECCS (SI) injection actuations.NEI 99-01 Basis:This IC addresses the difficulty associated with monitoring rapidly changing plant conditionsduring a transient without the ability to obtain SAFETY SYSTEM parameters from within theControl Room. During this condition, the margin to a potential fission product barrier challengeis reduced. It thus represents a potential substantial degradation in the level of safety of theplant.As used in this EAL, an "inability to monitor" means that values for one or more of the listedparameters cannot be determined from within the Control Room. This situation would requirea loss of all of the Control Room sources for the given parameter(s). For example, the reactorpower level cannot be determined from any analog, digital and recorder source within theControl Room.An event involving a loss of plant indications, annunciators and/or display systems is evaluatedin accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine ifan NRC event report is required. The event would be reported if it significantly impaired thecapability to perform emergency assessments. In particular, emergency assessmentsnecessary to implement abnormal operating procedures, emergency operating procedures,and emergency plan implementing procedures addressing emergency classification, accidentassessment, or protective action decision-making.This EAL is focused on a selected subset of plant parameters associated with the key safetyfunctions of reactivity control, core cooling [PuD4R] / RPV lev.,el [BWR] and RCS heat removal.The loss of the ability to determine one or more of these parameters from within the ControlRoom is considered to be more significant than simply a reportable condition. In addition, if allindication sources for one or more of the listed parameters are lost, then the ability todetermine the values of other SAFETY SYSTEM parameters may be impacted as well. Forexample, if the value for reactor vessel level [PWR] i RPV water level [.1r314RI cannot bedetermined from the indications and recorders on a main control board, the SPDS or the plantcomputer, the availability of other parameter values may be compromised as well.Fifteen minutes was selected as a threshold to exclude transient or momentary losses ofindication.Escalation of the emergency classification level would be via ICs FS1 or IC AS-I-RS1Page 191 of 276 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):1.2.3.4.5.FSAR Section 7.5DBD-EE-033 Detailed Control Room Design, 5.1.2, Figure 1SOP 906 Plant Process Computer System GuidelinesABN 906 Plant Process Computer System MalfunctionNEI 99-01 SA2Page 192 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:EAL:S -System Malfunction4 -RCS ActivityReactor coolant activity greater than Technical Specification allowablelimitsSU4.1 Unusual EventReactor coolant Dose Equivalent 1-131 specific activity greater than 60 pCi/gmORReactor coolant Dose Equivalent XE-133 specific activity greater than 500 pCi/gmMode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):NoneCPNPP Basis:This EAL addresses reactor coolant samples exceeding Technical Specification LCOs3.4.16.A and 3.4.16.B which are applicable in Modes 1, 2, and 3 and 4 (ref. 1). The TechnicalSpecification limits accommodate an iodine spike phenomenon that may occur followingchanges in thermal power. The Technical Specification LCO limits are established to minimizethe offsite radioactivity dose consequences in the event of a steam generator tube rupture(SGTR) accident (ref. 2).NEI 99-01 Basis:This IC addresses a reactor coolant activity value that exceeds an allowable limit specified inTechnical Specifications. This condition is a precursor to a more significant event andrepresents a potential degradation of the level of safety of the plant.Escalation of the emergency classification level would be via ICs FA1 or the RecognitionCategory A-R ICs.CPNPP Basis Reference(s):1. Technical Specifications Section 3.4.162. Technical Specifications Section B3.4.163. NEI 99-01 SU3Page 193 of 276 ATTACHMENT IEAL BasesCategory: S -System MalfunctionSubcategory: 4 -RCS ActivityInitiating Condition: Reactor coolant activity greater than Technical Specification allowablelimitsEAL:SU4.2 Unusual EventGross Failed Fuel Monitor, FFLu60 (u-RE-0406), High Alarm (RED)Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):NoneCPNPP Basis:This EAL addresses reactor coolant letdown line radiation levels sensed by FFLu60 (u-RE-0406) in excess of Technical Specification allowable limits. The High Alarm (RED) setpoint isbased on the Technical Specifications maximum allowable concentration of radioactivity in thereactor coolant (ref. 1, 2, 3). A Geiger-Mueller tube is mounted on the reactor coolant letdownline after the letdown heat exchanger to monitor fission-product activity. Detection of increasedsystem activity may be indicative of failed fuel. The monitor initiates Alert and High alarms inthe Control Room (PC-1 1 and Plant Computer) (ref. 3, 4, 5, 6, 7, 8).FFLu60 (u-RE-0406) has a range of 1 E-2 -1 E+7 pCi/ml.NEI 99-01 Basis:This IC addresses a reactor coolant activity value that exceeds an allowable limit specified inTechnical Specifications. This condition is a precursor to a more significant event andrepresents a potential degradation of the level of safety of the plant.Escalation of the emergency classification level would be via ICs FAI or the RecognitionCategory A-R ICs.CPNPP Basis Reference(s):1. Technical Specifications Section 3.4.162. ALM-3200 Alarm Procedure DRMS, Channel in High Alarm (RED), pg 543. DBD-EE-023 Radiation Monitoring System4. SWEC-NU(S)-174 Radiation Monitor Alarm Concentrations for Failed Fuel Monitors 1-RE-406 & 2-RE-4065. ABN-102 High Coolant Activity6. FSAR Section 11.5.2.7.117. FSAR Table 11.5-18. CHM-1 11 Primary Chemistry Accident Assessment Sampling ProgramPage 194 of 276 ATTACHMENT 1EAL Bases9. DBD-EE-023 Radiation Monitoring System10.NEI 99-01 SU3I Page 195 of 276 ATTACHMENT 1EAL BasesCategory: S -System MalfunctionSubcategory: 5 -RCS LeakageInitiating Condition: RCS leakage for 15 minutes or longerEAL:SU5.1 Unusual EventRCS unidentified or pressure boundary leakage greater than 10 gpm for greater than orequal to 15 min.ORRCS identified leakage greater than 25 gpm for greater than or equal to 15 min.ORUNISOLABLE leakage from the RCS to a location outside containment greater than 25gpm for greater than or equal to 15 min.(Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):NoneCPNPP Basis:RCS leakage outside of the containment that is not considered identified or unidentifiedleakage per Technical Specifications includes leakage via interfacing systems such as RCS tothe Component Cooling Water, or systems that directly see RCS pressure outside containmentsuch as Chemical & Volume Control System, Nuclear Sampling system and Residual HeatRemoval system (when in the shutdown cooling mode) (ref. 3, 6, 8)Isolating letdown is a standard abnormal operating procedure action and may preventunnecessary classification when a non-RCS leakage path, such as a CVCS leak, exists.Unidentified leakage and identified leakage are determined by performance of the RCS waterinventory balance. Pressure boundary leakage would first appear as unidentified leakage andcan only be positively identified by inspection (ref. 1). OPT-303 (ref. 1) is used to ensure RCSleakage is within Technical Specification limits (ref. 2). ABN-103 Attachments 1 and 3 (ref. 3)are used for excessive RCS leakage.Technical Specifications (ref. 4) defines RCS leakage as follows:Identified Leakage:o Leakage such as that from pump seals or valve packing (except reactor coolant pump(RCP) seal water injection or leakoff), that is captured and conducted to collectionsystems or a sump or collecting tankPage 196 of 276 ATTACHMENT 1EAL Baseso Leakage into the Containment atmosphere from sources that are both specificallylocated and known either not to interfere with the operation of leakage detectionsystems or not to be pressure boundary leakage.o Reactor Coolant System leakage through a steam generator to the Secondary System(primary to secondary leakage);" Unidentified Leakage: All leakage (except RCP seal water injection or leakoff) that is notidentified leakage." Pressure Boundary Leakage: Leakage (except primary to secondary leakage) through anon-isolable fault in an RCS component body, pipe wall, or vessel wall.Escalation of this EAL to the Alert level is via Category F, Fission Product Barrier Degradation,EAL FA1.1.NEI 99-01 Basis:This IC addresses RCS leakage which may be a precursor to a more significant event. In thiscase, RCS leakage has been detected and operators, following applicable procedures, havebeen unable to promptly isolate the leak. This condition is considered to be a potentialdegradation of the level of safety of the plant..EAL ..andE.AL.2The first and second EAL conditions are focused on a loss of mass fromthe RCS due to "unidentified leakage", "pressure boundary leakage" or "identified leakage" (asthese leakage types are defined in the plant Technical Specifications). EAL-#3The thirdcondition addresses an RCS mass loss caused by an UNISOLABLE leak through aninterfacing system. These EAL=s-conditions thus apply to leakage into the containment, asecondary-side system (e.g., steam generator tube leakage iO a PWR) or a location outside ofcontainment.The leak rate values for each E-AL-condition were selected because they are usuallyobservable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). EAL4#!-The firstcondition uses a lower value that reflects the greater significance of unidentified or pressureboundary leakage.The release of mass from the RCS due to the as-designed/expected operation of a relief valvedoes not warrant an emergency classification. FGF PWRS-,aAn emergency classification wouldbe required if a mass loss is caused by a relief valve that is not functioning asdesigned/expected (e.g., a relief valve sticks open and the line flow cannot be isolated). FGFBWRs,-a- stuck open Safety Relief Valve (SRV) oe SRV leakage is not on'sidered eitheridentified or unidentified leakage by Technical Specifications and, therefore, is not applicablete th~sFA--AThe 15-minute threshold duration allows sufficient time for prompt operator actions to isolatethe leakage, if possible.Escalation of the emergency classification level would be via ICs of Recognition Category A-Ror F.Page 197 of 276 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):I.2.3.4.5.6.7.8.9.OPT-303 Reactor Coolant System Water InventoryTechnical Specifications 3.4.13ABN-103 Excessive Reactor Coolant LeakageTechnical Specifications 1.1ABN-108 Shutdown Loss of CoolantFSAR 5.2.5.2FSAR 5.2.5.8ECA-1.2 LOCA Outside ContainmentNEI 99-01 SU4Page 198 of 276 ATTACHMENT 1EAL BasesCategory: S -System MalfunctionSubcategory: 6 -RPS FailureInitiating Condition: Automatic or manual trip fails to shut down the reactorEAL:SU6.1 Unusual EventAn automatic trip did not shut down the reactor as indicated by reactor power greater than5% after any RPS setpoint is exceededANDA subsequent automatic trip or manual trip action taken at the reactor control consoles(MCB reactor trip switches or deenergizing uB3 and uB4) is successful in shutting downthe reactor as indicated by reactor power less than or equal to 5% (Note 8)Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidlyinserted into the core, and does not include manually driving in control rods or implementation of boroninjection strategies.Mode Applicability:1 -Power OperationDefinition(s):NoneCPNPP Basis:The first condition of this EAL identifies the need to cease critical reactor operations byactuation of the automatic Reactor Protection System (RPS) trip function. A reactor trip isautomatically initiated by the RPS when certain continuously monitored parameters exceedpredetermined setpoints (ref. 1).Following a successful reactor trip, rapid insertion of the control rods occurs. Nuclear powerpromptly drops to a fraction of the original power level and then decays to a level severaldecades less with a negative startup rate. The reactor power drop continues until reactorpower reaches the point at which the influence of source neutrons on reactor power starts tobe observable. A predictable post-trip response from an automatic reactor trip signal shouldtherefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentationand a lowering of power into the source range. A successful trip has therefore occurred whenthere is sufficient rod insertion from the trip of RPS to bring the reactor power below theimmediate shutdown decay heat level of 5% (ref. 1, 2).For the purposes of emergency classification, successful manual trip actions are thosewhich can be quickly performed from the reactor control console; MCB reactor tripswitches or deenergizing uB3 and uB4. Reactor shutdown achieved by use of other tripactions specified in FR-S.1 Response to Nuclear Power Generation/ATWS (such as manuallyinsert control rods, opening the reactor trip and bypass breakers in the reactor switchgear,tripping the Rod Drive MG sets in the normal switchgear or emergency boration) do notconstitute a successful manual trip (ref. 2).Page 199 of 276 ATTACHMENT 1EAL BasesFollowing any automatic RPS trip signal, E-0.0 (ref. 1) and /FR-S.1 (ref. 2) prescribe insertionof redundant manual trip signals to back up the automatic RPS trip function and ensure reactorshutdown is achieved. Even if the first subsequent manual trip signal inserts all control rods tothe full-in position immediately after the initial failure of the automatic trip, the lowest level ofclassification that must be declared is an Unusual Event (ref. 2).In the event that the operator identifies a reactor trip is imminent and initiates a successfulmanual reactor trip before the automatic RPS trip setpoint is reached, no declaration isrequired. The successful manual trip of the reactor before it reaches its automatic trip setpointor reactor trip signals caused by instrumentation channel failures do not lead to a potentialfission product barrier loss. However, if subsequent manual reactor trip actions fail to reducereactor power to or below 5%, the event escalates to the Alert under EAL SA6.1.If by procedure, operator actions include the initiation of an immediate manual trip followingreceipt of an automatic trip signal and there are no clear indications that the automatic tripfailed (such as a time delay following indications that a trip setpoint was exceeded), it may bedifficult to determine if the reactor was shut down because of automatic trip or manual actions.If a subsequent review of the trip actuation indications reveals that the automatic trip did notcause the reactor to be shut down, then consideration should be given to evaluating the fuelfor potential damage, and the reporting requirements of 50.72 should be considered for thetransient event.NEI 99-01 Basis:This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor(trip [PrI, / Scram [81/14R]) that results in a reactor shutdown, and either a subsequentoperator manual action taken at the reactor control consoles or an automatic (trip fLrWR]--.scram r[{4R4) is successful in shutting down the reactor. This event is a precursor to a moresignificant condition and thus represents a potential degradation of the level of safety of theplant.I Following the failure on an automatic reactor (trip [PWR] I scram [BI',1.), operators willpromptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g.,I initiate a manual reactor (trip [P,44R] / sc,,ram [,WR.)). If these manual actions are successfulin shutting down the reactor, core heat generation will quickly fall to a level within thecapabilities of the plant's decay heat removal systems.If an initial manual reactor (trip [P-KR], / Gscam [,WRI) is unsuccessful, operators will promptlytake manual action at another location(s) on the reactor control consoles to shutdown thereactor (e.g., initiate a manual reactor (trip [PrW1RI / Scram [B"l'R])) using a different switch).Depending upon several factors, the initial or subsequent effort to manually (trip / sGrim[BW-R}) the reactor, or a concurrent plant condition, may lead to the generation of an automaticreactor (trip [PWR] / scram signal. If a subsequent manual or automatic (trip .PW.. ,sra is successful in shutting down the reactor, core heat generation will quickly fallto a level within the capabilities of the plant's decay heat removal systems.A manual action at the reactor control consoles is any operator action, or set of actions, whichcauses the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor(trip[PWR] / scram r],,. This action does not include manually driving in control rods orimplementation of boron injection strategies. Actions taken at back-panels or other locationsPage 200 of 276 ATTACHMENT 1EAL Baseswithin the Control Room, or any location outside the Control Room, are not considered to be"at the reactor control consoles".Taking the RoactoFr. Mode Sw~itcrh to61 SHT-DOWN is considered to be a mianual scram tion-.[BWR]The plant response to the failure of an automatic or manual reactor (trip [P,.R] / cram. [BVrR])will vary based upon several factors including the reactor power level prior to the event,availability of the condenser, performance of mitigation equipment and actions, otherconcurrent plant conditions, etc. If subsequent operator manual actions taken at the reactorcontrol consoles are also unsuccessful in shutting down the reactor, then the emergencyclassification level will escalate to an Alert via IC SA5SA6. Depending upon the plantresponse, escalation is also possible via IC FA1. Absent the plant conditions needed to meeteither IC SA5-SA6 or FA1, an Unusual Event declaration is appropriate for this event.A reactor shutdown is determined in accordance with applicable Emergency OperatingProcedure criteria.Should a reactor (trip [PV'RJ / [.. "R]) signal be generated as a result of plant work(e.g., RPS setpoint testing), the following classification guidance should be applied.* If the signal causes a plant transient that should have included an automatic reactor (trip[PWVR] / scram [BWR]) and the RPS fails to automatically shutdown the reactor, thenthis IC and the EALs are applicable, and should be evaluated.* If the signal does not cause a plant transient and the (trip [PWRI / scrGam [B,-R]) failureis determined through other means (e.g., assessment of test results), then this IC andthe EALs are not applicable and no classification is warranted.CPNPP Basis Reference(s):1. EOP-O.OA/B Reactor Trip or Safety Injection2. FR-S.1 Response to Nuclear Power Generation/ATWS3 NEI 99-01 SU5Page 201 of 276 ATTACHMENT IEAL BasesCategory: S -System MalfunctionSubcategory: 6 -RPS FailureInitiating Condition: Automatic or manual trip fails to shut down the reactorEAL:SU6.2 Unusual EventA manual trip did not shut down the reactor as indicated by reactor power greater than 5%after any manual trip action was initiatedANDA subsequent automatic trip or manual trip action taken at the reactor control console(MCB reactor trip switches or deenergizing uB3 and uB4) is successful in shutting downthe reactor as indicated by reactor power less than or equal to 5% (Note 8)Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidlyinserted into the core, and does not include manually driving in control rods or implementation of boroninjection strategies.Mode Applicability:1 -Power OperationDefinition(s):NoneCPNPP Basis:This EAL addresses a failure of a manually initiated trip in the absence of having exceeded anautomatic RTS trip setpoint and a subsequent automatic or manual trip is successful inshutting down the reactor (reactor power less than or equal to 5%). (ref. 1).Following a successful reactor trip, rapid insertion of the control rods occurs. Nuclear powerpromptly drops to a fraction of the original power level and then decays to a level severaldecades less with a negative startup rate. The reactor power drop continues until reactorpower reaches the point at which the influence of source neutrons on reactor power starts tobe observable. A predictable post-trip response from an automatic reactor trip signal shouldtherefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentationand a lowering of power into the source range. A successful trip has therefore occurred whenthere is sufficient rod insertion from the trip of RPS to bring the reactor power below theimmediate shutdown decay heat level of 5% (ref. 1, 2).For the purposes of emergency classification, successful manual trip actions are thosewhich can be quickly performed from the reactor control console; MCB reactor tripswitches or deenergizing uB3 and uB4. Reactor shutdown achieved by use of other tripactions specified in FR-S.1 Response to Nuclear Power Generation/ATWS (such as manuallyinsert control rods, opening the reactor trip and bypass breakers in the reactor switchgear,tripping the Rod Drive MG sets in the normal switchgear or emergency boration) do notconstitute a successful manual trip (ref. 2).Page 202 of 276 ATTACHMENT 1EAL BasesFollowing the failure of any manual trip signal, E-0.0 (ref. 1) and FR-S.1 (ref. 2) prescribeinsertion of redundant manual trip signals to back up the RPS trip function and ensure reactorshutdown is achieved. Even if a subsequent automatic trip signal or the first subsequentmanual trip signal inserts all control rods to the full-in position immediately after the initialfailure of the manual trip, the lowest level of classification that must be declared is an UnusualEvent (ref. 2).If both subsequent automatic and subsequent manual reactor trip actions in the Control Roomfail to reduce reactor power below the power associated with the safety system design (lessthan or equal to 5%) following a failure of an initial manual trip, the event escalates to an Alertunder EAL SA6.1.NEI 99-01 Basis:This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor(trip [PWR] / scram [BAR]) that results in a reactor shutdown, and either a subsequentoperator manual action taken at the reactor control consoles or an automatic (trip [PWk]-...am- is successful in shutting down the reactor. This event is a precursor to a moresignificant condition and thus represents a potential degradation of the level of safety of theplant.I Following the failure on an automatic reactor (trip [P,,rIq i scram operators willpromptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g.,I initiate a manual reactor (trip [PR] i scram [,WR])). If these manual actions are successfulin shutting down the reactor, core heat generation will quickly fall to a level within thecapabilities of the plant's decay heat removal systems.If an initial manual reactor (trip [PrK, R, scram [B",,q,) is unsuccessful, operators will promptlytake manual action at another location(s) on the reactor control consoles to shutdown thereactor (e.g., initiate a manual reactor (trip [Pr], / scram [Bl-r)) using a different switch).Depending upon several factors, the initial or subsequent effort to manually [P,/] / scram[BW-R) the reactor, or a concurrent plant condition, may lead to the generation of an automaticreactor (trip [PKR] i scram- [BIA'J) signal. If a subsequent manual or automatic (trip fP...scam is successful in shutting down the reactor, core heat generation will quickly fallto a level within the capabilities of the plant's decay heat removal systems.A manual action at the reactor control consoles is any operator action, or set of actions, whichcauses the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor#trip[PW'R]K ScGram ). This action does not include manually driving in control rods orimplementation of boron injection strategies. Actions taken at back-panels or other locationswithin the Control Room, or any location outside the Control Room, are not considered to be"at the reactor control consoles".Taking the Roactor Mode Switch to SHUTDOWN is considered to be a manual scram action-.fB9WRIThe plant response to the failure of an automatic or manual reactor (trip [PWR' / scram [BWR])will vary based upon several factors including the reactor power level prior to the event,availability of the condenser, performance of mitigation equipment and actions, otherconcurrent plant conditions, etc. If subsequent operator manual actions taken at the reactorcontrol consoles are also unsuccessful in shutting down the reactor, then the emergencyclassification level will escalate to an Alert via IC SA5SA6. Depending upon the plantPage 203 of 276 1 ATTACHMENT 1EAL Basesresponse, escalation is also possible via [C FAI. Absent the plant conditions needed to meeteither IC SA5-SA6 or FA1, an Unusual Event declaration is appropriate for this event.A reactor shutdown is determined in accordance with applicable Emergency OperatingProcedure criteria.I Should a reactor (trip [P"WR] ! S...m [B.. R]) signal be generated as a result of plant work(e.g., RPS setpoint testing), the following classification guidance should be applied.* If the signal causes a plant transient that should have included an automatic reactor (trip[PWR] [B,,-R]) and the RPS fails to automatically shutdown the reactor, thenthis IC and the EALs are applicable, and should be evaluated.e If the signal does not cause a plant transient and the (trip [PWR] / [BWR]) failureis determined through other means (e.g., assessment of test results), then this IC andthe EALs are not applicable and no classification is warranted.CPNPP Basis Reference(s):1.2.3.EOP-O.OA/B Reactor Trip or Safety InjectionFR-S.1 Response to Nuclear Power Generation/ATWSNEI 99-01 SU5Page 204 of 276 ATTACHMENT IEAL BasesCategory:Subcategory:Initiating Condition:S -System Malfunction2 -RPS FailureAutomatic or manual trip fails to shut down the reactor and subsequentmanual actions taken at the reactor control consoles are not successfulin shutting down the reactorEAL:SA6.1 AlertAn automatic or manual trip fails to shut down the reactor as indicated by reactor powergreater than 5%ANDManual trip actions taken at the reactor control console (MCB reactor trip switches ordeenergizing uB3 and uB4) are not successful in shutting down the reactor as indicated byreactor power greater than 5% (Note 8)Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidlyinserted into the core, and does not include manually driving in control rods or implementation of boroninjection strategies.Mode Applicability:1 -Power OperationDefinition(s):NoneCPNPP Basis:This EAL addresses any automatic or manual reactor trip signal that fails to shut down thereactor (reactor power less than or equal to 5%) followed by a subsequent manual trip that failsto shut down the reactor to an extent the reactor is producing energy in excess of the heat loadfor which the safety systems were designed (ref. 1).For the purposes of emergency classification, successful manual trip actions are thosewhich can be quickly performed from the reactor control console; MCB reactor tripswitches or deenergizing uB3 and uB4. Reactor shutdown achieved by use of other tripactions specified in FR-S.1 Response to Nuclear Power Generation/ATWS (such as manuallyinsert control rods, opening the reactor trip and bypass breakers in the reactor switchgear,tripping the Rod Drive MG sets in the normal switchgear or emergency boration) do notconstitute a successful manual trip (ref. 2).5% rated power is a minimum reading on the power range scale that indicates continuedpower production. It also approximates the decay heat which the shutdown systems weredesigned to remove and is indicative of a condition requiring immediate response to preventsubsequent core damage. Below 5%, plant response will be similar to that observed during anormal shutdown. Nuclear instrumentation can be used to determine if reactor power is greaterthan 5 % power (ref. 1, 2).Page 205 of 276 ATTACHMENT IEAL BasesEscalation of this event to a Site Area Emergency would be under EAL SS6.1 or EmergencyCoordinator judgment.NEI 99-01 Basis:This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor(trip [PWR] / Scram. [,LR]) that results in a reactor shutdown, and subsequent operatormanual actions taken at the reactor control consoles to shutdown the reactor are alsounsuccessful. This condition represents an actual or potential substantial degradation of thelevel of safety of the plant. An emergency declaration is required even if the reactor issubsequently shutdown by an action taken away from the reactor control consoles since thisevent entails a significant failure of the RPS.A manual action at the reactor control console is any operator action, or set of actions, whichcauses the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor(trip[PWR] / scram [9B4WR)). This action does not include manually driving in control rods orimplementation of boron injection strategies. If this action(s) is unsuccessful, operators wouldimmediately pursue additional manual actions at locations away from the reactor controlconsoles (e.g., locally opening breakers). Actions taken at back-panels or other locationswithin the Control Room, or any location outside the Control Room, are not considered to be"at the reactor control consoles".Taking the Mode SWitch to SHUTDOWN i,considered to be a manual scram action. [BWR]The plant response to the failure of an automatic or manual reactor (trip [PVVRD / scram [BrI/)will vary based upon several factors including the reactor power level prior to the event,availability of the condenser, performance of mitigation equipment and actions, otherconcurrent plant conditions, etc. If the failure to shut-down the reactor is prolonged enough tocause a challenge to the core cooling [PWR] / PV ..ate. level [BWR] or RCS heat removalsafety functions, the emergency classification level will escalate to a Site Area Emergency viaIC SS65. Depending upon plant responses and symptoms, escalation is also possible via ICFSI. Absent the plant conditions needed to meet either IC SS65 or FS1, an Alert declarationis appropriate for this event.It is recognized that plant responses or symptoms may also require an Alert declaration inaccordance with the Recognition Category F ICs; however, this IC and EAL are included toensure a timely emergency declaration.A reactor shutdown is determined in accordance with applicable Emergency OperatingProcedure criteria.CPNPP Basis Reference(s):1. EOP-O.OA/B Reactor Trip or Safety Injection2. FR-S.1 Response to Nuclear Power Generation/ATWS3. NEI 99-01 SA5Page 206 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:EAL:S -System Malfunction2 -RPS FailureInability to shut down the reactor causing a challenge to core cooling orRCS heat removalSS6.1 Site Area EmergencyAn automatic or manual trip fails to shut down the reactor as indicated by reactor powergreater than 5%ANDAll actions to shut down the reactor are not successful as indicated by reactor powergreater than 5%AND EITHER:* CSFST Core Cooling RED Path conditions met" CSFST Heat Sink RED Path conditions metMode Applicability:1 -Power OperationDefinition(s):NoneCPNPP Basis:This EAL addresses the following:* Any automatic reactor trip signal followed by a manual trip that fails to shut down thereactor to an extent the reactor is producing energy in excess of the heat load for whichthe safety systems were designed (EAL SA6.1), and" Indications that either core cooling is extremely challenged or heat removal is extremelychallenged.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 direct threatto the Fuel Clad and RCS barriers.Reactor shutdown achieved by use of FR-S.1 Response to Nuclear Power Generation/ATWS(such as manually insert control rods, opening the reactor trip and bypass breakers in thereactor switchgear, tripping the Rod Drive MG sets in the normal switchgear or emergencyboration) are also credited as a successful manual trip provided reactor power can be reducedbelow 5% before indications of an extreme challenge to either core cooling or heat removalexist (ref. 1, 2).5% rated power is a minimum reading on the power range scale that indicates continuedpower production. It also approximates the decay heat which the shutdown systems weredesigned to remove and is indicative of a condition requiring immediate response to preventPage 207 of 276 ATTACHMENT 1EAL Basessubsequent core damage. Below 5%, plant response will be similar to that observed during anormal shutdown. Nuclear instrumentation can be used to determine if reactor power is greaterthan 5% power (ref. 1, 2).Indication of continuing core cooling degradation is manifested by CSFST Core Cooling REDPath conditions being met. Specifically, Core Cooling RED Path conditions exist if either coreexit T/Cs are reading greater than or equal to 12001F (ref. 3).Indication of inability to adequately remove heat from the RCS is manifested by CSFST HeatSink RED Path conditions being met. Specifically, Heat Sink RED Path conditions exist ifnarrow range level in at least one steam generator is not greater than or equal to (43[50]%ACC) on Unit 1 or (10 [18]% ACC) on Unit 2 and total feedwater flow to the steam generatorsis less than or equal to 460 gpm (ref. 4).NEI 99-01 Basis:This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor(trip [P] ,'/ .cram [B,,R) that results in a reactor shutdown, all subsequent operator actionsto manually shutdown the reactor are unsuccessful, and continued power generation ischallenging the capability to adequately remove heat from the core and/or the RCS. Thiscondition will lead to fuel damage if additional mitigation actions are unsuccessful and thuswarrants the declaration of a Site Area Emergency.In some instances, the emergency classification resulting from this IC/EAL may be higher thanthat resulting from an assessment of the plant responses and symptoms against theRecognition Category F ICs/EALs. This is appropriate in that the Recognition Category FICs/EALs do not address the additional threat posed by a failure to shut-down the reactor. Theinclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency inresponse to prolonged failure to shutdown the reactor.A reactor shutdown is determined in accordance with applicable Emergency OperatingProcedure criteria.Escalation of the emergency classification level would be via IC AG--RG1 or FG1.CPNPP Basis Reference(s):1. EOP-0.OA/B Reactor Trip or Safety Injection2. FR-S.1 Response to Nuclear Power Generation/ATWS3. FR-C.1 Response to Inadequate Core Cooling4. FR-H.1 Response to Loss of Heat Sink5. NEI 99-01 SS5.I Page 208 of 276 ATTACHMENT 1EAL BasesCategory: S -System MalfunctionSubcategory: 7 -Loss of CommunicationsInitiating Condition: Loss of all onsite or offsite communications capabilitiesEAL:SU7.1 Unusual EventLoss of all Table S-4 onsite communication methodsORLoss of all Table S-4 offsite communication methodsORLoss of all Table S-4 NRC communication methodsTable S-4 Communication MethodsSystem Onsite Offsite NRCGai-tronics Page/Party (PA) XPlant Radios XPABX X X XPublic Telephone X X XFederal Telephone System (FTS) X XMode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):NoneCPNPP Basis:Onsite/offsite communications include one or more of the systems listed in Table S-4 (ref. 1,2).This EAL is the hot condition equivalent of the cold condition EAL CU5.1.NEI 99-01 Basis:This IC addresses a significant loss of on-site or offsite communications capabilities. While nota direct challenge to plant or personnel safety, this event warrants prompt notifications toOROs and the NRC.Page 209 of 276 ATTACHMENT 1EAL BasesThis IC should be assessed only when extraordinary means are being utilized to makecommunications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent tooffsite locations, etc.).EAL#4-The first EAL condition addresses a total loss of the communications methods used insupport of routine plant operations.EAL-#2The second EAL condition addresses a total loss of the communications methods usedto notify all OROs of an emergency declaration. The offsite (OROs. referred to here are-(seeDeveloper Notes) the State Department of Public Safety, Somervell and Hood County EOCs-EAL4#3The third EAL addresses a total loss of the communications methods used to notify theNRC of an emergency declaration.CPNPP Basis Reference(s):1. FSAR 9.5.22. DBD-EE-048 Communication System3. NEI 99-01 SU6I Page 210 of 276 ATTACHMENT 1EAL BasesCategory: S -System MalfunctionSubcategory: 8 -Containment FailureInitiating Condition: Failure to isolate containment or loss of containment pressure control.EAL:SU8.1 Unusual EventAny penetration is not isolated within 15 min. of a VALID containment isolation signalORContainment pressure greater than 18 psig with neither Containment Spray systemoperating per design for greater than or equal to 15 min.(Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Mode Applicability:I -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):VALID -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) by directobservation by plant personnel, such that doubt related to the indicator's operability, thecondition's existence, or the report's accuracy is removed. Implicit in this definition is the needfor timely assessment.CPNPP Basis:The Containment Spray System (CSS) is designed to remove heat from the Containmentenvironment following a LOCA, a main steam line break accident, or a feedwater line breakaccident. Each unit of the CPNPP is equipped with two redundant Containment spray trains,each designed to provide emergency Containment heat removal in the event of a LOCA. Thissystem, in conjunction with the ECCS, removes post-accident thermal energy from theContainment environment, thereby reducing the Containment pressure and temperature. Eachtrain includes two containment spray pumps, spray headers, nozzles, valves, and piping. Eachtrain is powered from a separate safeguard bus. (ref. 1)The Containment pressure setpoint (18 psig, ref. 2) is the pressure at which the ContainmentSpray System should actuate and begin performing its function. The design basis accident.analyses and evaluations assume the loss of one Containment Spray System train (ref. 1).NEI 99-01 Basis:This IC,-EAL addresses a failure of one or more containment penetrations to automaticallyisolate (close) when required by an actuation signal. It also addresses an event that results inhigh containment pressure with a concurrent failure of containment pressure control systems.Absent challenges to another fission product barrier, either condition represents potentialdegradation of the level of safety of the plant.Page 211 of 276 ATTACHMENT 1EAL BasesI For EAL--#-the first condition, the containment isolation signal must be generated as the resulton an off-normal/accident condition (e.g., a safety injection or high containment pressure); afailure resulting from testing or maintenance does not warrant classification. Thedetermination of containment and penetration status -isolated or not isolated -should bemade in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The15-minute criterion is included to allow operators time to manually isolate the requiredpenetrations, if possible.EAL #2The second condition addresses a condition where containment pressure is greaterthan the setpoint at which containment energy (heat) removal systems are designed toautomatically actuate, and less than one full train of equipment is capable of operating perdesign. The 15-minute criterion is included to allow operators time to manually start equipmentthat may not have automatically started, if possible. The inability to start the requiredequipment indicates that containment heat removal/depressurization systems (e.g.,containment sprays or ice condenser fans) are either lost or performing in a degraded manner.This event would escalate to a Site Area Emergency in accordance with IC FS1 if there were aconcurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers.CPNPP Basis Reference(s):1. FSAR Section 6.2.22. FRC-Z.1A/B Response to High Containment Pressure3. NEI 99-01 SU7IPage 212 of 276 ATTACHMENT IEAL BasesCategory: S -System MalfunctionSubcategory: 9 -Hazardous Event Affecting Safety SystemsInitiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the currentoperating modeEAL:SA9.1 AlertThe occurrence of any Table S-5 hazardous eventAND EITHER:* Event damage has caused indications of degraded performance in at least one trainof a SAFETY SYSTEM needed for the current operating mode" The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component orstructure needed for the current operating modeTable S-5 Hazardous Events" Seismic event (earthquake)" Internal or external FLOODING event* High winds or tornado strike" FIRE" EXPLOSION" Other events with similar hazard characteristicsas determined by the Emergency CoordinatorMode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):EXPLOSION -A rapid, violent and catastrophic failure of a piece of equipment due tocombustion, chemical reaction or overpressurization. A release of steam (from high energylines or components) or an electrical component failure (caused by short circuits, grounding,arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drivebelts or overheated electrical equipment do not constitute fires. Observation of flame ispreferred but is NOT required if large quantities of smoke and heat are observed.FLOODING -A condition where water is entering a room or area faster than installedequipment is capable of removal, resulting in a rise of water level within the room or area.SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/orPage 213 of 276 ATTACHMENT 1EAL Basesplacing it in the cold shutdown condition, including the ECCS. These are typically systemsclassified as safety-related (as defined in 10CFR50.2):Those structures, systems and components that are relied upon to remain functional duringand 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.VISIBLE DAMAGE -Damage to a component or structure that is readily observable withoutmeasurements, testing, or analysis. The visual impact of the damage is sufficient to causeconcern regarding the operability or reliability of the affected component or structure.CPNPP Basis:* The significance of seismic events are discussed under EAL HU2.1 (ref. 1)." Internal FLOODING may be caused by events such as component failures, equipmentmisalignment, or outage activity mishaps (ref. 2)." External flooding may be due to high lake level (ref. 3, 4).* Seismic Category I structures are analyzed to withstand a sustained, design wind velocityof at least 80 mph. (ref. 5)." Areas containing functions and systems required for safe shutdown of the plant areidentified by fire area (ref. 6, 7)." An explosion that degrades the performance of a SAFETY SYSTEM train or visiblydamages a SAFETY SYSTEM component or structure would be classified under this EAL.NEI 99-01 Basis:This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or astructure containing SAFETY SYSTEM components, needed for the current operating mode.This condition significantly reduces the margin to a loss or potential loss of a fission productbarrier, and therefore represents an actual or potential substantial degradation of the level ofsafety of the plant.EAL 1.b.!The first condition addresses damage to a SAFETY SYSTEM train that is inservice/operation since indications for it will be readily available. The indications of degradedperformance should be significant enough to cause concern regarding the operability orreliability of the SAFETY SYSTEM train.EAL .b.2The second condition addresses damage to a SAFETY SYSTEM component that isnot in service/operation or readily apparent through indications alone, or to a structurePage 214 of 276 ATTACHMENT 1EAL Basescontaining SAFETY SYSTEM components. Operators will make this determination based onthe totality of available event and damage report information. This is intended to be a briefassessment not requiring lengthy analysis or quantification of the damage.Escalation of the emergency classification level would be via IC FS1 or A&4-RS1.CPNPP Basis Reference(s):1. ABN-907 Acts of Nature2. CPNPP PRA Accident Sequence Analysis "Internal Flooding Sequences"3. FSAR Section 2.4.3.7 Flood Evaluations for Safe Shutdown Impoundment4. DBD-CS-071 Maximum Probable Flood5. FSAR Section 3.3.1.1 Wind Loadings6. CPNPP Fire Protection Report, Section 5.0 "Fire Safe Shutdown Equipment List"7. FSAR Section 7.4 Systems Required for Safe Shutdown8. NEI 99-01 SA9I Page 215 of 276 ATTACHMENT 1EAL BasesCategory F -Fission Product Barrier DegradationEAL Group: Hot Conditions (RCS temperature greater than2000F); 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 that precludesthe release of highly radioactive fission products to the environment. This concept relies onmultiple physical barriers any one of which, if maintained intact, precludes the release ofsignificant amounts of radioactive fission products to the environment. The primary fissionproduct barriers are:A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains thefuel pellets.B. Reactor Coolant System (RCS): The RCS Barrier includes the RCS primary side and itsconnections up to and including the pressurizer safety and relief valves, and otherconnections up to and including the primary isolation valves.C. Containment (CNTMT): The Containment Barrier includes the containment building andconnections up to and including the outermost containment isolation valves. This barrieralso includes the main steam, feedwater, and blowdown line extensions outside thecontainment building up to and including the outermost secondary side isolation valve.Containment Barrier thresholds are used as criteria for escalation of the ECL from Alertto a Site Area Emergency or a General Emergency.The EALs in this category require evaluation of the loss and potential loss thresholds listed inthe fission product barrier matrix of Table F-1 (Attachment 2). "Loss" and "Potential Loss"signify the relative damage and threat of damage to the barrier. "Loss" means the barrier nolonger assures containment of radioactive materials. "Potential Loss" means integrity of thebarrier is threatened and could be lost if conditions continue to degrade. The number ofbarriers that are lost or potentially lost and the following criteria determine the appropriateemergency classification level: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 barrierThe logic used for emergency classification based on fission product barrier monitoring shouldreflect the following considerations:* The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than theContainment Barrier.* Unusual Event ICs associated with RCS and Fuel Clad Barriers are addressed underSystem Malfunction ICs.Page 216 of 276 ATTACHMENT IEAL Bases* For accident conditions involving a radiological release, evaluation of the fission productbarrier thresholds will need to be performed in conjunction with dose assessments toensure correct and timely escalation of the emergency classification. For example, anevaluation of the fission product barrier thresholds may result in a Site Area Emergencyclassification while a dose assessment may indicate that an EAL for GeneralEmergency IC RG1 has been exceeded.* The fission product barrier thresholds specified within a scheme reflect plant-specificCPNPP design and operating characteristics." As used in this category, the term RCS leakage encompasses not just those typesdefined in Technical Specifications but also includes the loss of RCS mass to anylocation- inside the primary containment, an interfacing system, or outside of theprimary containment. The release of liquid or steam mass from the RCS due to the as-designed/expected operation of a relief valve is not considered to be RCS leakage." At the Site Area Emergency level, EAL users should maintain cognizance of how farpresent conditions are from meeting a threshold that would require a GeneralEmergency declaration. For example, if the Fuel Clad and RCS fission product barrierswere both lost, then there should be frequent assessments of containment radioactiveinventory and integrity. Alternatively, if both the Fuel Clad and RCS fission productbarriers were potentially lost, the Emergency Coordinator would have more assurancethat there was no immediate need to escalate to a General Emergency.I Page 217 of 276 ATTACHMENT 1EAL BasesCategory:Fission Product Barrier DegradationSubcategory: N/AInitiating Condition: Any loss or any potential loss of either Fuel Clad or RCSEAL:FAI.1 AlertAny loss or any potential loss of either Fuel Clad or RCS (Table F-I)Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):NoneCPNPP Basis:Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 (Attachment2) lists the fission product barrier thresholds, bases and references.At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavily than theContainment barrier. Unlike the Containment barrier, loss or potential loss of either the FuelClad or RCS barrier may result in the relocation of radioactive materials or degradation of corecooling capability. Note that the loss or potential loss of Containment barrier in combinationwith loss or potential loss of either Fuel Clad or RCS barrier results in declaration of a SiteArea Emergency under EAL FSI.1NEI 99-01 Basis:NoneCPNPP Basis Reference(s):1. NEI 99-01 FA1I Page 218 of 276 ATTACHMENT 1EAL 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 barriers (Table F-I)Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):NoneCPNPP Basis:Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 (Attachment2) lists the fission product barrier thresholds, bases and references.At the Site Area Emergency classification level, each barrier is weighted equally. A Site AreaEmergency 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 the proximityof present conditions with respect to the threshold for a General Emergency is important. Forexample, the existence of Fuel Clad and RCS Barrier loss thresholds in addition to offsite doseassessments would require continual assessments of radioactive inventory and Containmentintegrity in anticipation of reaching a General Emergency classification. Alternatively, if bothFuel Clad and RCS potential loss thresholds existed, the Emergency Coordinator would havegreater assurance that escalation to a General Emergency is less imminent.NEI 99-01 Basis:NoneCPNPP Basis Reference(s):1. NEI 99-01 FS1Page 219 of 276 ATTACHMENT 1EAL BasesCategory: Fission Product Barrier DegradationSubcategory: N/AInitiating Condition: Loss of any two barriers and loss or potential loss of third barrierEAL:FGI.1 General EmergencyLoss of any two barriersANDLoss or potential loss of third barrier (Table F-I)Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):NoneCPNPP Basis:Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 (Attachment2) 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 barrierNEI 99-01 Basis:NoneCPNPP Basis Reference(s):1. NEI 99-01 FG1Page 220 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesIntroductionTable 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 table isstructured so that each of the three barriers occupies adjacent columns. Each fission productbarrier column is further divided into two columns; one for Loss thresholds and one forPotential 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. RCS or SG Tube LeakageB. Inadequate Heat removalC. CNTMT Radiation / RCS ActivityD. CNTMT Integrity or BypassE. Emergency Coordinator JudgmentEach category occupies a row in Table F-1 thus forming a matrix defined by the categories.The intersection of each row with each Loss/Potential Loss column forms a cell in which one ormore fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for abarrier 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 category titleand number. For example, the first Fuel Clad barrier Loss in Category A would be assigned"FC Loss A.I," the third Containment barrier Potential Loss in Category C would be assigned"CNTMT P-Loss C.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 necessary toexceed 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 the applicablefission product barrier Loss and Potential Loss thresholds. This structure promotes asystematic approach to assessing the classification status of the fission product barriers.When equipped with knowledge of plant conditions related to the fission product barriers, theEAL-user first scans down the category column of Table F-i, locates the likely category andthen reads across the fission product barrier Loss and Potential Loss thresholds in thatcategory to determine if a threshold has been exceeded. If a threshold has not been exceeded,the EAL-user proceeds to the next likely category and continues review of the thresholds in thenew categoryIf the EAL-user determines that any threshold has been exceeded, by definition, the barrier islost or potentially lost -even if multiple thresholds in the same barrier column are exceeded,only that one barrier is lost or potentially lost. The EAL-user must examine each of the threefission product barriers to determine if other barrier thresholds in the category are lost orpotentially lost. For example, if containment radiation is sufficiently high, a Loss of the FuelClad and RCS barriers and a Potential Loss of the Containment barrier can occur. BarrierPage 221 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesLosses and Potential Losses are then applied to the algorithms given in EALs FG1.1, FSI.1,and FA1.1 to determine the appropriate emergency classification.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 222 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesTable F-1 Fission Product Barrier Threshold MatrixFuel Clad (FC) Barrier Reactor Coolant System (RCS) Barrier Containment (CNTMT) BarrierCategory Loss Potential Loss Loss Potential Loss Loss Potential Loss1. An automatic or manual ECCS 1. Operation of a standby chargingA (SI) actuation required by pump is required by EITHER:EITHER: | |||
* UNISOLABLE RCS leakage 1. A leaking or RUPTURED SG is NoneRCS or None None UNISOLABLE RCS | |||
* SG tube leakage FAULTED outside of containmentSG Tube leakage 2. CSFST Integrity-RED PathLeakage SG tube RUPTURE conditions met1. CSFST Core Cooling-ORANGEB Path conditions met 1. CSFST Heat Sink-RED Path 1. CSFST Core Cooling-RED PathInadequate 1. CSFST Core Cooling-RED 2. CSFST Heat Sink-RED Path conditions met conditions metPath conditions met conditions met None AND None ANDHeat AND Heat sink is required Restoration procedures notRemoval Heat sink is required effective within 15 min. (Note 1)RemovalHeat sink is requiredI. Containment radiation greater than85 RJhrCTEuo16 Containment HRRM(L-RE-6290A), or 1. Containment radiation greater than 1. Containment radiation greater thanC CTWu_17 Containment HRRM 5 R/hr 2,110 R/hrCNTMT W-RE-6290B) CTEu16 Containment HRRM CTEu16 Containment HRRMR NT 2. Dose equivalent 1-131 coolant None (L-RE-6290A), or None None (L-RE-6290A), orRadiation activity greater than 300 CTWu17 Containment HRRM CTWu17 Containment HRRMI RCS jCilco (L-RE-6290B) Lu-RE-6290B)Activity 3. Gross Failed Fuel Monitor,FFLu60 (Q-RE-0406),radiation greater than 1.0E04pCi/cc1. Containment isolation isrequiredAND EITHER: 1. CSFST Containment-RED Pathconditions metContainment integrity hasD been lost based on 2. Containment hydrogen concentratiorEmergency Coordinator greater than 4%CNTMT None None None None judgment 3. Containment pressure greater thanIntegrity UNISOLABLE pathwiay from 18 psig With neither Containmentor Bypass Containment to the environment Spray system train operatingexists greater than or equal to 15 rain.2. Indications of RCS leakage (Note 1)outside of ContainmentE 1. Any condition in the opinion of 1. Any condition in the opinion of 1. Any condition in the opinion of 1. Any condition in the opinion of the 1. Any condition in the opinion of 1. Any condition in the opinion of thethe Emergency Coordinator that the Emergency Coordinator that the Emergency Coordinator that Emergency Coordinator that the Emergency Coordinator that Emergency Coordinator thatEC indicates loss of the fuel clad indicates potential loss of the fuel indicates loss of the RCS barrier indicates potential loss of the RCS indicates loss of the Containment indicates potential loss of theJudgment barrier clad barrier barrier barrier Containment barrier[Document No.] Rev. 6 Page 223 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Fuel CladCategory: A. RCS or SG Tube LeakageDegradation Threat: LossThreshold:None[Document No.] Rev. 6 Page 224 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:Category:Degradation Threat:Threshold:Fuel CladA. RCS or SG Tube LeakagePotential LossLNone[[Document No.] Rev. 6 1 Page 225 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesFuel CladBarrier:Category:B. Inadequate Heat RemovalDegradation Threat: LossThreshold:1. CSFST Core Cooling-RED Path conditions metDefinition(s):NoneBasis:Plant-SpecificCritical Safety Function Status Tree (CSFST) Core Cooling-RED Path indicates significantcore exit superheating and core uncovery. The CSFSTs are normally monitored using theSPDS display on the Plant Computer (ref. 1).GenericThis reading indicates temperatures within the core are sufficient to cause significantsuperheating of reactor coolant.CPNPP Basis Reference(s):1. FRC-0.1A/B Response to Inadequate Core Cooling2. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.AS[Document No.] I Rev. 6 1 Page 226 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesFuel CladBarrier:Category:B. Inadequate Heat RemovalDegradation Threat: Potential LossThreshold:1. CSFST Core Cooling-ORANGE Path conditions metDefinition(s):NoneBasis:Plant-SpecificCritical Safety Function Status Tree (CSFST) Core Cooling-ORANGE path indicates indicatessubcooling has been lost and that some fuel clad damage may potentially occur. The CSFSTsare normally monitored using the SPDS display on the Plant Computer (ref. 1, 2).GenericThis reading indicates a reduction in reactor vessel water level sufficient to allow the onset ofheat-induced cladding damage.CPNPP Basis Reference(s):1. FRC-0.1A/B Response to Inadequate Core Cooling2. FRC-0.2A/B Response to Degraded Core Cooling3. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.AI[Document No.] I Rev. 6 Page 227 of 276 1 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Fuel CladCategory: B. Inadequate Heat RemovalDegradation Threat: Potential LossThreshold:2. CSFST Heat Sink-RED Path conditions metANDHeat sink is requiredDefinition(s):NoneBasis:Plant-SpecificIn combination with RCS Potential Loss B.1, meeting this threshold results in a Site AreaEmergency.Critical Safety Function Status Tree (CSFST) Heat Sink-RED Path indicates the ultimate heatsink function is under extreme challenge and that some fuel clad damage may potentiallyoccur (ref. 1).The CSFSTs are normally monitored using the SPDS display on the Plant Computer (ref. 1).The phrase "and heat sink required" precludes the need for classification for conditions inwhich RCS pressure is less than SG pressure or Heat Sink-RED path entry was createdthrough operator action directed by an ERG. For example, FRH-0.1 is entered from CSFSTHeat Sink-Red. Step 1 tells the operator to determine if heat sink is required by checking thatRCS pressure is greater than any non-faulted SG pressure and RCS temperature is greaterthan 3500F. If these conditions exist, Heat Sink is required. Otherwise, the operator is to eitherreturn to the procedure and step in effect and place RHR in service for heat removal. For largeLOCA events inside the Containment, the SGs are moot because heat removal through thecontainment heat removal systems takes place. Therefore, Heat Sink Red should not berequired and, should not be assessed for EAL classification because a LOCA event aloneshould not require higher than an Alert classification. (ref. 1).GenericThis condition indicates an extreme challenge to the ability to remove RCS-heat using thesteam generators (i.e., loss of an effective secondary-side heat sink). This conditionrepresents a potential loss of the Fuel Clad Barrier. In accordance with EOPs, there may beunusual accident conditions during which operators intentionally reduce the heat removalcapability of the steam generators; during these conditions, classification using threshold is notwarranted.I[Document No.] I Rev. 6 1 Page 228 of276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesCPNPP Basis Reference(s):1. FRH-O.1A/B Response to Loss of Secondary Heat Sink2. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.B[Document No.] Rev. 6 Page 229 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Fuel CladCategory: C. CNTMT Radiation / RCS ActivityDegradation Threat: LossThreshold:1. Containment radiation greater than 85 R/hrCTEu16 Containment HRRM (u-RE-6290A), orCTWu17 Containment HRRM (u-RE-6290B)Definition(s):NoneBasis:Plant-SpecificContainment radiation monitor readings greater than 85 R/hr indicate the release of reactorcoolant, with elevated activity indicative of fuel damage, into the Containment. The reading isderived assuming the instantaneous release and dispersal of the reactor coolant noble gasand iodine inventory associated with a 2% clad failures into the Containment atmosphere.Reactor coolant concentrations of this magnitude are several times larger than the maximumconcentrations (including iodine spiking) allowed within technical specifications and aretherefore indicative of fuel damage. This value is higher than that specified for RCS Loss C.1(ref. 2, 3).Per NUS-1 74, the design basis CPNPP RCS specific activity for 1% fuel defects is 340tICi/gm; therefore, a threshold corresponding to 2% fuel clad damage correlates to a coolantactivity of 680 pCi/gm. VL-03-000032 Figure 2A/2B (CRM2) corresponds to approximately50% clad damage released to the containment atmosphere. Figure 2A/2B provides severalpotential limits depending on the pressure of the RCS and the presence of containment spray.The high RCS pressure with containment spray is the most limiting threshold; however, perNEI 99-01, the fuel clad barrier loss threshold should represent a loss of both the fuel clad andRCS barriers. Therefore, the value of curve representing low RCS pressure with spray wasused. The change in dose rates based on amount of fuel defects is a linear function; therefore,the threshold at 2% fuel defects is: 2120R/hr *(2% / 50%) = 85 R/hr (ref. 2).The Containment High Range Radiation Monitors (HRRMs) provide indication of radiationlevels in Containment during and after postulated accidents. The monitors are two ion chamberdetectors located on the 905' level of Containment approximately 900 apart. The range of eachmonitor is 1 to 108 R/hr. The output of each detector is fed to an RM-80 located outsideContainment. The RM-80 provides monitoring, alarming, and recording functions for themonitor channel. The RM-80 works in conjunction with the PC-1 1, RM-21, and RM-23assemblies. (ref. 1)I [Document No.] Rev. 6 Page 230 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesGenericThe radiation monitor reading corresponds to an instantaneous release of all reactor coolantmass into the containment, assuming that reactor coolant activity equals 300 pCi/gm doseequivalent 1-131. Reactor coolant activity above this level is greater than that expected foriodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Sincethis condition indicates that a significant amount of fuel clad damage has occurred, itrepresents a loss of the Fuel Clad Barrier.The radiation monitor reading in this threshold is higher than that specified for RCS BarrierLoss threshold ,-.AC.1 since it indicates a loss of both the Fuel Clad Barrier and the RCSBarrier. Note that a combination of the two monitor readings appropriately escalates themrn..gc..Y classification I"ve'ECL to a Site Area Emergency.CPNPP Basis Reference(s):1. DBD-EE-023 Radiation Monitoring System2. Evaluation performed by Design Engineering & Analysis (Andrea Lemons) (AI-CR-2014-012646-15)3. EPP-312 Core Damage Assessment4. NEI 99-01 CNTMT Radiation / RCS Activity Fuel Clad Loss 3.Ai[Document No.] I Rev. 6 1 Page 231 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesFuel CladBarrier:Category:C. CNTMT Radiation / RCS ActivityDegradation Threat: LossThreshold:2. Dose equivalent 1-131 coolant activity greater than 300 pCi/ccDefinition(s):NoneBasis:Plant-SpecificNoneGenericThis threshold indicates that RCS radioactivity concentration is greater than 300 pCi/gm doseequivalent 1-131. Reactor coolant activity above this level is greater than that expected foriodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Sincethis condition indicates that a significant amount of fuel clad damage has occurred, itrepresents a loss of the Fuel Clad Barrier.There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.CPNPP Basis Reference(s):1. NEI 99-01 CNTMT Radiation / RCS Activity Fuel Clad Loss 3.B[Document No.] Rev. 6 Page 232 of276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Fuel CladCategory: C. CNTMT Radiation / RCS ActivityDegradation Threat: LossThreshold:3. Gross Failed Fuel Monitor, FFLu60 Lu-RE-0406), radiation greater than 1.0E04 pCi/mlIDefinition(s):NoneBasis:Plant-SpecificThe normal Chemical and Volume Control System (CVCS) charging and letdown flow pathallows purification of the reactor coolant and control of the RCS volume while maintaining acontinuous feed and bleed flow between the RCS and the CVCS. Reactor coolant is first"letdown" from the RCS through a regenerative heat exchanger, which minimizes heat lossesfrom the RCS. Additional cooling takes place in a letdown heat exchanger that acts as the heatsink for the system. Downstream of the letdown heat exchanger pressure control valve andupstream of the mixed bed demineralizers, the letdown stream passes by a Geiger-Muellerradiation detector, FFLu60 (u-RE-0406), mounted on the reactor coolant letdown line tomonitor coolant activity and warn of fission products in the letdown coolant if a fuel elementfailure occurs. Detection of increased coolant activity may be indicative of failed fuel. Themonitor initiates Alert and High alarms in the Control Room (PC-1 1 and Plant Computer). (ref.1).Core Damage Assessment Guidelines (VL-03-000032) which was incorporated into EPP-312"Core Damage Assessment" provides the basis for loss of the Fuel Cladding as monitored bythe Gross Failed Fuel Monitor. The setpoint recommended by Westinghouse is 1 E+04 pCi/ml(ref. 2, 3).FFLu60 (u-RE-0406) has a range of 1 E-2 -1 E+7 pCi/ml.GenericNoneCPNPP Basis Reference(s):1. DBD-EE-023 Radiation Monitoring System2. Evaluation performed by Design Engineering & Analysis (Andrea Lemons) (AI-CR-2014-012646-15)3. EPP-312 Core Damage Assessment4. NEI 99-01 Other Indications Fuel Clad Loss 5.A[Document No.] Rev. 6 Page 233 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Fuel CladCategory: C. CNTMT Radiation / RCS ActivityDegradation Threat: Potential LossThreshold:None[Document No.] Rev. 6 Page 234 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:Category:Degradation Threat:Threshold:Fuel CladD. CNTMT Integrity or BypassLossNoneI [Document No.] Rev. 6 1 Page 235 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:Category:Degradation Threat:Threshold:Fuel CladD. CNTMT Integrity or BypassPotential LossNoneI [Document No.] Rev. 6 Page 236 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesFuel CladBarrier:Category:E. Emergency Coordinator JudgmentDegradation Threat: LossThreshold:1. Any condition in the opinion of the Emergency Coordinator that indicates loss of theFuel Clad barrierDefinitNoneBasis:ion(s):Plant-SpecificThe Emergency Coordinator 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 relativelyshort period of time based on a projection of current safety system performance. Theterm "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 reliable indicators.This assessment should include instrumentation operability concerns, readings fromportable instrumentation and consideration of offsite monitoring results.* Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergency classificationdeclarations.GenericThis threshold addresses any other factors that are to be used by the Emergency DireotO-Coordinator in determining whether the Fuel Clad barrier is lostCPNPP Basis Reference(s):1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.AI [Document No.] Rev. 6 Page 237 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Fuel CladCategory: E. Emergency Coordinator JudgmentDegradation Threat: Potential LossThreshold:1. Any condition in the opinion of the Emergency Coordinator that indicates potential lossof the Fuel Clad barrierBasis:Plant-SpecificThe Emergency Coordinator 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 accident sequences.-Imminent barrier degradation exists if the degradation will likely occur within relativelyshort period of time based on a projection of current safety system performance. Theterm "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 reliable indicators.This assessment should include instrumentation operability concerns, readings fromportable instrumentation and consideration of offsite monitoring results.-Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergency classificationdeclarations.GenericThis threshold addresses any other factors that are to be used by the EmergencyCoordinatorDireGtei in determining whether the Fuel Clad barrier is potentially lost. TheEmergency D!reetE)FCoordinator should also consider whether or not to declare the barrierpotentially lost in the event that barrier status cannot be monitored.CPNPP Basis Reference(s):1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A[Document No.] Rev. 6 Page 238 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Reactor Coolant SystemCategory: A. RCS or SG Tube LeakageDegradation Threat: LossThreshold:1. An automatic or manual ECCS (SI) actuation required by EITHER:* UNISOLABLE RCS leakage* SG tube RUPTUREDefinition(s):UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally.RUPTURE -The condition of a steam generator in which primary-to-secondary leakage is ofsufficient magnitude to require a safety injection.Basis:Plant-SpecificECCS (SI) actuation is caused by (ref. 1):* Pressurizer low pressure less than 1820 psig* Steamline low pressure less than 610 psig* Containment high pressure greater than 3.0 psigGenericThis threshold is based on an UNISOLABLE RCS leak of sufficient size to require anautomatic or manual actuation of the Emergency Core Cooling System (ECCS). This conditionclearly represents a loss of the RCSBarrier.This threshold is applicable to unidentified and pressure boundary leakage, as well asidentified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacingsystem. The mass loss may be into any location -inside containment, to the secondary-side(i.e., steam generator tube leakage) or outside of containment.A steam generator with primary-to-secondary leakage of sufficient magnitude to require asafety injection is considered to be RUPTURED. If a RUPTURED steam generator is alsoFAULTED outside of containment, the declaration escalates to a Site Area Emergency sincethe Containment Barrier Loss threshold I.A will also be met.CPNPP Basis Reference(s):1. EOP-0.0A/B Reactor Trip or Safety Injection2. EOP-3.OA/B Steam Generator Tube Rupture3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Loss I.A[Document No.] I Rev. 6 Page 239 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Reactor Coolant SystemCategory: A. RCS or SG Tube LeakageDegradation Threat: Potential LossThreshold:1. Operation of a standby charging pump is required by EITHER:* UNISOLABLE RCS leakage* SG tube leakageDefinition(s):UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally.Basis:Plant-SpecificThe Chemical and Volume Control System (CVCS) includes three charging pumps (onepositive displacement pump and two centrifugal charging pumps) that take suction from thevolume control tank and return the cooled, purified reactor coolant to the RCS. The centrifugalcharging pumps in the CVCS also serve as the high-head safety injection pumps in theEmergency Core Cooling System. Positive displacement pump capacity is 98 gpm. Thecapacity of each centrifugal pump is 150 gpm. A second charging pump being required(positive displacement or centrifugal) is indicative of a substantial RCS leak. (ref. 1, 2, 3)GenericThis threshold is based on an UNISOLABLE RCS leak that results in the inability to maintainpressurizer level within specified limits by operation of a normally used charging (makeup)pump, but an ECCS (SI) actuation has not occurred. The threshold is met when an operatingprocedure, or operating crew supervision, directs that a standby charging (makeup) pump beplaced in service to restore and maintain pressurizer level.This threshold is applicable to unidentified and pressure boundary leakage, as well asidentified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacingsystem. The mass loss may be into any location -inside containment, to the secondary-side(i.e., steam generator tube leakage) or outside of containment.If a leaking steam generator is also FAULTED outside of containment, the declarationescalates to a Site Area Emergency since the Containment Barrier Loss threshold I.A will alsobe met.CPNPP Basis Reference(s):1. FSAR 9.3.42. FSAR Table 9.3-73. SOP-103A/B Chemical and Volume Control System4. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss I.A[Document No.] Rev. 6 Page 240 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesReactor Coolant SystemBarrier:Category:A. RCS or SG Tube LeakageDegradation Threat: Potential LossThreshold:2. CSFST Integrity-RED Path conditions metDefinition(s):NoneBasis:Plant-SpecificCritical Safety Function Status Tree (CSFST) RCS Integrity-RED path indicates the RCSbarrier is under significant challenge (ref. 1).GenericThis condition indicates an extreme challenge to the integrity of the RCS pressure boundaryI due to pressurized thermal shock -a transient that causes rapid RCS cooldown while the RCSis in Mode 3 or higher (i.e., hot and pressurized).CPNPP Basis Reference(s):1. FRP-0.1A/B Response to Imminent Pressurized Thermal Shock Condition2. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss I.BI [Document No.] Rev. 6 1 Page 241 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:Category:Degradation Threat:Threshold:Reactor Coolant SystemB. Inadequate Heat RemovalLossNone[Document No.] Rev. 6 Page 242 of2 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Reactor Coolant SystemCategory: B. Inadequate Heat RemovalDegradation Threat: Potential LossThreshold:1. CSFST Heat Sink-RED path conditions metANDHeat sink is requiredDefinition(s):NoneBasis:Plant-SpecificIn combination with FC Potential Loss B.2, meeting this threshold results in a Site AreaEmergency.Critical Safety Function Status Tree (CSFST) Heat Sink-RED Path indicates the ultimate heatsink function is under extreme challenge and that some fuel clad damage may potentiallyoccur (ref. 1).The CSFSTs are normally monitored using the SPDS display on the Plant Computer (ref. 1).The phrase "and heat sink required" precludes the need for classification for conditions inwhich RCS pressure is less than SG pressure or Heat Sink-RED path entry was createdthrough operator action directed by an ERG. For example, FRH-0.1 is entered from CSFSTHeat Sink-Red. Step 1 tells the operator to determine if heat sink is required by checking thatRCS pressure is greater than any non-faulted SG pressure and RCS temperature is greaterthan 3500F. If these conditions exist, Heat Sink is required. Otherwise, the operator is to eitherreturn to the procedure and step in effect and place RHR in service for heat removal. For largeLOCA events inside the Containment, the SGs are moot because heat removal through thecontainment heat removal systems takes place. Therefore, Heat Sink Red should not berequired and, should not be assessed for EAL classification because a LOCA event aloneshould not require higher than an Alert classification. (ref. 1).GenericThis condition indicates an extreme challenge to the ability to remove RCS heat using thesteam generators (i.e., loss of an effective secondary-side heat sink). This conditionrepresents a potential loss of the RCS Barrier. In accordance with EOPs, there may beunusual accident conditions during which operators intentionally reduce the heat removalcapability of the steam generators; during these conditions, classification using threshold is notwarranted.Meeting this threshold results in a Site Area Emergency because this threshold is identical toFuel Clad Barrier Potential Loss threshold 2B.2; both will be met. This condition warrants a[Document No.] I Rev. 6 Page 243 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesSite Area Emergency declaration because inadequate RCS heat removal may result in fuelheat-up sufficient to damage the cladding and increase RCS pressure to the point where masswill be lost from the system.CPNPP Basis Reference(s):1. FRH-O.1A/B Response to Loss of Secondary Heat Sink2. NEI 99-01 Inadequate Heat Removal RCS Loss 2.B[Document No.] Rev. 6 Page 244 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Reactor Coolant SystemCategory: C. CNTMT Radiation/ RCS ActivityDegradation Threat: LossThreshold:1. Containment radiation greater than 5 R/hrCTEu16 Containment HRRM (u-RE-6290A), orCTWu17 Containment HRRM (u-RE-6290B)Definition(s):N/ABasis:Plant-SpecificAs part of the elimination of the Post-Accident Sampling system, Westinghouse performedanalysis for Comanche Peak on Core Damage Assessment Guidelines (VL-03-000032) whichwas incorporated into EPP-312 "Core Damage Assessment". For setpoint CRM1,Westinghouse assumptions match the requirements of NEI 99-01 for RCS barrier loss with theexception that the level of radioactivity in the RCS is assumed to be at 10% of TechnicalSpecifications levels rather than 100% as recommended by NEI 99-01. However, this addsconservatism to this threshold. The limiting maximum value found in Figure 1A is 4.75 R/hr.This value is time dependent and corresponds to an hour after shutdown. This value has beenrounded to 5 R/hr for instrument readability (ref. 1, 3, 4, 5).The Containment High Range Radiation Monitors (HRRMs) provide indication of radiationlevels in Containment during and after postulated accidents. The monitors are two ion chamberdetectors located on the 905' level of Containment approximately 90' apart. The range of eachmonitor is 1 to 108 R/hr. The output of each detector is fed to an RM-80 located outsideContainment.The RM-80 provides monitoring, alarming, and recording functions for the monitor channel.The RM-80 works in conjunction with the PC-1 1, RM-21, and RM-23 assemblies. (ref. 2)GenericThe radiation monitor reading corresponds to an instantaneous release of all reactor coolantmass into the containment, assuming that reactor coolant activity equals TechnicalSpecification allowable limits. This value is lower than that specified for Fuel Clad Barrier Lossthreshold ,AC.1 since it indicates a loss of the RCS Barrier only.There is no Potential Loss threshold associated with RCS Activity I Containment Radiation.I [Document No.] Rev. 6 Page 245 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesCPNPP Basis Reference(s):1. Technical Specifications Table 3.3.3-12. DBD-EE-023 Radiation Monitoring System3. Evaluation performed by Design Engineering & Analysis (Andrea Lemons) (AI-CR-2014-012646-15)4. Technical Specifications B3.3.35. EPP-312 Core Damage Assessment6. NEI 99-01 CNTMT Radiation / RCS Activity RCS Loss 3.AI [Document No.] Rev. 6 Page 246 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Reactor Coolant SystemCategory: B. CNTMT Radiation/ RCS ActivityDegradation Threat: Potential LossThreshold:None[Document No.] Rev. 6 1 Page 247 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Reactor Coolant SystemCategory: D. CNTMT Integrity or BypassDegradation Threat: LossThreshold:NoneI [Document No.] I Rev. 6 Page 248 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Reactor Coolant SystemCategory: D. CNTMT Integrity or BypassDegradation Threat: Potential LossThreshold:NoneL[Document No.] I Rev. 6 Page 249 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesReactor Coolant SystemBarrier:Category:Degradation Threat:E. Emergency Coordinator JudgmentLossThreshold:Definition(s):NoneBasis:Plant-SpecificThe Emergency Coordinator judgment threshold addresses any other factors relevant todetermining if the RCS barrier is lost. Such a determination should include imminent barrierdegradation, barrier monitoring capability and dominant accident sequences." Imminent barrier degradation exists if the degradation will likely occur within relativelyshort period of time based on a projection of current safety system performance. Theterm "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 reliable indicators.This assessment should include instrumentation operability concerns, readings fromportable instrumentation and consideration of offsite monitoring results.* Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergency classificationdeclarations.GenericThis threshold addresses any other factors that may be used by the Emergency QieGteCoordinator in determining whether the RCS Barrier is lost.CPNPP Basis Reference(s):1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A[Document No.] Rev. 6 Page 250 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Reactor Coolant SystemCategory: E. Emergency Coordinator JudgmentDegradation Threat: Potential LossThreshold:1. Any condition in the opinion of the Emergency Coordinator that indicates potential lossof the RCS barrierDefinition(s):NoneBasis:Plant-SpecificThe Emergency Coordinator judgment threshold addresses any other factors relevant todetermining if the RCS barrier is potentially 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 relativelyshort period of time based on a projection of current safety system performance. Theterm "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 reliable indicators.This assessment should include instrumentation operability concerns, readings fromportable instrumentation and consideration of offsite monitoring results.* Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergency classificationdeclarations.GenericThis threshold addresses any other factors that may be used by the Emergency Direet4Coordinator in determining whether the RCS Barrier is potentially lost. The EmergencyDirector should also consider whether or not to declare the barrier potentially lost in the eventthat barrier status cannot be monitored.CPNPP Basis Reference(s):1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6.AI [Document No.] Rev. 6 1 Page 251 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: ContainmentCategory: A. RCS or SG Tube LeakageDegradation Threat: LossThreshold:1. A leaking or RUPTURED SG is FAULTED outside of containmentDefinition(s):FAULTED -The term applied to a steam generator that has a steam leak on the secondaryside of sufficient size to cause an uncontrolled drop in steam generator pressure or the steamgenerator to become completely depressurized.RUPTURED -The condition of a steam generator in which primary-to-secondary leakage is ofsufficient magnitude to require a safety injection.Basis:Plant-SpecificNone.GenericThis threshold addresses a leaking or RUPTURED Steam Generator (SG) that is alsoFAULTED outside of containment. The condition of the SG, whether leaking or RUPTURED,is determined in accordance with the thresholds for RCS Barrier Potential Loss 4-.A.1 and Loss4-A.1, respectively. This condition represents a bypass of the containment barrier.FAULTED is a defined term within the NEI 99-01 methodology; this determination is notnecessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if thepressure in a steam generator is decreasing uncontrollably .(part of the FAULTED definition).and the FAULTED steam generator isolation procedure is not entered because EOP user rulesare dictating implementation of another procedure to address a higher priority condition, thesteam generator is still considered FAULTED for emergency classification purposes.The FAULTED criterion establishes an appropriate lower bound on the size of a steam releasethat may require an emergency classification. Steam releases of this size are readilyobservable with normal Control Room indications. The lower bound for this aspect of the[Document No.] Rev. 6 Page 252 of276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and Basescontainment barrier is analogous to the lower bound criteria specified in IC SU4 for the fuelclad barrier (i.e., RCS activity values) and IC SU5 for the RCS barrier (i.e., RCS leak ratevalues).This threshold also applies to prolonged steam releases necessitated by operationalconsiderations such as the forced steaming of a leaking or RUPTURED steam generatordirectly to atmosphere to cooldown the plant, or to drive an auxiliary (emergency) feed waterpump. These types of conditions will result in a significant and sustained release of radioactivesteam to the environment (and are thus similar to a FAULTED condition). The inability toisolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss ofcontainment.Steam releases associated with the expected operation of a SG power operated relief valve orsafety relief valve do not meet the intent of this threshold. Such releases may occurintermittently for a short period of time following a reactor trip as operators process throughemergency operating procedures to bring the plant to a stable condition and prepare to initiatea plant cooldown. Steam releases associated with the unexpected operation of a valve (e.g., astuck-open safety valve) do meet this threshold.Following an SG tube leak or rupture, there may be minor radiological releases through asecondary-side system component (e.g., air ejectors, glad seal exhausters, valve packing,etc.). These types of releases do not constitute a loss or potential loss of containment butshould be evaluated using the Recognition Category A-R ICs.The emergenYcy clasification levelECLs resulting from primary-to-secondary leakage, with orwithout a steam release from the FAULTED SG, are summarized below.I [Document No.] I Rev. 6 Page 253 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesAffected SG is FAULTEDOutside of Containment?P-to-S Leak RateYesNoLess than or equal to 25 gpmGreater than 25 gpmRequires operation of a standbycharging (makeup) pump (RCSBarrier Potential Loss)Requires an automatic or manualECCS (SI) actuation (RCSBarrierLoss)No classificationUnusual Event perSU4SU5.1Site Area Emergency perFS1.1Site Area Emergency perFS1.1No classificationUnusual Event perSJ4SU5.1Alert per FAI.1Alert per FAI.1There is no Potential Loss threshold associated with RCS or SG Tube Leakage.CPNPP Basis Reference(s):1. EOP-3.0 Steam Generator Tube Rupture2. EOP-2.OA/B Faulted Steam Generator Isolation3. NEI 99-01 RCS or SG Tube Leakage Containment Loss I.A[Document No.] Rev. 6 Page 254 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: ContainmentCategory: A. RCS or SG Tube LeakageDegradation Threat: Potential LossThreshold:NoneI [Document No.] Rev. 6 Page 255 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:Category:Degradation Threat:Threshold:NoneContainmentB. Inadequate heat RemovalLossI [Document No.] Rev. 6 Page 256 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesContainmentBarrier:Category:B. Inadequate heat RemovalDegradation Threat: Potential LossThreshold:1. CSFST Core Cooling-RED Path conditions metANDRestoration procedures not effective within 15 min. (Note 1)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has beenexceeded, or will likely be exceeded.Definition(s):NoneBasis:Plant-SpecificCritical Safety Function Status Tree (CSFST) Core Cooling-RED path indicates significant coreexit superheating and core uncovery. The CSFSTs are normally monitored using the SPDSdisplay on the Plant Computer (ref. 1).The function restoration procedures are those emergency operating procedures that addressthe recovery of the core cooling critical safety functions. The procedure is considered effectiveif the temperature is decreasing or if the vessel water level is increasing (ref. 1).A direct correlation to status trees can be made if the effectiveness of the restorationprocedures is also evaluated. If core exit thermocouple (TC) readings are greater than 1,2000F(ref. 1), Fuel Clad barrier is also lost.GenericThis threshold addresses any other factors that may be used by the Emergency ir-eeGtCoordinator in determining whether the RCS Barrier is potentially lost. The EmergencyDirector should also consider whether or not to declare the barrier potentially lost in the eventthat barrier status cannot be monitored.CPNPP Basis Reference(s):1. FRC-O.1A/B Response to Inadequate Core Cooling2. NEI 99-01 Inadequate Heat Removal Containment Potential Loss 2.AI[Document No.] I Rev. 6 -Page 257 of 276 ý ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: ContainmentCategory: C. CNTMT Radiation/RCS ActivityDegradation Threat: LossThreshold:None[Document No.] Rev. 6 Page 258 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: ContainmentCategory: C. CNTMT Radiation/RCS ActivityDegradation Threat: Potential LossThreshold:1. Containment radiation greater than 1,110 R/hrCTEul 6 Containment HRRM (u-RE-6290A), orCTWu17 Containment HRRM (u-RE-6290B)Definition(s):NoneBasis:Plant-SpecificContainment radiation monitor readings greater than 1,110 R/hr indicate significant fueldamage well in excess of that required for loss of the RCS barrier and the Fuel Clad barrier.Regardless of whether the Containment barrier itself is challenged, this amount of activity incontainment could have severe consequences if released. It is, therefore, prudent to treat thisas a Potential Loss of the Containment barrier. (ref. 2, 3)The readings are higher than that specified for Fuel Clad Loss C.3 and RCS Loss C.1.Containment radiation readings at or above the Containment barrier Potential Loss threshold,therefore, signify a loss of two fission product barriers and Potential Loss of a third, indicatingthe need to upgrade the emergency classification to a General Emergency.The analysis performed in VL-03-00032 was used to determine the potential containment lossthreshold. Since the containment potential loss threshold also assumes a loss of the RCSbarrier, the Figure 3A (CRM3) curve which represents the dose response at low RCS pressurewith sprays present was used. The setpoint was developed on the assumption of 100% fuelrod rupture with 100% of the noble gas and 50% of the iodine and cesium in the RCS releasedto containment. With containment spray operating, the containment inventory of all fissionproducts except the noble gases are reduced by a factor of 100. Per Figure 3A, the value onehour after shutdown for 100% rod rupture is 5560 R/hr; therefore, the EAL threshold at 20%fuel defects is: 5,560 R/hr *20% = 1,112 R/hr (rounded to 1,110 R/hr for instrument readability)(ref. 2, 3).The Containment High Range Radiation Monitors (HRRMs) provide indication of radiationlevels in Containment during and after postulated accidents. The monitors are two ion chamberdetectors located on the 905' level of Containment approximately 90' apart. The range of eachmonitor is 1 to 108 R/hr. The output of each detector is fed to an RM-80 located outsideContainment. The RM-80 provides monitoring, alarming, and recording functions for themonitor channel. The RM-80 works in conjunction with the PC-1 1, RM-21, and RM-23assemblies. (ref. 1).I [Document No.] I Rev. 6 Page 259 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesGenericThe radiation monitor reading corresponds to an instantaneous release of all reactor coolantmass into the containment, assuming that 20% of the fuel cladding has failed. This level offuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss andRCS Barrier Loss thresholds.NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power PlantAccidents, indicates the fuel clad failure must be greater than approximately 20% in order forthere to be a major release of radioactivity requiring offsite protective actions. For thiscondition to exist, there must already have been a loss of the RCS Barrier and the Fuel CladBarrier. It is therefore prudent to treat this condition as a potential loss of containment whichwould then escalate the emergencY claSificatilo; levelECL to a General Emergency.CPNPP Basis Reference(s):1. DBD-EE-023 Radiation Monitoring System2. Evaluation performed by Design Engineering & Analysis (Andrea Lemons) (AI-CR-2014-012646-15)3. EPP-312 Core Damage Assessment4. NEI 99-01 CNTMT Radiation / RCS Activity Containment Potential Loss 3.AI [Document No.] Rev. 6 1 Page 260 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: ContainmentCategory: D. CNTMT Integrity or BypassDegradation Threat: LossThreshold:1. Containment isolation is requiredAND EITHER:* Containment integrity has been lost based on Emergency Coordinator judgment" UNISOLABLE pathway from containment to the environment existsDefinition(s):UNISOLABLE- An open or breached system line that cannot be isolated, remotely or locally.Basis:Plant-SpecificNoneGenericThese thresholds address a situation where containment isolation is required and one of twoconditions exists as discussed below. Users are reminded that there may be accident andrelease conditions that simultaneously meet both bulleted thresholds 4.A.1 and 4.A.2.4-A41First Threshold -Containment integrity has been lost, i.e., the actual containmentatmospheric leak rate likely exceeds that associated with allowable leakage (or sometimesreferred to as design leakage). Following the release of RCS mass into containment,containment pressure will fluctuate based on a variety of factors; a loss of containment integritycondition may (or may not) be accompanied by a noticeable drop in containment pressure.Recognizing the inherent difficulties in determining a containment leak rate during accidentconditions, it is expected that the Emergency DirectEr Coordinator will assess this thresholdusing judgment, and with due consideration given to current plant conditions, and availableoperational and radiological data (e.g., containment pressure, readings on radiation monitorsoutside containment, operating status of containment pressure control equipment, etc.).Refer to the middle piping run of Figure 9-F--41. Two simplified examples are provided. One isleakage from a penetration and the other is leakage from an in-service system valve.I [Document No.] Rev. 6 1 Page 261 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesDepending upon radiation monitor locations and sensitivities, the leakage could be detected byany of the four monitors depicted in the figure.Another example would be a loss or potential loss of the RCS barrier, and the simultaneousoccurrence of two FAULTED locations on a steam generator where one fault is located insidecontainment (e.g., on a steam or feedwater line) and the other outside of containment. In thiscase, the associated steam line provides a pathway for the containment atmosphere to escapeto an area outside the containment.Following the leakage of RCS mass into containment and a rise in containment pressure, theremay be minor radiological releases associated with allowable (design) containment leakagethrough various penetrations or system components. These releases do not constitute a lossor potential loss of containment but should be evaluated using the Recognition Category A--RICs.4-A.2Second Threshold -Conditions are such that there is an UNISOLABLE pathway for themigration of radioactive material from the containment atmosphere to the environment. Asused here, the term "environment" includes the atmosphere of a room or area, outside thecontainment, that may, in turn, communicate with the outside-the-plant atmosphere (e.g.,through discharge of a ventilation system or atmospheric leakage). Depending upon a varietyof factors, this condition may or may not be accompanied by a noticeable drop in containmentpressure.Refer to the top piping run of Figure 8-F--41. In this simplified example, the inboard andoutboard isolation valves remained open after a containment isolation was required (i.e.,containment isolation was not successful). There is now an UNISOLABLE pathway from thecontainment to the environment.The existence of a filter is not considered in the threshold assessment. Filters do not removefission product noble gases. In addition, a filter could become ineffective due to iodine and/orparticulate loading beyond design limits (i.e., retention ability has been exceeded) or watersaturation from steam/high humidity in the release stream.Leakage between two interfacing liquid systems, by itself, does not meet this threshold.I [Document No.] Rev. 6 1 Page 262 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesRefer to the bottom piping run of Figure 8-F--4-1. In this simplified example, leakage in an RCPseal cooler is allowing radioactive material to enter the Auxiliary Building. The radioactivitywould be detected by the Process Monitor. If there is no leakage from the closed watercooling system to the Auxiliary Building, then no threshold has been met. If the pumpdeveloped a leak that allowed steam/water to enter the Auxiliary Building, then secondthreshold-44B would be met. Depending upon radiation monitor locations and sensitivities, thisleakage could be detected by any of the four monitors depicted in the figure and cause the firstthreshold 4-A-_--to be met as well.Following the leakage of RCS mass into containment and a rise in containment pressure, theremay be minor radiological releases associated with allowable containment leakage throughvarious penetrations or system components. Minor releases may also occur if a containmentisolation valve(s) fails to close but the containment atmosphere escapes to an enclosedsystem. These releases do not constitute a loss or potential loss of containment but should beevaluated using the Recognition Category A-RICs.The status of the containment barrier during an event involving steam generator tube leakageis assessed using Loss Threshold 4-.A.1.CPNPP Basis Reference(s):1. NEI 99-01 CNTMT Integrity or Bypass Containment Loss 4.A[Document No.] Rev. 6 Page 263 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: ContainmentCategory: D. CNTMT Integrity or BypassDegradation Threat: LossThreshold:2. Indications of RCS leakage outside of containmentDefinition(s):NoneBasis:Plant-SpecificECA-1.2A/B LOCA Outside Containment (ref. 1) provides instructions to identify and isolate aLOCA outside of the containment. Potential RCS leak pathways outside containment include(ref. 1):" Residual Heat Removal* Safety Injection" Chemical & Volume Control" RCP seals" PZR/RCS Loop sample linesGenericContainment sump, temperature, pressure and/or radiation levels will increase if reactorcoolant mass is- leaking into the containment. If these parameters have not increased, then thereactor coolant mass may be leaking outside of containment (i.e., a containment bypasssequence). Increases in sump, temperature, pressure, flow and/or radiation level readingsoutside of the containment may indicate that the RCS mass is being lost outside ofcontainment.Unexpected elevated readings and alarms on radiation monitors with detectors outsidecontainment should be corroborated with other available indications to confirm that the sourceis a loss of RCS mass outside of containment. If the fuel clad barrier has not been lost,radiation monitor readings outside of containment may not increase significantly; however,other unexpected changes in sump levels, area temperatures or pressures, flow rates, etc.should be sufficient to determine if RCS mass is being lost outside of the containment.Refer to the middle piping run of Figure 9-F--41. In this simplified example, a leak has occurredat a reducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending uponradiation monitor locations and sensitivities, the leakage could be detected by any of the fourmonitors depicted in the figure and cause threshold 4AD.1 to be met as well.To ensure proper escalation of the emergency classification, the RCS leakage outside ofcontainment must be related to the mass loss that is causing the RCS Loss and/or PotentialLoss threshold 4-.A.1 to be met.[Document No.] Rev. 6 Page 264 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesCPNPP Basis Reference(s):1. ECA-1.2A/B LOCA Outside Containment2. NEI 99-01 CNTMT Integrity or Bypass Containment Loss[Document No.] Rev. 6 Page 265 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesFigure 1: Containment Integrity or Bypass Examples-easeEffluent oing Monitor 9O'' wa-----------I [Document No.] Rev. 6 Page 266 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: ContainmentCategory: D. CNTMT Integrity or BypassDegradation Threat: Potential LossThreshold:1. CSFST Containment-RED Path conditions metDefinition(s):NoneBasis:Plant-SpecificCritical Safety Function Status Tree (CSFST) Containment-RED path is entered if containmentpressure is greater than or equal to 50 psig and represents an extreme challenge to safetyfunction. The CSFSTs are normally monitored using the SPDS display on the Plant Computer(ref. 1).50 psig is the containment design pressure and is the pressure used to define CSFSTContainment Red Path conditions (ref. 2).GenericIf containment pressure exceeds the design pressure, there exists a potential to lose theContainment Barrier. To reach this level, there must be an inadequate core cooling conditionfor an extended period of time; therefore, the RCS and Fuel Clad barriers would already belost. Thus, this threshold is a discriminator between a Site Area Emergency and GeneralEmergency since there is now a potential to lose the third barrier.CPNPP Basis Reference(s):1. FRC-Z.1A/B Response to High Containment Pressure2. FSAR Table 6.2.1-13. NEI 99-01 CNTMT Integrity or Bypass Containment Potential Loss 4.AI [Document No.] Rev. 6 1 Page 267 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: ContainmentCategory: D. CNTMT Integrity or BypassDegradation Threat: Potential LossThreshold:2. Containment hydrogen concentration greater than 4%Definition(s):NoneBasis:Plant-SpecificIn the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a coreuncovery could result in an explosive mixture of dissolved gasses in Containment. However,Containment monitoring and/or sampling should be performed to verify this assumption and aGeneral Emergency declared if it is determined that an explosive mixture exists. A combustiblemixture can be formed when hydrogen gas concentration in the Containment atmosphere isgreater than 4% by volume. All hydrogen measurements are referenced to concentrations indry air even though the actual Containment environment may contain significant steamconcentrations. The plant has two hydrogen monitoring systems. Each monitoring systemconsists of four sensor modules and one microprocessor analyzer. Two sensors from eachContainment are coupled to one of the two hydrogen microprocessors located in the ControlRoom. Thus each microprocessor analyzer is shared by Units 1 and 2. The analyzer systemhas a range of 0-10% hydrogen by volume. The detector modules are located on the 905',873', and 860' elevations in Containment. A fourth detector is located on 832' level across fromthe loop room entrance for loops 1 and 4. Hydrogen concentration is displayed in the ControlRoom on u-AI-5506A/B and u-AI-5506C/D.Hydrogen concentration can also be displayed on the Plant Computer. Alarms at -3% areprovided for high hydrogen concentration, u-ALB-3A, window 3.7. If a hydrogen concentrationvalue can not be obtained from the hydrogen monitoring system, a grab sample from thecontainment PIG radiation monitor may be used to determine the hydrogen concentration (ref.1,2,3,4).To generate such levels of combustible gas, loss of the Fuel Clad and RCS barriers must haveoccurred. With the Potential Loss of the Containment barrier, the threshold hydrogenconcentration, therefore, will likely warrant declaration of a General Emergency.GenericThe existence of an explosive mixture means, at a minimum, that the containment atmospherichydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagrationlimit). A hydrogen burn will raise containment pressure and could result in collateral equipmentdamage leading to a loss of containment integrity. It therefore represents a potential loss ofthe Containment Barrier.[Document No.] Rev. 6 Page 268 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesCPNPP Basis Reference(s):1. FRC-O.1A/B Response to Inadequate Core Cooling, Attachment 52. FSAR Section 6.2.53. FSAR Table 7.5-7A4. CHM-1 11, Primary Chemistry Accident Assessment Sampling Program7. NEI 99-01 CNTMT Integrity or Bypass Containment Potential Loss 4.BI[Document No.] Rev. 6 1 Page 269 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: ContainmentCategory: D. CNTMT Integrity or BypassDegradation Threat: Potential LossThreshold:3. Containment pressure greater than 18 psig with neither Containment Spray systemtrain operating per design for greater than or equal to 15 min. (Note 1)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has beenexceeded, or will likely be exceeded.Definition(s):NoneBasis:Plant-SpecificThis threshold represents a Potential Loss of the Containment barrier because theContainment heat removal and depressurization equipment (but not including Containmentventing strategies) is either lost or degraded. The Containment Spray System (CSS) isdesigned to remove heat from the Containment environment following a LOCA, a main steamline break accident, or a feedwater line break accident. Each unit of the CPNPP is equippedwith two redundant Containment spray trains, each designed to provide emergencyContainment heat removal in the event of a LOCA. This system, in conjunction with the ECCS,removes postaccident thermal energy from the Containment environment, thereby reducingthe Containment pressure and temperature. Each train includes two containment spraypumps, spray headers, nozzles, valves, and piping. Each train is powered from a separatesafeguard bus. (ref. 1)The Containment pressure setpoint (18 psig, ref. 2) is the pressure at which the ContainmentSpray System should actuate and begin performing its function. The design basis accidentanalyses and evaluations assume the loss of one Containment Spray System train (ref. 1).GenericThis threshold describes a condition where containment pressure is greater than the setpointat which containment energy (heat) removal systems are designed to automatically actuate,and less than one full train of equipment is capable of operating per design. The 15-minutecriterion is included to allow operators time to manually start equipment that may not haveautomatically started, if possible. This threshold represents a potential loss of containment inthat containment heat removal/depressurization systems (e.g., containment sprays, icecondenser fans, etc., but not including containment venting strategies) are either lost orperforming in a degraded manner.I [Document No.] I Rev. 6 Page 270 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesCPNPP Basis Reference(s):1. FSAR Section 6.2.22. FRC-Z.1A/B Response to High Containment Pressure3. NEI 99-01 CNTMT Integrity or Bypass Containment Potential Loss 4.C[Document No.] Rev. 6 Page 271 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: ContainmentCategory: E. Emergency Coordinator JudgmentDegradation Threat: LossThreshold:1. Any condition in the opinion of the Emergency Coordinator that indicates loss of theContainment barrierDefinition(s):NoneBasis:Plant-SpecificThe Emergency Coordinator judgment threshold addresses any other factors relevant todetermining if the Primary Containment barrier is lost. Such a determination should includeimminent barrier degradation, barrier monitoring capability and dominant accident sequences.* Imminent barrier degradation exists if the degradation will likely occur within relativelyshort period of time based on a projection of current safety system performance. Theterm "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 reliable indicators.This assessment should include instrumentation operability concerns, readings fromportable instrumentation and consideration of offsite monitoring results.* Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergency classificationdeclarations.GenericThis threshold addresses any other factors that may be used by the Emergency QikeeterCoordinator in determining whether the Containment Barrier is lost.CPNPP Basis Reference(s):1. NEI 99-01 Emergency Director Judgment PC Loss 6.A[Document No.] Rev. 6 Page 272 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: ContainmentCategory: E. Emergency Coordinator JudgmentDegradation Threat: Potential LossThreshold:1. Any condition in the opinion of the Emergency Coordinator that indicates potential lossof the Containment barrierDefinition(s):NoneBasis:Plant-SpecificThe Emergency Coordinator judgment threshold addresses any other factors relevant todetermining if the Primary Containment barrier is potentially lost. Such a determination shouldinclude imminent barrier degradation, barrier monitoring capability and dominant accidentsequences." Imminent barrier degradation exists if the degradation will likely occur within relativelyshort period of time based on a projection of current safety system performance. Theterm "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 reliable indicators.This assessment should include instrumentation operability concerns, readings fromportable instrumentation and consideration of offsite monitoring results." Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergency classificationdeclarations.GenericThis threshold addresses any other factors that may be used by the Emergency DiQeetoCoordinator in determining whether the Containment Barrier is lost.CPNPP Basis Reference(s):1. NEI 99-01 Emergency Director Judgment PC Potential Loss 6.AI [Document No.] I Rev. 6 1 Page 273 of 276 ATTACHMENT 3Safe Operation & Shutdown Areas Tables R-3 & H-2 BasesBackgroundNEI 99-01 Revision 6 ICs AA3 and HA5 prescribe declaration of an Alert based on impededaccess to rooms or areas (due to either area radiation levels or hazardous gas concentrations)where equipment necessary for normal plant operations, cooldown or shutdown is located.These areas are intended to be plant operating mode dependent. Specifically the DevelopersNotes for AA3 and HA5 states:The "site-specific list of plant rooms or areas with entry-related mode applicability identified"should specify those rooms or areas that contain equipment which require a manual/localaction as specified in operating procedures used for normal plant operation, cooldown andshutdown. Do not include rooms or areas in which actions of a contingent or emergencynature would be performed (e.g., an action to address an off-normal or emergency conditionsuch as emergency repairs, corrective measures or emergency operations). In addition, thelist should specify the plant mode(s) during which entry would be required for each room orarea.The list should not include rooms or areas for which entry is required solely to performactions of an administrative or record keeping nature (e.g., normal rounds or routineinspections).Further, as specified in IC HA5:The list need not include the Control Room if adequate engineered safety/design featuresare in place to preclude a Control Room evacuation due to the release of a hazardous gas.Such features may include, but are not limited to, capability to draw air from multiple airintakes at different and separate locations, inner and outer atmospheric boundaries, or thecapability to acquire and maintain positive pressure within the Control Room envelope.I [Document No.] Rev. 6 Page 274 of 276 ATTACHMENT 3Safe Operation & Shutdown Areas Tables R-3 & H-2 BasesCPNPP Table R-3 and H-2 BasesA review of station operating procedures identified the following mode dependent in-plantactions and associated areas that are required for normal plant operation, cooldown orshutdown:Location- Modes- Modes-Safe Shutdown Area 1, 2 3, 4, 5, or 6Charging Pump Rooms SDC Equipment. Shut Down Cooling (SDC)-No entry required -No entry requiredInventory Control Equipment Inventory Control Equipment-Entry required during pump -Entry required during pumpstarts and stops starts and stopsReactivity Control.-Entry required during pumpstarts and stops-Containment Spray Pumps A Post-accident Containment Post-accident Containmentand B Pressure Control Pressure Control (modes 3 and-No entry required 4)-No entry required-SI Pumps A and B Post-accident ECCS Post-accident ECCS-No entry required -No entry required-Residual Heat Removal Pumps Post-accident ECCS Decay Heat Removal (Modes 4, 5,A and B -No entry required and 6-Entry required for pumpsstarts and stops-CVCS Valve Rooms, Auxiliary Inventory Control Equipment Inventory Control EquipmentBuilding 810' and 822' -Entry required during -Entry required during pumppump starts and stops starts and stopsReactivity Control. Reactivity Control.-Entry required during -Entry required during pumppump starts and stops starts and stops-Service Water Intake Structure Ultimate Heat Sink Equipment for Ultimate Heat Sink Equipment forHabitability Control, Habitability Control, ContainmentContainment Temperature, Temperature, and Shutdownand Shutdown Cooling Cooling-No entry required No entry required1E Switchgear Rooms 810', Electrical Power. Electrical Power.832', and 852' -Entry required for manual -Entry required for manualbreaker manipulations on breaker manipulations oncomponent operations, component operationsreactor startup andshutdownControl Building 807'Cable Electrical Power. Electrical Power.Spreading Room -No entry required -No entry requiredControl Building 792' UPS and Electrical Power. Electrical Power.Battery Rooms -No entry requiredEmergency Diesel Generators A Electrical Power. Electrical Power.& B -No entry required -No entry requiredEmergency Diesel Generators Electrical Power. Electrical Power.Day Tank Rooms -No entry required -No entry requiredControl Building 830' Control Continuously occupied, Continuously occupied, capableRoom capable of Ventilation Isolation of Ventilation Isolation mode,mode, covered under H-6 covered under H-6Control Building 840' Technical -No entry required No entry requiredSupport Center, [Document No.] Rev. 6 Page 275 of 276 ATTACHMENT 3Safe Operation & Shutdown Areas Tables R-3 & H-2 BasesLocation- Modes- Modes-Safe Shutdown Area 1, 2 3, 4, 5, or 6Control Building 778' Safety -No entry required -No entry requiredChiller RoomsAuxiliary Building 790' -No entry required -No entry requiredAuxiliary Building 810' other -No entry required -No entry requiredthan CVCS Valave Rooms andCharging Pump RoomsAuxiliary Building 830' -No entry required -No entry requiredAuxiliary Building 852' -No entry required -No entry requiredAuxiliary Building 873' -No entry required -No entry requiredAuxiliary Building 886' -No entry required -No entry requiredSafeguards 790' -No entry required -No entry requiredSafeguards 810' -No entry required -No entry requiredSafeguards 831' -No entry required -No entry requiredSafeguards 852' -No entry required -No entry requiredSafeguards 873' -No entry required -No entry requiredTurbine Building Elevations -No entry required -No entry requiredAux. Feedwater Pump Rooms Steam Generator Heat Removal Steam Generator Heat RemovalA, B, and Turbine Driven -No entry required -No entry requiredTable R-3 & H-2 ResultsTable R-3/H-2 Safe Operation & Shutdown Rooms/AreasRoom/Area Mode ApplicabilityCharging Pump Rooms 1, 2, 3, 4, 5, 6CVCS Valve Rooms 1,2, 3, 4, 5, 61 E Switchgear Rooms AllRHR Pump Rooms 4, 5, 6Plant Operating Procedures Reviewed1.2.3.4.5.IPO-003A/BIPO-005A/BIPO-001A/BIPO-002A/BSOP-1 036. SOP-1047. SOP-102I [Document No.] Rev. 6 Page 276 of 276 ATTACHMENT 5 TO TXX-15101CPNPP RADIOLOGICAL EFFLUENTEAL VALUES(7 PAGES) | |||
EAL Section RRevision 6Table R-1 Effluent MonitorClassification Thresholds ReviewSubmitted By:Signature: /JACmrygWiecheringEieerg'ercyb~'nningDate:Review~ed By:Signature: JA .f4 .Hrf Nt 4,Jeffery Hull, Emergency IFlgrhiing ManagerDate: 6 Signature: 16y ( " aer-Ky Fishencord, E-m g-enc-y Pla'n'ningSignaure:cr1 IDe6 O' onnor, Radiation Protection ManagerSignature:Date: __2 3Date:-- ........Andre'a Lemons, Senior EngineerI of 706/18/2015 EAL Section R Revision 6Table R-I Effluent Monitor Classification ThresholdsTable R- 1 Effluent Monitor Classification ThresholdsType Release Point Monitor GE SAE ALERT UEGaseous Plant Vent X-RE-5567 A+B --- --------6.52E-4*PVG-384 +PVG-385 uCi/ccPlant Vent X-RE-5570 A+B 4.0E07 | |||
* 4.0E06 | |||
* 4.0E05 | |||
* 4.0E04*(WRGM) uCi/sec uCi/sec uCi/sec uCi/secPVF-684 + PVF-685 uMain Steam 90 uCi/cc 9 uCi/cc .9 uCi/cc 2 X HighMSL-a78 M-RE-2325 Alarm SetpointMSL-i79 u-RE-2326MSL-*180 U-RE-2327MSL-u8l P-RE-2328Liquid Liquid Waste 2 X HighLWE-076 X-RE-__53 Alarm SetpointService Water u-RE.4269 ---- ---- 2 X HighSSW..,65 u-RE-4270 Alarm SetpointSSW_-A66 I I I* Total of the two monitors equal this value or greater.Reference Material for Table R-1 Effluent Monitor Classification ThresholdsType Release Point Monitor High Source for Range**Alarm AlarmGaseous Plant Vent X-RE-5567 A or B 3.26E-4 CLI-744-3 I E-06 to I E-02PVG-384 +PVG-385 uCi/cc uCilccPlant Vent X-RE-5570 A or B 2.0E04 CLI-744-3 IE-06 to IE+05(WRGM) uCi/sec uCifccPVF-684 + PVF-685 I E-04 to 1E+12uCi/Sec***Main Steam u-RE-2325 2.OE-1 PC-11 IE-01 to IE+03MSL--78 u-RE-2326 uCi/cc uCi/ccMSL-U79MSL-8U0 u--RE-2327MSL-1u81 u-RE-2328Liquid Liquid Waste X-RE-5253 2.48E-04 CLI-744-1 1 E-05 to 5E-02LWE-076 uCi/sec uCi/ccService Water u-RE-4269 4.OE-06 PC-I1 IE-05 to 5E-02SSw%-65 u-RE-4270 uCi/sec uCi/ccSSW-3j66 I I II** Range DBD-EE-023 Specified Instrument Range unless noted.*** Range is from the PC-11.2 of 706/18/2015 EAL Section R Revision 6Table R-1 Effluent Monitor Classification ThresholdsDose Projection AssumptionsThe dose projections used the following:Only one unit was used to make dose projections.One mile was used for Exclusion Area Boundary.Reactor has been Shutdown for one hour.The event start time is same as Reactor Shutdown.Accident Type for Vent release was a LOCA.Filtration: NoStability Class: DWind Speed: 10 miles per hourPlant Vent UE AssumptionsPVG-384 +PVG-385 X-RE-5567 A+B (This monitor is used for an UE only)High Alarm Setpoint: 3.26E-4 uCi/ccPlant Vent Total: 6.52E-04 uCi/ccFlow Rate: 140,000 scfmReactor has been Shut Down for one hour.Projected Dose using CPAMPEDE 8.0PROJECTED DOSE I Mile 2 Mile 5 Mile 10 MileTEDE (Rem) 0.002 0.0008 0.0002 0.0001CDE (Rem) 0.009 0.003 0.0009 0.0003Plume Shine (R/hr) 0.0002 0.0001 0.0000 0.00003 of 706/18/2015 EAL Section R Revision 6Table R-1 Effluent Monitor Classification ThresholdsPlant Vent (WRGM) UE AssumptionsPlant Vent (WRGM) PVF-684 + PVF-685 X-RE-5570 A+BHigh Alarm Setpoint: 2.0E04 uCi/secCPAMPEDEValue U1 Value U22E4 uCi/Sec 2E4 uCi/SecPlant Vent Total: 4E04 uCi/secProjected Dose using CPAMPEDE 8.0PROJECTED DOSE I Mile 2 Mile 5 Mile 10 MileTEDE (Rem) 0.001 0.0005 0.0001 0.0000CDE (Rem) 0.006 0.002 0.0005 0.0002Plume Shine (R/hr) 0.00 0.00 0.00 0.00Plant Vent (WRGM) ALERTPlant Vent (WRGM) PVF-684 + PVF-68510 mRem TEDE or 50 mRem thyroid CDEHigh Alarm Setpoint: 2.0E04 uCi/secAssumptionsX-RE-5570 A+BCPAMPEDEValue UI I Value U22E5 uCi/Sec 2E5 uCi/SecPlant Vent Total: 4.0E05 uCi/secProjected Dose using CPAMPEDE 8.0PROJECTED DOSE I Mile 2 Mile 5 Mile 10 MileTEDE (Rem) 0.013 0.005 0.001 0.0005CDE (Rem) 0.061 0.022 0.006 0.002Plume Shine (R/hr) 0.001 0.00 0.00 0.004 of 706/18/2015 EAL Section R Revision 6Table R-1 Effluent Monitor Classification ThresholdsPlant Vent (WRGM) SAE AssumptionsPlant Vent (WRGM) PVF-684 + PVF-685 X-RE-5570 A+B100 mRem TEDE or 500 mRem thyroid CDEHigh Alarm Setpoint: 2.0E04 uCi/secICPAMPEDEValue U I Value U22E6 uCi/Sec 2E6 uCi/SecPlant Vent Total: 4E06 uCi/secProjected Dose using CPAMPEDE 8.0PROJECTED DOSE I Mile 2 Mile 5 Mile 10 MileTEDE (Rem) 0.132 0.047 0.012 0.005CDE (Rem) 0.577 0.208 0.054 0.020Plume Shine (R/hr) 0.010 0.003 0.001 0.00Plant Vent (WRGM) GE AssumptionsPlant Vent (WRGM) PVF-684 + PVF-685 X-RE-5570 A+B1000 mRem TEDE or 5000 mRem thyroid CDEHigh Alarm Setpoint: 2.0E04 uCi/secCPAMPEDEValue U 1 Value U22E7 uCi/Sec 2E7 uCi/SecPlant Vent Value: 4.0E07 uCi/secProjected Dose using CPAMPEDE 8.0PROJECTED DOSE 1 Mile 2 Mile 5 Mile 10 MileTEDE (Rem) 1.320 0.474 0.124 0.045CDE (Rem) 5.780 2.080 0.542 0.1 99Plume Shine (R/hr) 0.102 10.037 0.010 0.0045 of 706/18/2015 EAL Section R Revision 6Table R-1 Effluent Monitor Classification ThresholdsMain Steam UE AssumptionsThe setpoints differ on the main steam lines so an absolute number cannot be derived.Following are the current setpoints from the PC-1 1:Main Steam Line MonitorHigh (uCi/cc) [ Alert(uCi/ec)2.OE-01 1.OE-01Dose Projection AssumptionsThe dose projections used the following:Only one unit was used to make dose projections.One mile was used for Exclusion Area Boundary.Reactor has been Shutdown for one hour.The event start time is same as Reactor Shutdown.Accident Type for Main Steam Line monitors release was a SGTR.Flow Rate 120,000 lb/hrFiltration: NoStability Class: DWind Speed: 10 miles per hourMain Steam UE AssumptionsHigh Alarm Setpoint: 2.OE-1 uCi/ccMSL Total: 0.4 uCi/ccProjected Dose using CPAMPEDE 8.0PROJECTED DOSE 1 Mile 2 Mile 5 Mile 10 MileTEDE (Rem) 0.002 0.0006 0.0002 0.0001CDE (Rem) 0.020 0.007 0.002 0.0007Plume Shine (R/hr) 0.00 0.00 0.00 0.006 of 706/18/2015 EAL Section R Revision 6Table R-1 Effluent Monitor Classification ThresholdsMain Steam ALERT Assumptions10 mRem TEDE or 50 mRem thyroid CDEPlant Vent Value: 0.9 uCi/ccProjected Dose using CPAMPEDE 8.0PROJECTED DOSE I Mile 2 Mile 5 Mile 10 MileTEDE (Rem) 0.004 0.001 0.0004 0.0001CDE (Rem) 0.046 0.017 0.004 0.002Plume Shine (R/hr) 0.003 0.001 0.00 0.00Main Steam SAE Assumptions100 mRem TEDE or 500 mRem thyroid CDEPlant Vent Value: 9.0 uCi/ccProjected Dose using CPAMPEDE 8.0PROJECTED DOSE I Mile 2 Mile 5 Mile 10 MileTEDE (Rem) 0.041 0.015 0.004 0.001CDE (Rem) 0.461 0.166 0.043 0.016Plume Shine (R/hr) 0.002 0.001 10.00 0.00Main Steam GE Assumptions1000 mRem TEDE or 5000 mRem thyroid CDEPlant Vent Value: 90 uCi/ecProjected Dose using CPAMPEDE 8.0PROJECTED DOSE I Mile 2 Mile 5 Mile 10 MileTEDE (Rem) 0.405 0.146 0.038 0.014CDE (Rem) 4.610 1.660 0.433 0.158Plume Shine (R/hr) 0.020 0.007 10.002 0.0017 of 706/18/2015 ATTACHMENT 6 TO TXX-15101EMERGENCY ACTION LEVELWALLCHARTSFOR CPNPP(3 PAGES)}} | |||
Revision as of 00:00, 7 June 2018
| ML15191A164 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 06/24/2015 |
| From: | Luminant Generation Co, Luminant Power |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML15191A175 | List: |
| References | |
| CP-201500668, TXX-15101 | |
| Download: ML15191A164 (289) | |
Text
{{#Wiki_filter:ATTACHMENT 4 TO TXX-15101EMERGENCY ACTION LEVELTECHNICAL BASES(REDLINE AND STRIKEOUT VERSION)(276 PAGES) LuminantComanche Peak Nuclear Power PlantEPP-201Emergency Action Level Technical Bases DocumentPrepared by:Print NameSignatureDateTechnical Reviewer:Print NameSignatureDateReviewer:Print NameSignatureDateApproval:Print NameSignatureDateEffective Date:DRAFT D2 6/24/15I Page 1 of 276 I TABLE OF CONTENTSSECTION PAGE1.0 PURPOSE .............................................................................................................................. 32.0 DISCUSSION .......................................................................................................................... 32.1 Background ......................................................................................................................... 32.2 Fission Product Barriers .................................................................................................. 42.3 Fission Product Barrier Classification Criteria ................................................................ 42.4 EAL Organization ........................................................................................................ 52.5 Technical Bases Information ........................................................................................... 72.6 Operating M ode Applicability ......................................................................................... 82.7 Unit Designation .......................................................................................................... 83.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS .......................................... 93.1 General Considerations .................................................................................................. 93.2 Classification Methodology ........................................................................................... 1
04.0 REFERENCES
..................................................................................................................... 134.1 Developm ental ................................................................................................................... 134.2 Im plementing ..................................................................................................................... 135.0 DEINITIONS, ACRONYMS & ABBREVIATIONS ............................................................. 146.0 CPNPP TO NEI 99-01 Rev. 6 EAL CROSS-REFERENCE .............................................. 207.0 ATTACHM ENTS ................................................................................................................... 241 Emergency Action Level Technical Bases ........................................................... 25Cate-gory R Abnormal Rad Release / Rad Effluent .................................... 25Category E ISFSI ...................................................................................... 65Cate-gory C Cold Shutdown / Refueling System Malfunction ...................... 68Category H Hazards ...................................................................................... 112Cate-gory S System Malfunction ........... ; ........................................................ 155Cate-gory F Fission Product Barrier Degradation .......................................... 2002 Fission Product Barrier Loss / Potential LossMatrix and Bases .................................................................................................... 2053 Safe Operation & Shutdown Areas Tables R-3 & H-2 Bases ................................. 258Page 2 of 276 1.0 PURPOSEThis document provides an explanation and rationale for each Emergency Action Level (EAL)included in the EAL Upgrade Project for Comanche Peak Nuclear Power Plant (CPNPP). Itshould be used to facilitate review of the CPNPP EALs and provide historical documentationfor future reference. Decision-makers responsible for implementation of EPP-201,"Assessment of Emergency Action Levels, Emergency Classification and Plan Activation,"may use this document as a technical reference in support of EAL interpretation. Thisinformation may assist the Emergency Coordinator in making classifications, particularly thoseinvolving judgment or multiple events. The basis information may also be useful in trainingand for explaining event classifications to off-site officials.The expectation is that emergency classifications are to be made as soon as conditions arepresent and recognizable for the classification, but within 15 minutes or less in all cases ofconditions present. Use of this document for assistance is not intended to delay theemergency classification.Because the information in a basis document can affect emergency classification decision-making (e.g., the Emergency Coordinator refers to it during an event), the NRC staff expectsthat changes to the basis document will be evaluated in accordance with the provisions of 10CFR 50.54(q).2.0 DISCUSSION2.1 BackgroundEALs are the plant-specific indications, conditions or instrument readings that are utilized toclassify emergency conditions defined in the CPNPP Plant Radiological EmergencyResponse Plan (RERP).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) Revisions 4 and 5 were 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 address conditionsthat may be postulated to occur at permanently Defueled Stations and IndependentSpent Fuel Storage Installations (ISFSIs).* Simplifying the fission product barrier EAL threshold for a Site Area Emergency.Subsequently, Revision 6 of NEI 99-01 has been issued which incorporates resolutions tonumerous implementation issues including the NRC EAL Frequently Asked Questions (FAQs).Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Action Levelsfor Non-Passive Reactors," November 2012 (ADAMS Accession Number ML12326A805) (ref.4.1.1), CPNPP conducted an EAL implementation upgrade project that produced the EALsdiscussed hereinPage 3 of 276 2.2 Fission Product BarriersFission product barrier thresholds represent threats to the defense in depth design conceptthat precludes the release of radioactive fission products to the environment. This conceptrelies on multiple physical barriers, any one of which, if maintained intact, precludes therelease of significant amounts of radioactive fission products to the environment.Many of the EALs derived from the NEI methodology are fission product barrier thresholdbased. That is, the conditions that define the EALs are based upon thresholds that representthe loss or potential loss of one or more of the three fission product barriers. "Loss" and"Potential Loss" signify the relative damage and threat of damage to the barrier. A "Loss"threshold means the barrier no longer assures containment of radioactive materials. A"Potential Loss" threshold implies an increased probability of barrier loss and decreasedcertainty of maintaining the barrier.The primary fission product barriers are:A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains thefuel pellets.B. Reactor Coolant System (RCS): The RCS Barrier includes the RCS primary side and itsconnections up to and including the pressurizer safety and relief valves, and otherconnections up to and including the primary isolation valves.C. Containment (CNTMT): The Containment Barrier includes the containment building andconnections up to and including the outermost containment isolation valves. This barrieralso includes the main steam, feedwater, and blowdown line extensions outside thecontainment building up to and including the outermost secondary side isolation valve.Containment Barrier thresholds are used as criteria for escalation of the ECL from Alertto a Site Area Emergency or a General Emergency2.3 Fission Product Barrier Classification CriteriaThe following criteria are the bases for event classification related to fission product barrierloss or potential loss:Alert:Any loss or any potential loss of either Fuel Clad or RCS barrierSite 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 4 of 276 2.4 EAL OrganizationThe CPNPP EAL scheme includes the following features:0 Division of the EAL set into three broad groups:o EALs applicable under any 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, Hot Standby,Startup, or Power 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, Refueling orDefueled mode.The purpose of the groups is to avoid review of hot condition EALs when the plant is ina cold condition and avoid review of cold condition EALs when the plant is in a hotcondition. This approach significantly minimizes the total number of EALs that must bereviewed by the EAL-user for a given plant condition, reduces EAL-user reading burdenand, thereby, speeds identification of the EAL that applies to the emergency.0 Within each group, assignment of EALs to categories and subcategories:Category and subcategory titles are selected to represent conditions that are operationallysignificant to the EAL-user. The CPNPP EAL categories are aligned to and represent the NEI99-01"Recognition Categories." Subcategories are used in the CPNPP scheme as necessaryto further divide the EALs of a category into logical sets of possible emergency classificationthresholds. The CPNPP EAL categories and subcategories are listed below.Page 5 of 276 EAL Groups, Categories and SubcategoriesEAL Group/Category EAL SubcategoryAny Operating Mode:R -Abnormal Rad Levels / Rad Effluent I -Radiological Effluent2 -Irradiated Fuel Event3 -Area Radiation LevelsH -Hazards and Other Conditions 1 -SecurityAffecting Plant Safety 2 -Seismic Event3 -Natural or Technological Hazard4 -Fire5 -Hazardous Gases6 -Control Room Evacuation7 -Emergency Coordinator JudgmentE -ISFSI I -Confinement BoundaryHot Conditions:S -System Malfunction 1 -Loss of Emergency AC Power2 -Loss of Vital DC Power3 -Loss of Control Room Indications4 -RCS Activity5 -RCS Leakage6 -RPS Failure7 -Loss of Communications8 -Containment Failure9 -Hazardous Event Affecting Safety SystemsF -Fission Product Barrier Degradation NoneCold Conditions:C -Cold Shutdown / Refueling System 1 -RCS LevelMalfunction 2 -Loss of Emergency AC Power3 -RCS Temperature4 -Loss of Vital DC Power5 -Loss of Communications6 -Hazardous Event Affecting Safety SystemsThe primary tool for determining the emergency classification level is the EAL ClassificationMatrix. The user of the EAL Classification Matrix may (but is not required to) consult the EALTechnical Bases Document in order to obtain additional information concerning the EALsunder classification consideration. The user should consult Section 3.0 and Attachments I & 2of this document for such information.I Page 6 of 276 2.5 Technical Bases InformationEAL technical bases are provided in Attachment 1 for each EAL according to EAL group (Any,Hot, Cold), EAL category (R, C, H, S, E and F) and EAL subcategory. A summary explanationof each category and subcategory is given at the beginning of the technical bases discussionsof the EALs included in the category. For each EAL, the following information is provided:Category Letter & TitleSubcategiorv Number & TitleInitiating Condition (IC)Site-specific description of the generic IC given in NEI 99-01 Rev. 6.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 each EALidentifier:1. First character (letter): Corresponds to the EAL category as described above (R, C,H, S, E 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). If acategory does not have a subcategory, this character is assigned the number one(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)Exact wording of the EAL as it appears in the EAL Classification MatrixPage 7 of 276 Mode ApplicabilityOne or more of the following plant operating conditions comprise the mode to which eachEAL is applicable: 1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown, 5 -Cold Shutdown, 6 -Refueling, D -Defueled, or Any. (See Section 2.6 for operating modedefinitions)Definitions:If the EAL wording contains a defined term, the definition of the term is included in thissection. These definitions can also be found in Section 5.1.Basis:A Plant-Specific basis section that provides CPNPP-relevant information concerning theEAL. This is followed by a Generic basis section that provides a description of the rationalefor the EAL as provided in NEI 99-01 Rev. 6.CPNPP Basis Reference(s):Site-specific source documentation from which the EAL is derived.2.6 Operating Mode Applicability (ref. 4.1.8)1 Power OperationKeff greater than or equal to 0.99 and reactor thermal power greater than 5%2 StartupKeff greater than or equal t 0.99 and reactor thermal power-< 5%3 Hot StandbyKeff less than 0.99 and average coolant temperature greater than or equal to 3501F4 Hot ShutdownKeff less than 0.99 and average coolant temperature 3501F greater than Tavg greaterthan 200 OF and all reactor vessel head closure bolts fully tensioned5 Cold ShutdownKeff less than 0.99 and average coolant temperature < 200OF6 RefuelinqOne or more reactor vessel head closure bolts are less than fully tensionedD DefueledAll reactor fuel removed from reactor pressure vessel (full core off load during refuelingor extended outage).The plant operating mode that exists at the time that the event occurs (prior to any protectivesystem or operator action being initiated in response to the condition) should be compared tothe mode applicability of the EALs. If a lower or higher plant operating mode is reached beforethe emergency classification is made, the declaration shall be based on the mode that existedat the time the event occurred.2.7 Unit DesignationThe specific unit designator (1 or 2) is represented within these instructions by the symbol "u".The appropriate unit digit may be substituted for this symbol to obtain the unit specificequipment number (Example u-FK-121 represents 1-FK-121 for Unit 1 and 2-FK-121 for Unit2). For equipment or components that are common or non unit-specific the "X" designator isused. (Example X-RE-6272 represents a radiation monitor that is common to both units).Page 8 of 276 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS3.1 General ConsiderationsWhen making an emergency classification, the Emergency Coordinator must consider allinformation having a bearing on the proper assessment of an Initiating Condition (IC). Thisincludes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability,Notes, and the informing basis information. In the Recognition Category F matrices, EALs arebased on loss or potential loss of Fission Product Barrier Thresholds.3.1.1 Classification TimelinessNRC regulations require the licensee to establish and maintain the capability to assess,classify, and declare an emergency condition within 15 minutes after the availability ofindications to plant operators that an emergency action level has been exceeded and topromptly declare the emergency condition as soon as possible following identification of theappropriate emergency classification level. The NRC staff has provided guidance onimplementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, EmergencyPlanning for Nuclear Power Plants" (ref. 4.1.11).3.1.2 Valid IndicationsAll emergency classification assessments shall be based upon valid indications, reports orconditions. A valid indication, report, or condition, is one that has been verified throughappropriate means such that there is no doubt regarding the indicator's operability, thecondition's existence, or the report's accuracy. For example, verification could beaccomplished through an instrument channel check, response on related or redundantindicators, or direct observation by plant personnel.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) by directobservation by plant personnel, such that doubt related to the indicator's operability, thecondition's existence, or the report's accuracy is removed. Implicit in this definition is the needfor timely assessment.3.1.3 Imminent ConditionsFor ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), theEmergency Coordinator should not wait until the applicable time has elapsed, but shoulddeclare the event as soon as it is determined that the condition has exceeded, or will likelyexceed, the applicable time. If an ongoing radiological release is detected and the releasestart time is unknown, it should be assumed that the release duration specified in the IC/EALhas been exceeded, absent data to the contrary.3.1.4 Planned vs. Unplanned EventsA planned work activity that results in an expected event or condition which meets or exceedsan EAL does not warrant an emergency declaration provided that: 1) the activity proceeds asplanned, and 2) the plant remains within the limits imposed by the operating license. Suchactivities include planned work to test, manipulate, repair, maintain or modify a system orcomponent. In these cases, the controls associated with the planning, preparation andexecution of the work will ensure that compliance is maintained with all aspects of theoperating license provided that the activity proceeds and concludes as expected. Events orconditions of this type may be subject to the reporting requirements of 10 § CFR 50.72 (ref.4.1.4).Page 9 of 276 3.1.5 Classification Based on AnalysisThe assessment of some EALs is based on the results of analyses that are necessary toascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments,chemistry sampling, RCS leak rate calculation, etc.). For these EALs, the EAL wording or theassociated basis discussion will identify the necessary analysis. In these cases, the 15-minutedeclaration period starts with the availability of the analysis results that show the threshold tobe exceeded (i.e., this is the time that the EAL information is first available). The NRC expectslicensees to establish the capability to initiate and complete EAL-related analyses within areasonable period of time (e.g., maintain the necessary expertise on-shift).3.1.6 Emergency Coordinator JudgmentWhile the EALs have been developed to address a full spectrum of possible events andconditions which may warrant emergency classification, a provision for classification based onoperator/management experience and judgment is still necessary. The NEI 99-01 EALscheme provides the Emergency Coordinator with the ability to classify events and conditionsbased upon judgment using EALs that are consistent with the Emergency Classification Level(ECL) definitions (refer to Category H). The Emergency Coordinator will need to determine ifthe effects or consequences of the event or condition reasonably meet or exceed a particularECL definition. A similar provision is incorporated in the Fission Product Barrier Tables;judgment may be used to determine the status of a fission product barrier.3.2 Classification MethodologyTo make an emergency classification, the user will compare an event or condition (i.e., therelevant plant indications and reports) to an EAL(s) and determine if the EAL has been met orexceeded. The evaluation of an EAL must be consistent with the related Operating ModeApplicability and Notes. If an EAL has been met or exceeded, the associated IC is likewisemet, the emergency classification process "clock" starts, and the ECL must be declared inaccordance with plant procedures no later than fifteen minutes after the process "clock"started.When assessing an EAL that specifies a time duration for the off-normal condition, the "clock"for the EAL time duration runs concurrently with the emergency classification process "clock."For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01 (ref. 4.1.11).3.2.1 Classification of Multiple Events and ConditionsWhen multiple emergency events or conditions are present, the user will identify all met orexceeded EALs. The highest applicable ECL identified during this review is declared. Forexample:* If an Alert EAL and a Site Area Emergency EAL are met, whether at one unit or at twodifferent units, a Site Area Emergency should be declared.There is no "additive" effect from multiple EALs meeting the same ECL. For example:* If two Alert EALs are met, whether at one unit or at two different units, an Alert shouldbe declared.Related guidance concerning classification of rapidly escalating events or conditions isprovided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance forEmergency Notifications During Quickly Changing Events (ref. 4.1.2).Page 10 of 276 3.2.2 Consideration of Mode Changes During ClassificationThe mode in effect at the time that an event or condition occurred, and prior to any plant oroperator response, is the mode that determines whether or not an IC is applicable. If an eventor condition occurs, and results in a mode change before the emergency is declared, theemergency classification level is still based on the mode that existed at the time that the eventor condition was initiated (and not when it was declared). Once a different mode is reached,any new event or condition, not related to the original event or condition, requiring emergencyclassification should be evaluated against the ICs and EALs applicable to the operating modeat the time of the new event or condition.For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicablein the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is enteredduring the subsequent plant response. In particular, the fission product barrier EALs areapplicable only to events that initiate in the Hot Shutdown mode or higher.3.2.3 Classification of Imminent ConditionsAlthough EALs provide specific thresholds, the Emergency Coordinator must remain alert toevents or conditions that could lead to meeting or exceeding an EAL within a relatively shortperiod of time (i.e., a change in the ECL is IMMINENT). If, in the judgment of the EmergencyCoordinator, meeting an EAL is IMMINENT, the emergency classification should be made as ifthe EAL has been met. While applicable to all emergency classification levels, this approach isparticularly important at the higher emergency classification levels since it provides additionaltime for implementation of protective measures.3.2.4 Emergency Classification Level Upgrading and DowngradingAn ECL may be downgraded when the event or condition that meets the highest IC and EALno longer exists, and other site-specific downgrading requirements are met. If downgradingthe ECL is deemed appropriate, the new ECL would then be based on a lower applicable IC(s)and EAL(s). The ECL may also simply be terminated.As noted above, guidance concerning classification of rapidly escalating events or conditions isprovided in RIS 2007-02 (ref. 4.1.2).3.2.5 Classification of Short-Lived EventsEvent-based ICs and EALs define a variety of specific occurrences that have potential oractual safety significance. By their nature, some of these events may be short-lived and, thus,over before the emergency classification assessment can be completed. If an event occursthat meets or exceeds an EAL, the associated ECL must be declared regardless of itscontinued presence at the time of declaration. Examples of such events include anearthquake or a failure of the reactor protection system to automatically trip the reactorfollowed by a successful manual trip.3.2.6 Classification of Transient ConditionsMany of the ICs and/or EALs employ time-based criteria. These criteria will require that theIC/EAL conditions be present for a defined period of time before an emergency declaration iswarranted. In cases where no time-based criterion is specified, it is recognized that sometransient conditions may cause an EAL to be met for a brief period of time (e.g., a few secondsto a few minutes). The following guidance should be applied to the classification of theseconditions.Page 11 of 276 EAL momentarily met during expected plant response -In instances where an EAL is brieflymet during an expected (normal) plant response, an emergency declaration is not warrantedprovided that associated systems and components are operating as expected, and operatoractions are performed in accordance with procedures.EAL momentarily met but the condition is corrected prior to an emergency declaration -If anoperator takes prompt manual action to address a condition, and the action is successful incorrecting the condition prior to the emergency declaration, then the applicable EAL is notconsidered met and the associated emergency declaration is not required. For illustrativepurposes, consider the following example:An ATWS occurs and the high pressure ECCS systems fail to automatically start. RPVlevel rapidly decreases and the plant enters an inadequate core cooling condition (apotential loss of both the fuel clad and RCS barriers). If an operator manually starts ahigh pressure ECCS system in accordance with an EOP step and clears the inadequatecore cooling condition prior to an emergency declaration, then the classification shouldbe based on the ATWS only.It is important to stress that the 15-minute emergency classification assessment period(process clock) is not a "grace period" during which a classification may be delayed to allowthe performance of a corrective action that would obviate the need to classify the event.Emergency classification assessments must be deliberate and timely, with no undue delays.The provision discussed above addresses only those rapidly evolving situations when anoperator is able to take a successful corrective action prior to the Emergency Coordinatorcompleting the review and steps necessary to make the emergency declaration. Thisprovision is included to ensure that any public protective actions resulting from the emergencyclassification are truly warranted by the plant conditions.3.2.7 After-the-Fact Discovery of an Emergency Event or ConditionIn some cases, an EAL may be met but the emergency classification was not made at the timeof the event or condition. This situation can occur when personnel discover that an event orcondition existed which met an EAL, but no emergency was declared, and the event orcondition no longer exists at the time of discovery. This may be due to the event or conditionnot being recognized at the time or an error that was made in the emergency classificationprocess.In these cases, no emergency declaration is warranted; however, the guidance contained inNUREG-1022 (ref. 4.1.3) is applicable. Specifically, the event should be reported to the NRCin accordance with 10 CFR § 50.72 (ref. 4.1.4) within one hour of the discovery of theundeclared event or condition. The licensee should also notify appropriate State and localagencies in accordance with the agreed upon arrangements.3.2.8 Retraction of an Emergency DeclarationGuidance on the retraction of an emergency declaration reported to the NRC is discussed inNUREG-1022 (ref. 4.1.3).Page 12 of 276 1
4.0 REFERENCES
4.1 Developmental4.1.1 NEI 99-01 Revision 6, Methodology for the Development of Emergency ActionLevels for Non-Passive Reactors, ADAMS Accession Number ML12326A8054.1.2 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications DuringQuickly Changing Events, February 2, 2007.4.1.3 NUREG-1022 Event Reporting Guidelines: 10CFR50.72 and 50.734.1.4 10 § CFR 50.72 Immediate Notification Requirements for Operating NuclearPower Reactors4.1.5 10 § CFR 50.73 License Event Report System4.1.6 CPNPP Emergency Plan Appendix E, Complex and Owner Controlled Area4.1.7 CPNPP FSAR Section 2.1.1 Site Location and Description4.1.8 Technical Specifications Table 1.1-1 Modes4.1.9 OPT-408A/B Refueling Containment Penetration Verification4.1.10 ODA-207 Guidelines on the Preparation and Review of Operations Procedures4.1.11 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for NuclearPower Plants4.1.12 IPO-01OA/B Reactor Coolant System Reduced Inventory Operations4.1.13 Technical Specifications 3.9.44.1.14 CPNPP Offsite Dose Calculation Manual (ODCM)4.2 Implementing4.2.1 EPP-201, Assessment of Emergency Action Levels, Emergency Classificationand Plan Activation4.2.2 NEI 99-01 Rev. 6 to CPNPP EAL Comparison Matrix4.2.3 CPNPP EAL MatrixPage 13 of 276 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS5.1 Definitions (ref. 4.1.1 except as noted)Selected terms used in Initiating Condition and Emergency Action Level statements are set inall capital letters (e.g., ALL CAPS). These words are defined terms that have specificmeanings as used in this document. The definitions of these terms are provided below.AlertEvents are in process, 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 hostile action. Anyreleases are expected to be small fractions of the EPA Protective Action Guideline exposurelevels.Containment ClosureThe procedurally defined actions taken to secure containment and its associated structures,systems, and components as a functional barrier to fission product release under shutdownconditions. Containment closure means that all potential escape paths are closed or capableof being closed (ref.4.1.13).A. All penetrations providing direct access from Containment atmosphere to outsideatmosphere are closed except:" Penetrations with automatic valves capable of being closed by an operable CVI" Penetrations under administrative controls (e.g., Control Room notified and designatedperson to close if required by fuel handling accident)B. Equipment hatch is closed and held in place by 4 bolts, or is capable of being closed andheld in place by 4 boltsC. One emergency airlock door is closedD. One personnel airlock door is capable of being closed (ref. 4.1.9)EPA PAGsEnvironment Protection Agency Protective Action Guidelines. The EPA PAGs are expressed interms of dose commitment: 1 Rem TEDE or 5 Rem CDE Thyroid. Actual or projected offsiteexposures in excess of the EPA PAGs requires CPNPP to recommend protective actions forthe general public to offsite planning agencies.Exclusion Area BoundaryExclusion Area Boundary is a synonymous term for Site Boundary. CPNPP FSAR Section2.1.1.3 and Figure 2.1-2 define the Exclusion Area Boundary. This boundary is used forestablishing effluent release limits with respect to the requirements of IOCFR20 (ref. 4.1.7).See also CPNPP Emergency Plan Appendix E, Complex and Owner Controlled Area (ref.4.1.6) and CCNPP ODCM Section 5.0 Design Features (ref. 4.1.14).ExplosionA rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemicalreaction or overpressurization. A release of steam (from high energy lines or components) oran electrical component failure (caused by short circuits, grounding, arcing, etc.) should notPage 14 of 276 automatically be considered an explosion. Such events require a post-event inspection todetermine if the attributes of an explosion are present.FaultedThe term applied to a steam generator that has a steam leak on the secondary side ofsufficient size to cause an uncontrolled drop in steam generator pressure or the steamgenerator to become completely depressurized.FireCombustion characterized by heat and light. Sources of smoke such as slipping drive belts oroverheated electrical equipment do not constitute fires. Observation of flame is preferred butis NOT required if large quantities of smoke and heat are observed.FloodinaA condition where water is enteringq a room or area faster than installed equipment is capableof removal, resulting in a rise of water level within the room or area.General EmergencyEvents are in process or have occurred which involve actual or imminent substantial coredegradation or melting with potential for loss of containment integrity or hostile actions thatresult in an actual loss of physical control of the facility. Releases can be reasonably expectedto exceed EPA Protective Action Guideline exposure levels offsite for more than the immediatesite area.HostageA person(s) held as leverage against the station to ensure that demands will be met by thestation.Hostile ActionAn act toward CPNPP or its personnel that includes the use of violent force to destroyequipment, take hostages, and/or intimidate the licensee to achieve an end. This includesattack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices usedto deliver destructive force. Other acts that satisfy the overall intent may be included. Hostileaction should not be construed to include acts of civil disobedience or felonious acts that arenot part of a concerted attack on CPNPP. Non-terrorism-based EALs should be used toaddress such activities (i.e., this may include violent acts between individuals in the ownercontrolled area).Hostile ForceOne or more individuals who are engaged in a determined assault, overtly or by stealth anddeception, equipped with suitable weapons capable of killing, maiming, or causing destruction.ImminentThe trajectory of events or conditions is such that an EAL will be met within a relatively shortperiod of time regardless of mitigation or corrective actions.ImPede(d)Personnel access to a room or area is hindered to an extent that extraordinary measures arenecessary to facilitate entry of personnel into the affected room/area (e.g., requiring use ofprotective equipment, such as SCBAs, that is not routinely employed).Page 15 of 276 Independent Spent Fuel Storage Installation (ISFSI)A complex that is designed and constructed for the interim storage of spent nuclear fuel andother radioactive materials associated with spent fuel storage.MaintainTake appropriate action to hold the value of an identified parameter within specified limits.Normal LevelsSappliod to radiological lIdEALs, the highest reading in the past t'A'cty-foue orexcluding th.e current peak value.Owner Controlled AreaAs shown in CPNPP Emergency Plan Appendix E, Complex and Owner Controlled Area.ProjectileAn object directed toward a Nuclear Power Plant that could cause concern for its continuedoperability, reliability, or personnel safety.Protected AreaAn area encompassed by physical barriers and to which access is controlled. The ProtectedArea refers to the designated security area around the process buildings and is depicted inFSAR Figure 1.2-1 Plot Plan (ref. 4.1.7).RCS 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).Reduced InventoryPlant condition when fuel is in the reactor vessel and Reactor Coolant System level is < 80inches above core plate (829'8") (ref. 4.1.12).Refueling PathwayThe reactor refueling cavity, spent fuel pool and fuel transfer canal comprise the refuelingpathway.RupturedThe condition of a steam generator in which primary-to-secondary leakage is of sufficientmagnitude to require a safety injection.RestoreTake the appropriate action required to return the value of an identified parameter to theapplicable limits.Safety SystemA system required for safe plant operation, cooling down the plant and/or placing it in the coldshutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in IOCFR50.2):Those structures, systems and components that are relied upon to remain functional duringand following design basis events to assure:Page 16 of 276 (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 could resultin 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 degradation tothe level of safety of the plant. A security condition does not involve a hostile action.Site Area EmergencyEvents are in process or have occurred which involve actual or likely maior failures of plantfunctions needed for protection of the public or hostile actions that result in intentional damageor malicious acts; (1) toward site personnel or equipment that could lead to the likely failure ofor; (2) that prevent effective access to equipment needed for the protection of the public. Anyreleases are not expected to result in exposure levels which exceed EPA Protective ActionGuidelines exposure levels beyond the site boundary.Site BoundarySee EXCLUSION AREA BOUNDARYUnisolableAn open or breached system line that cannot be isolated, remotely or locally.UnplannedA parameter change or an event that is not 1) the result of an intended evolution or 2) anexpected plant response to a transient. The cause of the parameter change or event may beknown or unknown.Unusual EventEvents are in process or have occurred which indicate a potential degradation in the level ofsafety of the plant or indicate a security threat to facility protection has been initiated. Noreleases of radioactive material requiring offsite response or monitoring are expected unlessfurther degradation of safety systems occurs.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) by directobservation by plant personnel, such that doubt related to the indicator's operability, thecondition's existence, or the report's accuracy is removed. Implicit in this definition is the needfor timely assessment.Visible DamageDamage to a component or structure that is readily observable without measurements, testing,or analysis. The visual impact of the damage is sufficient to cause concern regarding theoperability or reliability of the affected component or structure.Page 17 of 276 5.2 Abbreviations/AcronymsOF ....................................................................................................... Degrees Fahrenheit0 .......................................................................................................................... D e g re e sAC ....................................................................................................... Alternating CurrentAPDG .......................................................................... Alternate Power Diesel GeneratorATW S ...................................................................... Anticipated Transient W ithout ScramCPNPP .................................................................. Com anche Peak Nuclear Power PlantCDE ....................................................................................... Com m itted Dose EquivalentCFR ..................................................................................... Code of Federal RegulationsCNTM T .......................................................................................................... Containm entCSFST ....................................................................... Critical Safety Function Status TreeDBA ............................................................................................... Design Basis AccidentDC ............................................................................................................... Direct CurrentEAL ............................................................................................. Em ergency Action LevelECCS ............................................................................ Em ergency Core Cooling SystemECL ................................................................................. Em ergency Classification LevelEO F .................................................................................. Em ergency O perations FacilityEO P ............................................................................... Em ergency Operating ProcedureEPA .............................................................................. Environm ental Protection AgencyERG ................................................................................ Em ergency Response G uidelineEPIP ................................................................ Em ergency Plan Im plem enting ProcedureESF ........................................................................................ Engineered Safety FeatureESW ........................................................................................ Em ergency Service W aterFAA ................................................................................. Federal Aviation Adm inistrationFBI ................................................................................... Federal Bureau of InvestigationFEMA ............................................................... Federal Em ergency M anagem ent AgencyFSAR .................................................................................... Final Safety Analysis ReportG E ..................................................................................................... General Em ergencyIC ......................................................................................................... Initiating ConditionIPEEE ................. Individual Plant Examination of External Events (Generic Letter 88-20)Keff ......................................................................... Effective Neutron M ultiplication FactorLCO .................................................................................. Lim iting Condition of OperationLER ............................................................................................... Licensee Event ReportLOCA ......................................................................................... Loss of Coolant AccidentLW R ................................................................................................... Light W ater ReactorM PC ................................... M axim um Perm issible Concentration/M ulti-Purpose Canisterm R, m Rem , m rem , m REM .............................................. m illi-Roentgen Equivalent M anM SL ........................................................................................................ M ain Steam LineM W .................................................................................................................... M egawattNEI ........................................................ ..................................... Nuclear Energy InstitutePage 18 of 276 NESP ................................................................... National Environm ental Studies ProjectNPP .................................................................................................. Nuclear Power PlantNRC ................................................................................ Nuclear Regulatory Com m issionNSSS ................................................................................ Nuclear Steam Supply SystemNO RAD ................................................... North Am erican Aerospace Defense Com m and(NO )UE ................................................................................ Notification of Unusual EventO BE ...................................................................................... O perating Basis EarthquakeO CA ............................................................................................... Owner Controlled AreaO DCM ............................................................................ Off-site Dose Calculation M anualO RO ................................................................................. Offsite Response OrganizationOTO ............................................................................... Off-Norm al O perating ProcedurePA .............................................................................................................. Protected AreaPAG ........................................................................................ Protective Action GuidelinePRA/PSA ..................... Probabilistic Risk Assessment / Probabilistic Safety AssessmentPW R ....................................................................................... Pressurized W ater ReactorPSIG ................................................................................ Pounds per Square Inch GaugeR ........................................................................................................................ RoentgenRCC ............................................................................................ Reactor Control ConsoleRCS ............................................................................................ Reactor Coolant SystemRem , rem , REM ....................................................................... Roentgen Equivalent M anRETS ......................................................... Radiological Effluent Technical SpecificationsRPS ....................................................................................... Reactor Protection SystemR(P)V ....................................................................................... Reactor (Pressure) VesselRVLIS ................................................................. Reactor Vessel Level Indicating SystemSAR ............................................................................................... Safety Analysis ReportSBO ......................................................................................................... Station BlackoutSCBA ....................................................................... Self-Contained Breathing ApparatusSG ......................................................................................................... Steam GeneratorSI .............................................................................................................. Safety InjectionO DCM ............................................................................. Offsite Dose Calculation M anualSPDS ........................................................................... Safety Param eter Display SystemSRO ............................................................................................ Senior Reactor O peratorTEDE ............................................................................... Total Effective Dose EquivalentTOAF .................................................................................................... Top of Active FuelTSC ... ...................................................................................... Technical Support CenterW O G ................................................................................... W estinghouse Ow ners GroupI Page 19 of 276 6.0 CPNPP-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCEThis cross-reference is provided to facilitate association and location of a CPNPP EAL withinthe NEI 99-01 IC/EAL identification scheme. Further information regarding the development ofthe CPNPP EALs based on the NEI guidance can be found in the EAL Comparison Matrix.CPNPP NEI 99-01 Rev. 6EAL IC ExampleEALRU1.1 AU1 1,2RU1.2 AU1 3RU2.1 AU2 IRA1.1 AA1 1RA1.2 AA1 2RAU.3 AA1 3RA1.4 AA1 4RA2.1 AA2 1RA2.2 AA2 2RA2.3 AA2 3RA3.1 AA3 1RA3.2 AA3 2RS1.1 AS1 1RS1.2 AS1 2RS1.3 AS1 3RS2.1 AS2 1RGI.1 AG1 1RG1.2 AG1 2RG1.3 AG1 3RG2.1 AG2 1CU1.1 CU1 1Page 20 of 276 CPNPP NEI 99-01 Rev. 6EAL IC ExampleEALCU1.2 CUI 2CU2.1 CU2 1CU3.1 CU3 1CU3.2 CU3 2CU4.1 CU4 1CU5.1 CU5 1,2,3CAI.1 CA1 1CAI.2 CA1 2CA2.1 CA2 1CA3.1 CA3 1,2CA6.1 CA6 1CS1.1 CS1 1CS1.2 CS1 2CS1.3 CS1 3CGI.1 CG1 2FAI.1 FA1 IFSI.1 FS1 1FG1.1 FG1 1HU1.1 HU1 1HU1.2 HUI 2HU1.3 HUI 3HU2.1 HU2 IHU3.1 HU3 1HU3.2 HU3 2HU3.3 HU3 3Page 21 of 276 CPNPP NEI 99-01 Rev. 6EAL IC ExampleEALHU3.4 HU3 4HU4.1 HU4 1HU4.2 HU4 2HU4.3 HU4 3HU4.4 HU4 4HU7.1 HU7 1HA1.1 HAl 1HA1.2 HAl 2HA5.1 HA5 IHA6.1 HA6 1HA7.1 HA7 1HSl.1 HS1 IHS6.1 HS6 1HS7.1 HS7 1HG1.1 HG1 1HG7.1 HG7 1su1.1 Sul 1SU3.1 SU2 1SU4.1 SU3 1SU4.2 SU3 2SU5.1 SU4 1,2,3SU6.1 SU5 ISU6.2 SU5 2SU7.1 SU6 1,2,3SU8.1 SU7 1Page 22 of 276 CPNPP NEI 99-01 Rev. 6EAL IC ExampleEALSA1.1 SA1 1SA3.1 SA2 1SA6.1 SA5 1SA9.1 SA9 1SS1.1 SS1 1SS2.1 SS8 1SS6.1 SS5 1SGl.1 SG1 1SG1.2 SG8 1EUI.1 E-HU1 1Page 23 of 276 7.0 ATTACHMENTS7.1 Attachment 1, Emergency Action Level Technical Bases7.2 Attachment 2, Fission Product Barrier Matrix and BasisPage 24 of 276 ATTACHMENT 1EAL BasesCategory 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 barriers becauseof the elevated potential for offsite radioactivity release. Degradation of fission productbarriers though is not always apparent via non-radiological symptoms. Therefore, directindication of elevated radiological effluents or area radiation levels are appropriate symptomsfor emergency classification.At lower levels, abnormal radioactivity releases may be indicative of a failure of containmentsystems or precursors to more significant releases. At higher release rates, offsite radiologicalconditions may result which require offsite protective actions. Elevated area radiation levels inplant may also be indicative of the failure of containment systems or preclude access to plantvital equipment necessary to ensure plant safety.Events of this category pertain to the following subcategories:1. Radiological EffluentDirect indication of effluent radiation monitoring systems provides a rapid assessmentmechanism to determine releases in excess of classifiable limits. Projected offsite doses,actual offsite field measurements or measured release rates via sampling indicate dosesor dose rates above classifiable limits.2. Irradiated Fuel EventConditions indicative of a loss of adequate shielding or damage to irradiated fuel maypreclude access to vital plant areas or result in radiological releases that warrantemergency classification.3. Area Radiation LevelsSustained general area radiation levels which may preclude access to areas requiringcontinuous occupancy also warrant emergency classification.I Page 25 of 276 1 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:R -Abnormal Rad Levels / Rad Effluent1 -Radiological EffluentRelease of gaseous or liquid radioactivity greater than 2 times theODCM limits for 60 minutes or longerEAL:RUI.1 Unusual EventReading on any Table R-1 effluent radiation monitor greater than column "UE" for greaterthan or equal to 60 min.(Notes 1, 2, 3)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the releaseduration has exceeded the specified time limit.Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path isisolated, the effluent monitor reading is no longer VALID for classification purposes.Table R-1 Effluent Monitor Classification ThresholdsRelease Point Monitor GE SAE Alert UEPlant Vent X-RE-5567............6.52E-4 1Cic/,mlPVG384 + PVG385 A + BPlant Vent(WRGM)X-RE-5570(WRGM) A + B 4.0E+7 ý.Ci/sec 4.OE+6 p.Ci/sec 4.0E+5 pCi/sec 4.OE+4 iiCi/sec0PVF684 + PVF685S Main SteamMSLu78 u-RE-2325MSLu79 u-RE-2326 90 pici/ml* 9.0 p.Cil/ml* 0.9 t4ci/ml* 2 x high alarmMSLu80 u-RE-2327 setpoint*MSLu81 u-RE-2328Liquid Waste X-RE-5253 2 x high alarmLWE-076 setpointService Water"J SSei65 u-RE-4269 2 x high alarmu-RE-4270 setpointSSWu66* with reactor shutdownMode Applicability:AllI Page 26 of 276 ATTACHMENT 1EAL BasesDefinition(s):NoneCPNPP Basis:The column "UE" gaseous and liquid release values in Table R-1 represent two times thealarm setpoint of the specified monitors. The setpoints are established to ensure the ODCMrelease limits are not exceeded. (ref. 1)Plant Vent Monitors sample both plant vent stacks prior to discharge to the environment. Theydetect normal operational levels of noble gases. The noble gas detectors (X-RE-5567A, B) canbe used as backups to the wide range gas monitors (X-RE-5570A, B). These monitorscommunicate with the RM-23s in the Control Room. Indication and annunciation are providedin the Control Room for alert and high radiation levels and monitor failure. (ref. 2)The WRGM system is a gaseous effluent monitoring system composed of two identicalmonitors used for detection of noble gas releases through the two plant vent stacks. Exhaustfrom the main turbine gland steam condenser exhauster is routed to the vent stacks formonitoring prior to release. Particulate and iodine grab samples may also be obtained from theWRGM. These monitors also initiate the automatic closure of the gas release valve in thewaste gas processing system on detection of high radiation. Indication and annunciation areprovided in the Control Room for alert and high radiation levels and monitor failure. (ref. 2)There are four online Main Steam Line Monitors (MSL) for each steam generator. Each oneconsists of a shielded, Category II seismic detector mounted adjacent to a main steam line, aremote RM-80 microprocessor and a remote customer interface junction box. The RM-80associated with the MSL monitor communicates with the PC-1I1 CRT console computer. (ref.2).Plant Liquid Waste Processing System (LWPS) discharge is continuously monitored by ashielded gamma sensitive (Nal(T1)) scintillation detector. When a LWPS discharge is required,normally locked-closed control valves can be opened directing flow through a path containing aradiation monitor (X-RE-5253) and a control valve which discharges waste to the circulatingwater discharge tunnel. The control valves are administratively controlled with a key-operatedswitch selectable to closed, automatic, or "key-held" open modes. In the automatic position,the valve will close on monitor high radiation alarm or monitor failure signals. Indication andannunciation are provided on the Waste Processing System (WPS) control panel for alert ormonitor failure alarm and in the Control Room for alert, high, and monitor failure alarms. (ref. 2)Service Water monitors are provided to monitor the Service Water System for radiation sinceleakage from radioactive fluid systems could cause potential radioactive leakage to theenvironment. A shielded gamma sensitive scintillation (Nal(T1)) detector is located in an off-line sample assembly downstream of each component cooling water heat exchanger tomonitor service water being discharged. Indication and annunciation are provided at theControl Room RMS console. (ref. 2)NEI 99-01 Basis:This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time(e.g., an uncontrolled release). It includes any gaseous or liquid radiological release,Page 27 of 276 ATTACHMENT IEAL Basesmonitored or un-monitored, including those for which a radioactivity discharge permit isnormally prepared.Nuclear power plants incorporate design features intended to control the release of radioactiveeffluents to the environment. Further, there are administrative controls established to preventunintentional releases, and to control and monitor intentional releases. The occurrence of anextended, uncontrolled radioactive release to the environment is indicative of degradation inthese features and/or controls.Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plant conditionsalone. The inclusion of both plant condition and radiological effluent EALs more fully addressesthe spectrum of possible accident events and conditions.Classification based on effluent monitor readings assumes that a release path to theenvironment is established. If the effluent flow past an effluent monitor is known to havestopped due to actions to isolate the release path, then the effluent monitor reading is nolonger valid for classification purposes.Releases should not be prorated or averaged. For example, a release exceeding 4 timesrelease limits for 30 minutes does not meet the EAL.I EAL--#-I-This EAL addresses normally occurring continuous radioactivity releases frommonitored gaseous or liquid effluent pathways.EAL #t2 This EAL addergenscs radioactioity leases that cause c-lucRt radiation monitoreadings to oxceed 2 times, the limit established by a radioactivity discharge permit. This EALWill typically be associated with planned batch rcleases fErom non continuous release pathways(e.g., r-adwaste, waste gas).EAL= 13 This E=AL addresses uncontrollcd gaseous or liquid relcascs that are detected bysample analyses or enviMromental sur~eys, particularly on unmonitored pathways (e.g., spoillsof radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.)-.Escalation of the emergency classification level would be via IC AAI-RPAI.CPNPP Basis Reference(s):1. CPNPP ODCM Unit 1 and 22. DBD-EE-023 Radiation Monitoring System3. EAL Section R Revision 6 Table R-1 Effluent Monitor Classification Thresholds Review4. NEI 99-01 AU1Page 28 of 276 1 ATTACHMENT 1EAL BasesCategory: R -Abnormal Rad Levels / Rad EffluentSubcategory: 1 -Radiological EffluentInitiating Condition: Release of gaseous or liquid radioactivity greater than 2 times theODCM limits for 60 minutes or longer.EAL:RUI.2 Unusual EventSample analysis for a gaseous or liquid release indicates a concentration or release rate> 2 x ODCM limits for greater than or equal to 60 min. (Notes 1, 2)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the releaseduration has exceeded the specified time limit.Mode Applicability:AllDefinition(s):NoneCPNPP Basis:NoneNEI 99-01 Basis:This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time(e.g., an uncontrolled release). It includes any gaseous or liquid radiological release,monitored or un-monitored, including those for which a radioactivity discharge permit isnormally prepared.Nuclear power plants incorporate design features intended to control the release of radioactiveeffluents to the environment. Further, there are administrative controls established to preventunintentional releases, and to control and monitor intentional releases. The occurrence of anextended, uncontrolled radioactive release to the environment is indicative of degradation inthese features and/or controls.Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plant conditionsalone. The inclusion of both plant condition and radiological effluent EALs more fully addressesthe spectrum of possible accident events and conditions.Cla~ssfication basecd on cfflucnt mon~itor readings assumes that a release path to henvionmet is established. if the effluent flo)w past an effluent monitor is known to havostopped due to actions to isolatc the release path, thcn the effluent moniRtorF reading is nolonger valid fra f purposes.Page 29 of 276 ATTACHMENT 1EAL BasesReleases should not be prorated or averaged. For example, a release exceeding 4 timesrelease limits for 30 minutes does not meet the EAL.EAL #1 This E.A.L , Addresses normally occurring continuous radioactiVity .eleas' .frommonioredgaseous or liquid effluent pathways.EAL-- -2 This EAL addresses radioactiVity releases that c-ause -ffluent rditomntrreadings to exceed 2 times the li~mit esta-bli-shed by a radioactiv~ity discharge permit. This ,A~LWil typically be asSocGiate ithpland- bhatch releases from non continuous release pathways(e.g., radwaste, waste gas).EAL-#3 --This EAL addresses uncontrolled gaseous or liquid releases that are detected bysample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spillsof radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.).Escalation of the emergency classification level would be via IC AA4-RA1.CPNPP Basis Reference(s):1. CPNPP ODCM Unit 1 and 22. NEI 99-01 AU1Page 30 of 276 ATTACHMENT 1EAL BasesCategory: R -Abnormal Rad Levels / Rad EffluentSubcategory: 1 -Radiological EffluentInitiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dosegreater than 10 mrem TEDE or 50 mrem thyroid CDEEAL:RA1.1 AlertReading on any Table R-1 effluent radiation monitor greater than column "ALERT" forgreater than or equal to 15 min. (Notes 1, 2, 3, 4)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the releaseduration has exceeded the specified time limit.Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path isisolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4 The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be usedfor emergency classification assessments until the results from a dose assessment using actualmeteorology are available.Table R-1 Effluent Monitor Classification ThresholdsRelease Point Monitor GE SAE Alert UEPlant Vent X-RE-5567.............6.52E-4 pCi/mlPVG384 + PVG385 A + BPlant Vent(WRGM)X-RE-5570(WRGM) X- 7 A + B 4.OE+7 ýCi/sec 4.OE+6 4.OE+5 pci/sec 4.OE+4 pCi/sec0 PVF684 + PVF685' Main SteamMSLu78 u-RE-2325MSLu79 u-RE-2326 90 pci/ml* 9.0 pci/ml* 0.9 j, ci/ml* 2xhigh alarmMSLu8O u-RE-2327MSLu81 u-RE-2328Liquid Waste X-RE-5253 2 x high alarmLWE-076 setpointM Servic Water u-RE-4269 2 x high alarmSSWu65u-RE-4270 setpointSSWu66* with reactor shutdownMode Applicability:AllI Page 31 of 276 ATTACHMENT 1EAL BasesDefinition(s):NoneCPNPP Basis:This EAL address gaseous radioactivity releases, that for whatever reason, cause effluentradiation monitor readings corresponding to site boundary doses that exceed either:0 10 mRem TEDE* 50 mRem CDE ThyroidThe column "ALERT" gaseous effluent release values in Table R-1 correspond to calculateddoses of 1 % (10% of the SAE thresholds) of the EPA Protective Action Guidelines (TEDE orCDE Thyroid) (ref. 1).NEI 99-01 Basis:This IC addresses a release of gaseous or liquid radioactivity that results in projected or actualoffsite doses greater than or equal to 1 % of the EPA Protective Action Guides (PAGs). Itincludes both monitored and un-monitored releases. Releases of this magnitude represent anactual or potential substantial degradation of the level of safety of the plant as indicated by aradiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolledrelease).Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plant conditionsalone. The inclusion of both plant condition and radiological effluent EALs more fully addressesthe spectrum of possible accident events and conditions.The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDEwas established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to theenvironment is established. If the effluent flow past an effluent monitor is known to havestopped due to actions to isolate the release path, then the effluent monitor reading is nolonger valid for classification purposes.Escalation of the emergency classification level would be via IC ASI-RSI.CPNPP Basis Reference(s):1. EAL Section R Revision 6 Table R-1 Effluent Monitor Classification Thresholds Review2. NEI 99-01 AA1Page 32 of 276 ATTACHMENT IEAL BasesCategory: R -Abnormal Rad Levels / Rad EffluentSubcategory: 1 -Radiological EffluentInitiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dosegreater than 10 mrem TEDE or 50 mrem thyroid CDEEAL:RA1.2 AlertDose assessment using actual meteorology indicates doses greater than 10 mrem TEDEor 50 mrem thyroid CDE at or beyond the EXCLUSION AREA BOUNDARY (Notes 3, 4)Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path isisolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: The pre-calculated effluent monitor values presented in EALs RAI.1, RS1.1 and RGI.1 should be usedfor emergency classification assessments until the results from a dose assessment using actualmeteorology are available.Mode Applicability:AllDefinition(s):EXCLUSION AREA BOUNDARY- Exclusion Area Boundary is a synonymous term for SiteBoundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion AreaBoundary. This boundary is used for establishing effluent release limits with respect to therequirements of 1OCFR20. See also CPNPP Emergency Plan Appendix E, Complex andOwner Controlled Area and CCNPP ODCM Section 5.0 Design Features.CPNPP Basis:Dose assessments are performed by computer-based method (ref. 1)NEI 99-01 Basis:This IC addresses a release of gaseous or liquid radioactivity that results in projected or actualoffsite doses greater than or equal to 1 % of the EPA Protective Action Guides (PAGs). Itincludes both monitored and un-monitored releases. Releases of this magnitude represent anactual or potential substantial degradation of the level of safety of the plant as indicated by aradiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolledrelease).Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plant conditionsalone. The inclusion of both plant condition and radiological effluent EALs more fully addressesthe spectrum of possible accident events and conditions.The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDEwas established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to theenvironment is established. If the effluent flow past an effluent monitor is known to havePage 33 of 276 ATTACHMENT 1EAL Basesstopped due to actions to isolate the release path, then the effluent monitor reading is nolonger valid for classification purposes.Escalation of the emergency classification level would be via IC AS-14RSI.CPNPP Basis Reference(s):1. EPP-303 Operation of Computer Based Dose Assessment System2. NEI 99-01 AA1Page 34 of 276 ATTACHMENT 1EAL BasesCategory: R -Abnormal Rad Levels / Rad EffluentSubcategory: I -Radiological EffluentInitiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dosegreater than 10 mrem TEDE or 50 mrem thyroid CDEEAL:RA1.3 AlertAnalysis of a liquid effluent sample indicates a concentration or release rate that wouldresult in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond theEXCLUSION AREA BOUNDARY for 60 min. of exposure (Notes 1, 2)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the releaseduration has exceeded the specified time limit.Mode Applicability:AllDefinition(s):EXCLUSION AREA BOUNDARY -Exclusion Area Boundary is a synonymous term for SiteBoundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion AreaBoundary. This boundary is used for establishing effluent release limits with respect to therequirements of 10CFR20. See also CPNPP Emergency Plan Appendix E, Complex andOwner Controlled Area and CCNPP ODCM Section 5.0 Design Features.CPNPP Basis:Dose assessments based on liquid releases are performed per Offsite Dose CalculationManual (ref. 1).NEI 99-01 Basis:This IC addresses a release of gaseous or liquid radioactivity that results in projected or actualoffsite doses greater than or equal to 1 % of the EPA Protective Action Guides (PAGs). Itincludes both monitored and un-monitored releases. Releases of this magnitude represent anactual or potential substantial degradation of the level of safety of the plant as indicated by aradiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolledrelease).Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plant conditionsalone. The inclusion of both plant condition and radiological effluent EALs more fully addressesthe spectrum of possible accident events and conditions.The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDEwas established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Page 35 of 276 ATTACHMENT 1EAL BasesClassification based on effluent monitor readings assumes that a release path to theenvironment is established. If the effluent flow past an effluent monitor is known to havestopped due to actions to isolate the release path, then the effluent monitor reading is nolonger valid for classification purposes.Escalation of the emergency classification level would be via IC AS4RS1.CPNPP Basis Reference(s):1. CPNPP Offsite Dose Calculation Manual2. NEI 99-01 AA1Page 36 of 276 ATTACHMENT 1EAL BasesCategory: R -Abnormal Rad Levels I Rad EffluentSubcategory: 1 -Radiological EffluentInitiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dosegreater than 10 mrem TEDE or 50 mrem thyroid CDEEAL:RA1.4 AlertField survey results indicate EITHER of the following at or beyond the EXCLUSION AREABOUNDARY:" Closed window dose rates greater than 10 mR/hr expected to continue for greaterthan or equal to 60 min." Analyses of field survey samples indicate thyroid CDE greater than 50 mrem for 60min. of inhalation.(Notes 1,2)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the releaseduration has exceeded the specified time limit.Mode Applicability:AllDefinition(s):EXCLUSION AREA BOUNDARY- Exclusion Area Boundary is a synonymous term for SiteBoundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion AreaBoundary. This boundary is used for establishing effluent release limits with respect to therequirements of 10CFR20. See also CPNPP Emergency Plan Appendix E, Complex andOwner Controlled Area and CCNPP ODCM Section 5.0 Design Features.CPNPP Basis:EPP-309 Onsite/In-Plant Radiological Surveys and Offsite Radiological Monitoring providesguidance for emergency or post-accident radiological environmental monitoring (ref. 1).NEI 99-01 Basis:This IC addresses a release of gaseous or liquid radioactivity that results in projected or actualoffsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). Itincludes both monitored and un-monitored releases. Releases of this magnitude represent anactual or potential substantial degradation of the level of safety of the plant as indicated by aradiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolledrelease).Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plant conditionsalone. The inclusion of both plant condition and radiological effluent EALs more fully addressesPage 37 of 276 ATTACHMENT 1EAL Basesthe spectrum of possible accident events and conditions.The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDEwas established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to theenvironment is established. If the effluent flow past an effluent monitor is known to havestopped due to actions to isolate the release path, then the effluent monitor reading is nolonger valid for classification purposes.Escalation of the emergency classification level would be via IC AS4RSI.CPNPP Basis Reference(s):1. EPP-309 Onsite/In-Plant Radiological Surveys and Offsite Radiological Monitoring2. NEI 99-01 AA1Page 38 of 276 1 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:R -Abnormal Rad Levels / Rad Effluent1 -Radiological EffluentRelease of gaseous radioactivity resulting in offsite dose greater than100 mrem TEDE or 500 mrem thyroid CDEEAL:RSI.1 Site Area EmergencyReading on any Table R-1 effluent radiation monitor greater than column "SAE" forgreater than or equal to 15 min.(Notes 1, 2, 3, 4)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the releaseduration has exceeded the specified time limit.Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path isisolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: The pre-calculated effluent monitor values presented in EALs RAI.1, RSI.1 and RGI.1 should be usedfor emergency classification assessments until the results from a dose assessment using actualmeteorology are available.Table R-1 Effluent Monitor Classification ThresholdsRelease Point Monitor GE SAE Alert UEPlant Vent X-RE-5567PVG384 + PVG385 A + BPlant Vent(WRGM)X-RE-5570(WRGM) +- 7 A + B 4.OE+7 [.Ci/sec 4.OE+6 ýtCi/sec 4.0E+5 pCi/sec 4.0E+4 p.Ci/sec0 PVF684 + PVF685C Main SteamMSLu78 u-RE-2325MSLu79 u-RE-2326 90 [Ci/ml* 9.0 [.ci/ml* 0.9 j.ci/ml* 2 x high alarmMSLu80 u-RE-2327 setpoint*MSLu81 u-RE-2328Liquid Waste X-RE-5253 ...............-2 x high alarmLWE-076 setpointr" Service Water* Srcu-RE-4269 2 x high alarmS SSWu65 ----.--.-----.2x.hgh.larSSw65 u-RE-4270 setpointSSWu66* with reactor shutdownMode Applicability:AllDefinition(s):Page 39 of 276 ATTACHMENT 1EAL BasesNoneCPNPP Basis:This EAL address gaseous radioactivity releases, that for whatever reason, cause effluentradiation monitor readings corresponding to site boundary doses that exceed either:* 100 mRem TEDE0 500 mRem CDE ThyroidThe column "SAE" gaseous effluent release value in Table R-1 corresponds to calculateddoses of 10% of the EPA Protective Action Guidelines (TEDE or CDE Thyroid) (ref. 1).NEI 99-01 Basis:This IC addresses a release of gaseous radioactivity that results in projected or actual offsitedoses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includesboth monitored and un-monitored releases. Releases of this magnitude are associated withthe failure of plant systems needed for the protection of the public.Radiological effluent EALs are also-included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plant conditionsalone. The inclusion of both plant condition and radiological effluent EALs more fully addressesthe spectrum of possible accident events and conditions.The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDEwas established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to theenvironment is established. If the effluent flow past an effluent monitor is known to havestopped due to actions to isolate the release path, then the effluent monitor reading is nolonger valid for classification purposes.Escalation of the emergency classification level would be via IC AG4RGI.CPNPP Basis Reference(s):1. EAL Section R Revision 6 Table R-1 Effluent Monitor Classification Thresholds Review2. NEI 99-01 AS1I Page 40 of 276 ATTACHMENT IEAL BasesCategory: R -Abnormal Rad Levels I Rad EffluentSubcategory: 1 -Radiological EffluentInitiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than100 mrem TEDE or 500 mrem thyroid CDEEAL:RS1,2 Site Area EmergencyDose assessment using actual meteorology indicates doses greater than 100 mrem TEDEor500 mrem thyroid CDE at or beyond the EXCLUSION AREA BOUNDARY (Notes 3, 4)Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path isisolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: The pre-calculated effluent monitor values presented in EALs RAI.1, RSI.1 and RG1.1 should be usedfor emergency classification assessments until the results from a dose assessment using actualmeteorology are available.Mode Applicability:AllDefinition(s):EXCLUSION AREA BOUNDARY -Exclusion Area Boundary is a synonymous term for SiteBoundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion AreaBoundary. This boundary is used for establishing effluent release limits with respect to therequirements of 10CFR20. See also CPNPP Emergency Plan Appendix E, Complex andOwner Controlled Area and CCNPP ODCM Section 5.0 Design Features.CPNPP Basis:Dose assessments are performed by computer-based method (ref. 1)NEI 99-01 Basis:This IC addresses a release of gaseous radioactivity that results in projected or actual offsitedoses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includesboth monitored and un-monitored releases. Releases of this magnitude are associated withthe failure of plant systems needed for the protection of the public.Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plant conditionsalone. The inclusion of both plant condition and radiological effluent EALs more fully addressesthe spectrum of possible accident events and conditions.The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDEwas established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to theenvironment is established. If the effluent flow past an effluent monitor is known to havestopped due to actions to isolate the release path, then the effluent monitor reading is noI Page 41 of 276 ATTACHMENT 1EAL Baseslonger valid for classification purposes.Escalation of the emergency classification level would be via IC AG-I-RGI.CPNPP Basis Reference(s):1. EPP-303 Operation of Computer Based Dose Assessment System2. NEI 99-01 AS1Page 42 of 276 ATTACHMENT 1EAL BasesCategory: R -Abnormal Rad Levels / Rad EffluentSubcategory: 1 -Radiological EffluentInitiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than100 mrem TEDE or 500 mrem thyroid CDEEAL:RS1.3 Site Area EmergencyField survey results indicate EITHER of the following at or beyond the EXCLUSION AREABOUNDARY:" Closed window dose rates greater than 100 mR/hr expected to continue for greaterthan or equal to 60 min." Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for 60min. of inhalation.(Notes 1, 2)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the releaseduration has exceeded the specified time limit.Mode Applicability:AllDefinition(s):EXCLUSION AREA BOUNDARY -Exclusion Area Boundary is a synonymous term for SiteBoundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion AreaBoundary. This boundary is used for establishing effluent release limits with respect to therequirements of 10CFR20. See also CPNPP Emergency Plan Appendix E, Complex andOwner Controlled Area and CCNPP ODCM Section 5.0 Design Features.CPNPP Basis:EPP-309 Onsite/In-Plant Radiological Surveys and Offsite Radiological Monitoring providesguidance for emergency or post-accident radiological environmental monitoring (ref. 1).NEI 99-01 Basis:This IC addresses a release of gaseous radioactivity that results in projected or actual offsitedoses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includesboth monitored and un-monitored releases. Releases of this magnitude are associated withthe failure of plant systems needed for the protection of the public.Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plant conditionsalone. The inclusion of both plant condition and radiological effluent EALs more fully addressesthe spectrum of possible accident events and conditions.Page 43 of 276 ATTACHMENT 1EAL BasesThe TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDEwas established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.ClassificAtion on effluent readings as.sum.es. that a releaso path to theenViGronment is established. if the cifluent flo)W past an effluent moenitor is kno)Wn to haVostepped due to- ac-tions to isolate the relcasc path, then the enffluent monitor rea;ding is nolonger valid for classification purposes.Escalation of the emergency classification level would be via IC AG4-RGI.CPNPP Basis Reference(s):1. EPP-309 Onsite/In-Plant Radiological Surveys and Offsite Radiological Monitoring2. NEI 99-01 AS1Page 44 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:R -Abnormal Rad Levels / Rad Effluent1 -Radiological EffluentRelease of gaseous radioactivity resulting in offsite dose greater than1,000 mrem TEDE or 5,000 mrem thyroid CDEEAL:RGI.1 General EmergencyReading on any Table R-1 effluent radiation monitor greater than column "GE" for greaterthan or equal to 15 min.(Notes 1,2, 3, 4)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the releaseduration has exceeded the specified time limit.Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path isisolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RSI.1 and RGI.1 should be usedfor emergency classification assessments until the results from a dose assessment using actualmeteorology are available.Table R-1 Effluent Monitor Classification ThresholdsRelease Point Monitor GE SAE Alert UEPlant Vent X-RE-5567.............6.52E-4 pCi/mlPVG384 + PVG385 A + BPlant Vent(WRGM)X-RE-5570S(WRGM) X-E-7 A + B 4.OE+7 luCi/sec 4.OE+6 liCi/sec 4.OE+5 [iCi/sec 4.OE+4 pCi/sec0 PVF684 + PVF685Main SteamMSLu78 u-RE-2325MSLu79 u-RE-2326 90 p.ci/ml* 9.0 ILci/ml* 0.9 pLCi/ml* 2 x high alarmMSLu80 u-RE-2327 setpoint*MSLu81 u-RE-2328Liquid Waste X-RE-5253 ...............-2 x high alarmLWE-076 setpoint"7 Service Wateru-RE-4269 2 x high alarm----..-------..x.hgh.larSSu65 u-RE-4270 setpointSSWu66* with reactor shutdownMode Applicability:AllPage 45 of 276 ATTACHMENT 1EAL BasesDefinition(s):NoneCPNPP Basis:This EAL address gaseous radioactivity releases, that for whatever reason, cause effluentradiation monitor readings corresponding to site boundary doses that exceed either:* 1000 mRemTEDE* 5000 mRem CDE ThyroidThe column "GE" gaseous effluent release values in Table R-1 correspond to calculated dosesof 100% of the EPA Protective Action Guidelines (TEDE or CDE Thyroid) (ref. 1).NEI 99-01 Basis:This IC addresses a release of gaseous radioactivity that results in projected or actual offsitedoses greater than or equal to the EPA Protective Action Guides (PAGs). It includes bothmonitored and un-monitored releases. Releases of this magnitude will require implementationof protective actions for the public.Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plant conditionsalone. The inclusion of both plant condition and radiological effluent EALs more fully addressesthe spectrum of possible accident events and conditions.The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE wasestablished in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to theenvironment is established. If the effluent flow past an effluent monitor is known to havestopped due to actions to isolate the release path, then the effluent monitor reading is nolonger valid for classification purposes.CPNPP Basis Reference(s):1. EAL Section R Revision 6 Table R-1 Effluent Monitor Classification Thresholds Review2. NEI 99-01 AG1I Page 46 of 276 ATTACHMENT 1EAL BasesCategory: R -Abnormal Rad Levels / Rad EffluentSubcategory: 1 -Radiological EffluentInitiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than1,000 mrem TEDE or 5,000 mrem thyroid CDEEAL:RGI.2 General EmergencyDose assessment using actual meteorology indicates doses greater than 1,000 mremTEDE or5,000 mrem thyroid CDE at or beyond the EXCLUSION AREA BOUNDARY (Notes 3, 4)Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path isisolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: The pre-calculated effluent monitor values presented in EALs RAI.1, RS1.1 and RGI.1 should be usedfor emergency classification assessments until the results from a dose assessment using actualmeteorology are available.Mode Applicability:AllDefinition(s):EXCLUSION AREA BOUNDARY- Exclusion Area Boundary is a synonymous term for SiteBoundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion AreaBoundary. This boundary is used for establishing effluent release limits with respect to therequirements of 10CFR20. See also CPNPP Emergency Plan Appendix E, Complex andOwner Controlled Area and CCNPP ODCM Section 5.0 Design Features.CPNPP Basis:Dose assessments are performed by computer-based method (ref. 1)NEI 99-01 Basis:This IC addresses a release of gaseous radioactivity that results in projected or actual offsitedoses greater than or equal to the EPA Protective Action Guides (PAGs). It includes bothmonitored and un-monitored releases. Releases of this magnitude will require implementationof protective actions for the public.Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plant conditionsalone. The inclusion of both plant condition and radiological effluent EALs more fully addressesthe spectrum of possible accident events and conditions.The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE wasestablished in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to theenvironment is established. If the effluent flow past an effluent monitor is known to havePage 47 of 276 ATTACHMENT 1EAL Basesstopped due to actions to isolate the release path, then the effluent monitor reading is nolonger valid for classification purposes.CPNPP Basis Reference(s):1. EPP-303 Operation of Computer Based Dose Assessment System2. NEI 99-01 AG1Page 48 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:R -Abnormal Rad Levels / Rad Effluent1 -Radiological EffluentRelease of gaseous radioactivity resulting in offsite dose greater than1,000 mrem TEDE or 5,000 mrem thyroid CDEEAL:RG1.3 General EmergencyField survey results indicate EITHER of the following at or beyond the EXCLUSION AREABOUNDARY:" Closed window dose rates greater than 1,000 mR/hr expected to continue for greaterthan or equal to 60 min." Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrem for60 min. of inhalation.(Notes 1, 2)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the releaseduration has exceeded the specified time limit.Mode Applicability:AllDefinition(s):EXCLUSION AREA BOUNDARY- Exclusion Area Boundary is a synonymous term for SiteBoundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion AreaBoundary. This boundary is used for establishing effluent release limits with respect to therequirements of 10CFR20. See also CPNPP Emergency Plan Appendix E, Complex andOwner Controlled Area and CCNPP ODCM Section 5.0 Design Features.CPNPP Basis:EPP-309 Onsite/In-Plant Radiological Surveys and Offsite Radiological Monitoring providesguidance for emergency or post-accident radiological environmental monitoring (ref. 1).NEI 99-01 Basis:This IC addresses a release of gaseous radioactivity that results in projected or actual offsitedoses greater than or equal to the EPA Protective Action Guides (PAGs). It includes bothmonitored and un-monitored releases. Releases of this magnitude will require implementationof protective actions for the public.Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plant conditionsalone. The inclusion of both plant condition and radiological effluent EALs more fully addressesthe spectrum of possible accident events and conditions.Page 49 of 276 ATTACHMENT 1EAL BasesThe TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE wasestablished in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a rolcase path to theenvionmet is established. if the effluent flow past an cifluent monitor is known to aestopped duo to cin to- isolate the release path, then the effluont moenitor reading is nolo~nger valid for clasSiffication purposes.CPNPP Basis Reference(s):1. EPP-309 Onsite/In-Plant Radiological Surveys and Offsite Radiological Monitoring2. NEI 99-01 AG1Page 50 of 276 1 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:EAL:R -Abnormal Rad Levels / Rad Effluent2 -Irradiated Fuel EventUnplanned loss of water level above irradiated fuelRU2.1 Unusual EventUNPLANNED water level drop in the REFUELING PATHWAY as indicated by low waterlevel alarm or indicationANDUNPLANNED rise in corresponding area radiation levels as indicated by any Table R-2area radiation monitorsTable R-2 SFP & Refueling Cavity Area Radiation MonitorsSFP:* SFP001, LRAM SFP 2 E WALL (X-RE-6272)" SFP002, LRAM SFP 2 N WALL (X-RE-6273)" SFP003, LRAM SFP 1 E WALL (X-RE-6274)* SFP004, LRAM SFP 1 S WALL (X-RE-6275)Refueling Cavity:* RFCulO, LRAM W REFUEL CAV860 (u-RE-6251)* RFCu12, LRAM E REFUEL CAV 860 (u-RE-6253)Mode Applicability:AllDefinition(s):UNPLANNED-. A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient. The cause of the parameter changeor event may be known or unknown.REFUELING PATHWAY-. The reactor refueling cavity, spent fuel pool and fuel transfer canalcomprise the refueling pathway.CPNPP Basis:Indication of decreasing level includes Spent Fuel Pool Panel (FB 810) (ref. 2):" RFL CAVITY 1 LEVEL LO (upender area) (SFP-3.3)* RFL CAVITY 2 LEVEL LO (upender area) (SFP-3.7)" SFP 1 LEVEL LO (SFP-3.9)* SFP 2 LEVEL LO (SFP-3.10)Page 51 of 276 ATTACHMENT 1EAL Bases* RFL CAVITY 1 LEVEL LO (vessel area) (SFP-4.3)* RFL CAVITY 2 LEVEL LO (vessel area) (SFP-4.7)* SFP 1 TRANSFER CANAL LEVEL LO (SFP-4.9)* SFP 2 TRANSFER CANAL LEVEL LO (SFP-4.10)Allowing level to decrease could result in spent fuel being uncovered, reducing spent fueldecay heat removal and creating an extremely hazardous radiation environment. TechnicalSpecification Section 3.7.15 (ref. 4) requires at least 23 ft of water above the Spent Fuel Poolstorage racks (857' 31/2") (ref. 2). Technical Specification Section 3.9.7 (ref. 5) requires at least23 ft of water above the Reactor Vessel flange in the refueling cavity (856' 11" in refuelingcavity or 407" above core plate) (ref. 6). During refueling, this maintains sufficient water level inthe fuel transfer canal, refueling cavity, and SFP to retain iodine fission product activity in thewater in the event of a fuel handling accident. ABN-909, Spent Fuel Pool/Refueling CavityMalfunctions, provides appropriate guidance to restore and maintain normal water levels in thefuel transfer canal, refueling cavity, and SFP, and to determine if water levels have droppedbelow the Technical Specification LCOs (ref. 2). The fuel transfer canal is only of concern inassessing this EAL when irradiated fuel transfer is in progress, in which case the spent fuelpool gates are open and connected to the fuel transfer canal.NEI 99-01 Basis:This IC addresses a decrease in water level above irradiated fuel sufficient to cause elevatedradiation levels. This condition could be a precursor to a more serious event and is alsoindicative of a minor loss in the ability to control radiation levels within the plant. It is thereforea potential degradation in the level of safety of the plant.A water level decrease will be primarily determined by indications from available levelinstrumentation. Other sources of level indications may include reports from plant personnel(e.g., from a refueling crew) or video camera observations (if available). A significant drop inthe water level may also cause an increase in the radiation levels of adjacent areas that can bedetected by monitors in those locations.The effects of planned evolutions should be considered. For example, a refueling bridge arearadiation monitor reading may increase due to planned evolutions such as lifting of the reactorvessel head or movement of a fuel assembly. Note that this EAL is applicable only in caseswhere the elevated reading is due to an unplanned loss of water level.A drop in water level above irradiated fuel within the reactor vessel may be classified inaccordance Recognition Category C during the Cold Shutdown and Refueling modes.Escalation of the emergency classification level would be via IC AA2RA2.CPNPP Basis Reference(s):1. ABN-908 Fuel Handling Accident2. ABN-909 Spent Fuel Pool/Refueling Cavity Malfunctions3. 1-ALB-6B SFPCS TRBL4. Technical Specifications 3.7.15 Fuel Storage Area Water Level5. Technical Specifications 3.9.7 Refueling Cavity Water Level6. RFO-102A/B Refueling OperationsPage 52 of 276 ATTACHMENT IEAL Bases7. NEI 99-01 AU2Category: R -Abnormal Rad Levels / Rad EffluentSubcategory: 2 -Irradiated Fuel EventInitiating Condition: Significant lowering of water level above, or damage to, irradiated fuelEAL:RA2.1 Unusual EventUncovery of irradiated fuel in the REFUELING PATHWAYMode Applicability:AllDefinition(s):REFUELING PATHWAY-. The reactor refueling cavity, spent fuel pool and fuel transfer canalcomprise the refueling pathway.CPNPP Basis:None.NEI 99-01 Basis:This IC addresses events that have caused imminent or actual damage to an irradiated fuelassembly, or a significant lowering of water level within the spent fuel pool" see- .evek,.eFNetes). These events present radiological safety challenges to plant personnel and areprecursors to a release of radioactivity to the environment. As such, they represent an actualor potential substantial degradation of the level of safety of the plant.This IC applies to irradiated fuel that is ic~ensed for dry s~torage up to the point that theloaded storage cask is sealed. Once sealed, damage to a loaded cask causiRg loss of theCONFINEMIIENT BOUNDARY is classified in .c-, ... r ccord .A. ;ith IC E HUl.Esc-alation of the emergency would be bhasead on either Recognition Categor,' A Or CEAL-#-#This EAL escalates from AUJ2-RU2.1 in that the loss of level, in the affected portion of theREFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiatedfuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation(e.g., reports from personnel or camera images), as well as significant changes in water andradiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of availableindications, reports and observations.While an area radiation monitor could detect an increase in a dose rate due to a lowering ofwater level in some portion of the REFUELING PATHWAY, the reading may not be a reliableindication of whether or not the fuel is actually uncovered. To the degree possible, readingsshould be considered in combination with other available indications of inventory loss.A drop in water level above irradiated fuel within the reactor vessel may be classified inPage 53 of 276 ATTACHMENT 1EAL Basesaccordance Recognition Category C during the Cold Shutdown and Refueling modes.EAL4#2ThN" F.AI .. .nf rirnch'.. miftr~rn h,' ..tnirraiatd fel, Damaging eventS May include thc dropping, bumping or binding of aassembly, Or dro)pping a heavy load onto an assembly. A risc OR readingG on radiatoM9111101 -SHRUIG 08 GGR1S.-P-FP-Wyn W i0Iii tf vdd o'i"1bi-~tW1EAL.Hut t vn e.#3aue adigaý -trF= .f1.43 wt~ ~i tth""i r iti h n~ n n h ~~ in~n n~~t --.--. ....--. ..-.-. -prevent significafnt; do;se con Rsequences from direct gamma radiation to personnel performingemeratinn5G in the 'Anlnit'i of the speRnt fuel nee[ ThkS GE)RditiGR refkr~ts a ".innffictnt less of-. .....-J- .----.v..................... .... ...............ent Tuel pool water and thlus it is Also a PrecursorF to a less T me anniity toIn4 I I fI Idaacquately cool the irradiated Tuei assemoigis stored in the pool.Escalation of the emergency classification level would be via ICs AS-1-RS1 9F .AS2 (see.A,.DevýeFp-ee)CPNPP Basis Reference(s):1. ABN-909 Spent Fuel Pool/Refueling Cavity Malfunctions2. NEI 99-01 AA2Category: R -Abnormal Rad Levels / Rad EffluentSubcategory- 2 -Irradiated Fuel EventInitiating Condition: Significant lowering of water level above, or damage to, irradiated fuelEAL:RA2.2 AlertDamage to irradiated fuel resulting in a release of radioactivityANDHigh alarm on any of the following:" Any Table R-2 area radiation monitors* CAGu97, CNTMT AIR PIG GAS (u-RE-5503)* CAPu98, CNTMT AIR PIG PART (u-RE-5502)* CAIu99, CNTMT AIR PIG IODINE (u-RE-5566)" FBV088, FB VENT EXH (X-RE-5700)Table R-2 SFP & Refueling Cavity Area Radiation MonitorsSFP:* SFP001, LRAM SFP 2 E WALL (X-RE-6272)* SFP002, LRAM SFP 2 N WALL (X-RE-6273)" SFP003, LRAM SFP I E WALL (X-RE-6274)" SFP004, LRAM SFP 1 S WALL (X-RE-6275)Page 54 of 276 ATTACHMENT 1EAL BasesRefueling Cavity:* RFCulO, LRAM W REFUEL CAV860 (u-RE-6251)" RFCu12, LRAM E REFUEL CAV 860 (u-RE-6253)Mode Applicability:AllDefinition(s):NoneCPNPP Basis:The specified radiation monitors are those expected to see increase area radiation levels as aresult of damage to irradiated fuel (ref. 1, 2).The bases for the SFP ventilation radiation High alarm and the SFP and containment arearadiation readings are a spent fuel handling accident (ref. 1). In the Fuel Handling Building, afuel assembly could be dropped in the fuel transfer canal or in the SFP. Should a fuelassembly be dropped in the fuel transfer canal or in the SFP and release radioactivity above aprescribed level, the Fuel Handling Building ventilation monitors sound an alarm, alertingpersonnel to the problem.NEI 99-01 Basis:This IC addresses events that have caused imminent or actual damage to an irradiated fuelassembly, or a significant lowering of water level within the spent fuel pool(See -Deve..,..Netes). These events present radiological safety challenges to plant personnel and areprecursors to a release of radioactivity to the environment. As such, they represent an actualor potential substantial degradation of the level of safety of the plant.This IC applies to irradiated fuel that is lic-ensed fo-r dry storage up to the point that theloaded- storFage cask is sealed. Once sealed, damage to a loaded cask causing los6 of theCONFINEMI1ENT BOUNDARYis classified in ac.r.d. ,....with ICE HU.Escalation of the emergency would be based on either Recognition Category A-Ror C ICs.EAl lt--hs E,ATL ealates from in that the loss .. of level, inR the affected portion REFUELING P'HWY is f sufficin mgiuethaeruldinuncoYer' of irdaefuel. IndicatioNs of irradiated fuel uncover' may include direct or indirect visual l bselVati1(e.g., reports from personn~el or caer imaes), as well as significant changes in water andradiation levels, or other plant parametes.. o.m.putationa aids may also be used (e.g., a boiloff cu.e). Classification .f aR event using this EA.L should be based n the totality of availableindications, reports and obsevations.While an area radiation monitor could detect an increase in a dose rate due to alowering of water level in som e portion of the REF UELING PIADTHWAIAV the Feading may not be-A reliable indication of whetheror not t~he fu-'el is actually uncovered. To the degree possible,readings should bhe cosidered in combination With o-th-ravailable indicationS of invento' losAdrop in water level above irradiated fuel within the reactor vessel ma be classifiediaccordance Recognition Categor,' C duFrig the Co-ld Shutdo'A' and Refueling modes.Page 55 of 276 ATTACHMENT 1EAL BasesThis EAL addresses a release of radioactive material caused by mechanical damage toirradiated fuel. Damaging events may include the dropping, bumping or binding of anassembly, or dropping a heavy load onto an assembly. A rise in readings on radiationmonitors should be considered in conjunction with in-plant reports or observations of apotential fuel damaging event (e.g., a fuel handling accident).EAL i,,Spent fuel poel waterlevel at this value is within the lower end of the level range necessar,. to prevent S do6e conSequences from direct gamma radiation to perSonnel p.. rrming operations in thevicinity of the spent fuel pool. This condition reflects a signific-ant loss of spent fuel pool wtei nventor,' and thus it is also a prec-ursor to a loss of the ability to adequately cool the irdaefuel assembles stored in the pool.Escalation of the emergency classification level would be via ICes ASI--RS1 9FAS2-(seee-AS22D'evelepeF-Notes).CPNPP Basis Reference(s):1. ABN-908 Fuel Handling Accident2. ABN-909 Spent Fuel Pool/Refueling Cavity Malfunctions3. NEI 99-01 AA2I Page 56 of 276 ATTACHMENT IEAL BasesCategory:Subcategory:R -Abnormal Rad Levels / Rad Effluent2 -Irradiated Fuel EventInitiating Condition: Significant lowering of water level above, or damage to, irradiated fuelEAL:RA2.3 AlertLowering of spent fuel pool level to El. 844.3' (Level 2)Mode Applicability:AllDefinition(s):NoneCPNPP Basis:Post-Fukushima order EA-12-051 (ref. 1) required the installation of reliable SFP levelindication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of thefuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3) (ref. 1).Level 2 is the level that is adequate to provide substantial radiation shielding for a personstanding on the spent fuel pool operating deck. It represents the range of water level whereany necessary operations in the vicinity of the spent fuel pool can be completed withoutsignificant dose consequences from direct gamma radiation from the stored spent fuel.Comanche Peak designated as Level 2 the water level 10 feet (+/- 1.0 foot) above the top of thefuel racks (El 844' -2.75" rounded to 844.3' indicated) (ref. 2).The enhanced SFP level instruments (X-LI-4876, 4878, 4877, 4879) do not have indicationavailable in the control room and must be read remotely outside of the control room.NEI 99-01 Basis:This IC addresses events that have caused imminent or actual damage to an irradiated fuelassembly, or a significant lowering of water level within the spent fuel pool- (see Nte4. These events present radiological safety challenges to plant personnel and areprecursors to a release of radioactivity to the environment. As such, they represent an actualor potential substantial degradation of the level of safety of the plant.This IC applies to irradiated fuel that is licensed for storage up to the point that theloaded storage cask is sealed. Oncc sealed, damage to a loaded cask causing loss of theCONFINEMENT BOUINDARY is classified on with IC E HUE--Escalation of the emergency would be based on either Recognition Category A-Ror C1('q~ PA L44T1,kQ P~A L imi==t= m Ul 2 on- th-n+ +j-P Aflq Af I'PUP' on th'-' 'ff4PL~tp -~.nndon nf ih,-nR=FUELIN 1F=1 HAY, is of sufficient magnitude to hav.e resulted in uncoVery of irrFadiatedfuel. Indications of iRradiated fuel uncovery may include direct or indirect visual observation(e.g., reports from personnel or caer iags), as well as significant changes in water andr-adiatien levels, or other plant par-amieteirs. Coamputational aids may also be used (e.g., a boilindicationS, repo-ts and ,bseR,'atkiN4.Page 57 of 276 ATTACHMENT 1EAL BasesWhile an area radiatienR uld detect aninreae a dose rate due to al-w-oing ef water level in some portion of the REFUELING PATHlw Y, the reading may not bea reliable indicatio enf whether r icnt the fuel is acually urcgvered. To the degree possible,readings should be onsidered ini cofbiatioe with other available indieation9res of i gniVefto loA drfp in water level above inradiated fuel within the reacutor el ay be clasified inaccordance Recognition Categor,' C during the Cold Shu~tdoWn and Refueling moedes.This EAL addrpesrspes a- rpelpease o-f radioactive maeral cased by mnechan~ical damnage toirradiated fuel. Damaging evenss may include the dropping, bLmping or binding oassembly, or dropping a heavy load onRto an assembly. A rise in readings on radiatomonitors should be considered in conjunction With in plant reports or o~bservations of apotential fuel damaging event (e.g., a fuel handling accident).AL#Spent fuel pool water level at this value is within the lower end of the level rangenecessary to prevent significant dose consequences from direct gamma radiation to personnelperforming operations in the vicinity of the spent fuel pool. This condition reflects a significantloss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability toadequately cool the irradiated fuel assembles stored in the pool.Escalation of the emergency classification level would be via ICs AS4-RS1-, A$2 A9/- p A,,., -I -Page 58 of 276 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):1. NRC EA-12-51 Issuance of Order to Modify Licenses with Regard to Reliable Spent FuelPool Instrumentation2. TXX-13103 Overall Integrated Plan in Response to March 12,2012 Commission OrderModifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (OrderNumber EA-12-051), Response to Request for Additional Information3 NEI 99-01 AA2Page 59 of 276 ATTACHMENT 1EAL BasesCategory: R -Abnormal Rad Levels / Rad EffluentSubcategory: 2- Irradiated Fuel EventInitiating Condition: Spent fuel pool level at the top of the fuel racksEAL:RS2.1 Site Area EmergencyLowering of spent fuel pool level to El. 835.3' (Level 3)Mode Applicability:AllDefinition(s):NoneCPNPP Basis:Post-Fukushima order EA-12-051 (ref. 1) required the installation of reliable SFP levelindication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of thefuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3) (ref. 1).Level 3 is the level where fuel remains covered and actions to implement make-up wateraddition should no longer be deferred. Level 3 corresponds nominally (i.e., +/- 1 foot) to thehighest point of any fuel rack seated in the spent fuel pool. Level 3 is defined in this manner toprovide the maximum range of information to operators, decision makers and emergencyresponse personnel. Comanche Peak designated as Level 3 the water level greater than 1 footabove the top of the fuel storage racks plus the accuracy of the SFP level instrument channel(El. 835' -2.75" rounded to 835.3' indicated). Designation of this level as Level 3 isconservative; its selection assures that the fuel will remain covered, and at that point therewould be no functional or operational reason to defer action to implement the addition of make-up water to the pool (ref. 2).The enhanced SFP level instruments (X-LI-4876, 4878, 4877, 4879) do not have indicationavailable in the control room and must be read remotely outside of the control room.NEI 99-01 Basis:This 1C-EAL addresses a significant loss of spent fuel pool inventory control and makeupcapability leading to IMMINENT fuel damage. This condition entails major failures of plantfunctions needed for protection of the public and thus warrant a Site Area Emergencydeclaration.It is recognized that this IC would likely not be met until well after another Site Area EmergencyIC was met; however, it is included to provide classification diversity.Escalation of the emergency classification level would be via IC AG1 or AG2RG2.I Page 60 of 276 1 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):1. NRC EA-12-51 Issuance of Order to Modify Licenses with Regard to Reliable Spent FuelPool Instrumentation2. TXX-13103 Overall Integrated Plan in Response to March 12,2012 Commission OrderModifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (OrderNumber EA-12-051), Response to Request for Additional Information3. NEI 99-01 AS2I Page 61 of 276 ATTACHMENT 1EAL BasesCategory: R -Abnormal Rad Levels / Rad EffluentSubcategory: 2 -Irradiated Fuel EventInitiating Condition: Spent fuel pool level cannot be restored to at least the top of the fuelracks for 60 minutes or longerEAL:RG2.1 General EmergencySpent fuel pool level cannot be restored to at least El. 835.3' (Level 3) for greater than orequal to 60 min. (Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Mode Applicability:AllDefinition(s):NoneCPNPP Basis:Post-Fukushima order EA-12-051 (ref. 1) required the installation of reliable SFP levelindication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of thefuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3) (ref. 1).Level 3 is the level where fuel remains covered and actions to implement make-up wateraddition should no longer be deferred. Level 3 corresponds nominally (i.e., +/- 1 foot) to thehighest point of any fuel rack seated in the spent fuel pool. Level 3 is defined in this manner toprovide the maximum range of information to operators, decision makers and emergencyresponse personnel. Comanche Peak designated as Level 3 the water level greater than 1 footabove the top of the fuel storage racks plus the accuracy of the SFP level instrument channel(El. 835' -2.75" rounded to 835.3' indicated). Designation of this level as Level 3 isconservative; its selection assures that the fuel will remain covered, and at that point therewould be no functional or operational reason to defer action to implement the addition of make-up water to the pool (ref. 2).The enhanced SFP level instruments (X-LI-4876, 4878, 4877, 4879) do not have indicationavailable in the control room and must be read remotely outside of the control room.NEI 99-01 Basis:This I,-EAL addresses a significant loss of spent fuel pool inventory control and makeupcapability leading to a prolonged uncovery of spent fuel. This condition will lead to fueldamage and a radiological release to the environment.It is recognized that this IC would likely not be met until well after another General EmergencyIC was met; however, it is included to provide classification diversity.Page 62 of 276 ATTACHMENT IEAL BasesCPNPP Basis Reference(s):1. NRC EA-12-51 Issuance of Order to Modify Licenses with Regard to Reliable Spent FuelPool Instrumentation2. TXX-13103 Overall Integrated Plan in Response to March 12,2012 Commission OrderModifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (OrderNumber EA-12-051), Response to Request for Additional Information3. NEI 99-01 AG2Page 63 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:EAL:R -Abnormal Rad Levels / Rad Effluent3 -Area Radiation LevelsRadiation levels that IMPEDE access to equipment necessary fornormal plant operations, cooldown or shutdownRA3.1 AlertDose rates greater than 15 mR/hr in EITHER of the following areas:Control RoomCRM048 (X-RE-6281) or CRM049 (X-RE-6282)ORCentral Alarm Station (by survey)Mode Applicability:AllDefinition(s):NoneCPNPP Basis:X-RE-6281 and X-RE-6282 are the installed Control Room area radiation monitors and may beused to assess this EAL threshold (range of 1 E-4 to 1 E+5 mR/hr). However, no permanentlyinstalled area radiation monitoring is installed in the CAS and therefore this threshold must beassessed via local radiation survey (ref. 1).NEI 99-01 Basis:This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to precludeor impede personnel from performing actions necessary to maintain normal plant operation, orto perform a normal plant cooldown and shutdown. As such, it represents an actual orpotential substantial degradation of the level of safety of the plant. The ...Emergqency Coordinator should consider the cause of the increased radiation levels anddetermine if another IC may be applicable.For EAr. #2, an Alert delara-onj, is warranted ifoperatinG mode in effect at the time of the elevated radiation levels. The emerenc'-S;~ ~atcnnir'i~ntinn ~hrh~r ~nt" kint ini' nr'~rv t th~ tmr~nf hr~ ----....- .J-increased radiation levels. Access should be considered as impeded if extraordinary.- aa , .a ara.,aa co .,+a fn H;+ I+a .+ -Fr, ~r arrr .nna Ltnia l, a +k^ ad raa rnIa -a Ia ----. .IVinstalling temporary shielding, reqIu"irig Use of non rou1tine protective equipment, requesting an4extension in dose limits beyond nremal administrative lmt)An emergency declaration is not wratdif any of the folloWing conditions apply-.aThe plant is in an operating mode differet than the mode specified for the affectedroom/area (i.e., entry is not required during the operating mode in effect at the time otPage 64 of 276 ATTACHMENT 1EAL Basesthe elevated radiation levels). For example, the plant is in Mode 1 whcn the radiationincrease occurFs, and the proGedures used for normnal opcration, cooldoWn andshutdW-n dGo nt require ent' ; into the affected room until Mode 4.alThe incfeased radiation levels are a result of a planed acgtivity that incRlucomnpensator,' measures which addrcss the temporar,' inaGeeSSibility of a room or area(e.g., radiography, spcnt filter or re-sin transfer, etc.).*The action for which room/area entr,'is required is of an administrative or recorFdkeeping nature (e.g., nr9mal rounds or routine inspections).-*The access control measures are of a conservative or precauitionar,' natur~e, and would1not actually prevent or impede a required action-.Escalation of the emergency classification level would be via Recognition Category AýR, C or FICs.CPNPP Basis Reference(s):1. DBD-EE-023 Radiation Monitoring System2. NEI 99-01 AA3Page 65 of 276 ATTACHMENT IEAL BasesCategory: R -Abnormal Rad Levels / Rad EffluentSubcategory: 3- Area Radiation LevelsInitiating Condition: Radiation levels that IMPEDE access to equipment necessary fornormal plant operations, cooldown or shutdownEAL:RA3.2 AlertAn UNPLANNED event results in radiation levels that prohibit or IMPEDE access to anyTable R-3 rooms or areas (Note 5)Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, thenno emergency classification is warranted.Table R-3 Safe Operation & Shutdown Rooms/AreasRoom/Area Mode ApplicabilityCharging Pump Rooms 1, 2, 3, 4, 5, 6CVCS Valve Rooms 1, 2, 3, 4, 5, 61 E Switchgear Rooms AllRHR Pump Rooms 4, 5, 6Mode Applicability:AllDefinition(s):IMPEDE(D) -Personnel access to a room or area is hindered to an extent that extraordinarymeasures are necessary to facilitate entry of personnel into the affected room/area(e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).UNPLANNED-. A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient. The cause of the parameter changeor event may be known or unknown.CPNPP Basis:For this EAL, area or room access is considered impeded if radiation levels require locked highradiation controls to to be imposed.If the equipment in the listed room or area was already inoperable, or out-of-service, before theevent occurred, then no emergency should be declared since the event will have no adverseimpact beyond that already allowed by Technical Specifications at the time of the event.The list of plant rooms or areas with entry-related mode applicability identified specify thoserooms or areas that contain equipment which require a manual/local action as specified inoperating procedures used for normal plant operation, cooldown and shutdown. Rooms orareas in which actions of a contingent or emergency nature would be performed (e.g., anaction to address an off-normal or emergency condition such as emergency repairs, correctivePage 66 of 276 ATTACHMENT 1EAL Basesmeasures or emergency operations) are not included. In addition, the list specifies the plantmode(s) during which entry would be required for each room or area (ref. 1).NEI 99-01 Basis:This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to precludeor impede personnel from performing actions necessary to maintain normal plant operation, orto perform a normal plant cooldown and shutdown. As such, it represents an actual orpotential substantial degradation of the level of safety of the plant. The ergiRyDirecteFEmergency Coordinator should consider the cause of the increased radiation levelsand determine if another IC may be applicable.I For EAL- #2RA3.2, an Alert declaration is warranted if entry into the affected room/area is, ormay be, procedurally required during the plant operating mode in effect at the time of theelevated radiation levels. The emergency classification is not contingent upon whether entry isactually necessary at the time of the increased radiation levels. Access should be consideredas impeded if extraordinary measures are necessary to facilitate entry of personnel into theaffected room/area (e.g., installing temporary shielding, requiring use of non-routine protectiveequipment, requesting an extension in dose limits beyond normal administrative limits).An emergency declaration is not warranted if any of the following conditions apply:" The plant is in an operating mode different than the mode specified for the affectedroom/area (i.e., entry is not required during the operating mode in effect at the time ofthe elevated radiation levels). For example, the plant is in Mode 1 when the radiationincrease occurs, and the procedures used for normal operation, cooldown andshutdown do not require entry into the affected room until Mode 4." The increased radiation levels are a result of a planned activity that includescompensatory measures which address the temporary inaccessibility of a room or area(e.g., radiography, spent filter or resin transfer, etc.)." The action for which room/area entry is required is of an administrative or recordkeeping nature (e.g., normal rounds or routine inspections)." The access control measures are of a conservative or precautionary nature, and wouldnot actually prevent or impede a required action.Escalation of the emergency classification level would be via Recognition Category AR, C or FICs.CPNPP Basis Reference(s):1. Attachment 3 Safe Operation & Shutdown Areas Tables R-3 & H-2 Bases2. NEI 99-01 AA3Page 67 of 276 ATTACHMENT IEAL BasesCategory E -Independent Spent Fuel Storage Installation (ISFSI)EAL Group: Any (EALs in this category are applicable to anyplant condition, hot or cold.)An independent spent fuel storage installation (ISFSI) is a complex that is designed andconstructed for the interim storage of spent nuclear fuel and other radioactive materialsassociated with spent fuel storage. A significant amount of the radioactive material containedwithin a canister must escape its packaging and enter the biosphere for there to be asignificant environmental effect resulting from an accident involving the dry storage of spentnuclear fuel.An Unusual Event is declared on the basis of the occurrence of an event of sufficientmagnitude that a loaded cask confinement boundary is damaged or violated.IPage 68 of 276 ATTACHMENT 1EAL BasesCategory: ISFSISubcategory: Confinement BoundaryInitiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARYEAL:EUl.1 Unusual EventDamage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contactradiation reading greater than EITHER:* 60 mrem/hr (-r + rq) on the top of the overpack* 600 mrem/hr (T" + rj on the side of the overpack (excluding inlet and outlet ducts)Mode Applicability:AllDefinition(s):CONFINEMENT BOUNDARY-. The barrier(s) between spent fuel and the environment oncethe spent fuel is processed for dry storage. As applied to the CPNPP ISFSI, theCONFINEMENT BOUNDARY is defined to be the Multi-Purpose Canister (MPC).CPNPP Basis:The ISFSI includes the dry-cask storage system, the cask transfer facility, onsite transporter,and the storage pads. The dry-cask storage system is the HI-STORM 100 System. This is acanister-based storage system that stores spent nuclear fuel in a vertical orientation. Itconsists of three discrete components: the MPC, the HI-TRAC 125 Transfer Cask, and the HI-STORM 100 System Overpack. The MPC provides the confinement boundary for the storedfuel. The HI-TRAC 125 Transfer Cask provides radiation shielding and structural protection ofthe MPC during transfer operations, while the storage overpack provides radiation shieldingand structural protection of the MPC during storage (ref. 1).The value shown represents 2 times the limits specified in the ISFSI Certificate of ComplianceTechnical Specification section 5.7.4 for radiation external to a loaded MPC overpack (ref. 1).NEI 99-01 Basis:This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of astorage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storagebeginning at the point that the loaded storage cask is sealed. The issues of concern are thecreation of a potential or actual release path to the environment, degradation of one or morefuel assemblies due to environmental factors, and configuration changes which could causechallenges in removing the cask or fuel from storage.The existence of "damage" is determined by radiological survey. The technical specificationI multiple of "2 times", which is also used in Recognition Category A-RIC RAU1, is used here toI Page 69 of 276 ATTACHMENT IEAL Basesdistinguish between non-emergency and emergency conditions. The emphasis for thisclassification is the degradation in the level of safety of the spent fuel cask and not themagnitude of the associated dose or dose rate. It is recognized that in the case of extremedamage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may bedetermined based on measurement of a dose rate at some distance from the cask.Security-related events for ISFSIs are covered under ICs HU1 and HAl.CPNPP Basis Reference(s):1. HI-2104635, HI-STORM Certificate of Compliance Appendix A Technical SpecificationSection 5.7.42. RPI-792 HI-STORM Overpack Surface Dose Rates3. NEI 99-01 E-HU1Page 70 of 276 ATTACHMENT 1EAL BasesCategory C -Cold Shutdown / Refueling System MalfunctionEAL Group: Cold Conditions (RCS temperature < 2000F); EALsin this category are applicable only in one or morecold 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) duringthese periods, the consequences of any given initiating event can vary greatly. For example, aloss of decay heat removal capability that occurs at the end of an extended outage has lesssignificance than a similar loss occurring during the first week after shutdown. Compoundingthese events is the likelihood that instrumentation necessary for assessment may also beinoperable. The cold shutdown and refueling system malfunction EALs are based onperformance capability to the extent possible with consideration given to RCS integrity,containment closure, and fuel clad integrity for the applicable operating modes (5 -ColdShutdown, 6 -Refueling, D -Defueled).The events of this category pertain to the following subcategories:1. RCS LevelRCS water level is directly related to the status of adequate core cooling and, therefore,fuel clad integrity.2. Loss of Emerqency AC PowerLoss of emergency plant electrical power can compromise plant safety system operabilityincluding decay heat removal and emergency core cooling systems which may benecessary to ensure fission product barrier integrity. This category includes loss of onsiteand offsite power sources for 6.9 KV AC emergency buses.3. RCS TemperatureUncontrolled or inadvertent temperature or pressure increases are indicative of a potentialloss of safety functions.4. Loss of Vital DC PowerLoss of emergency plant electrical power can compromise plant safety system operabilityincluding decay heat removal and emergency core cooling systems which may benecessary to ensure fission product barrier integrity. This category includes loss of power toor degraded voltage on the 125V DC vital buses.5. Loss of CommunicationsCertain events that degrade plant operator ability to effectively communicate with essentialpersonnel within or external to the plant warrant emergency classification.6. Hazardous Event Affecting Safety SystemsCertain hazardous natural and technological events may result in visible damage to ordegraded performance of safety systems warranting classification.Page 71 of 276 1 ATTACHMENT 1EAL BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 1 -RCS LevelInitiating Condition: UNPLANNED loss of RCS inventory for 15 minutes or longerEAL:CUI.1 Unusual EventUNPLANNED loss of reactor coolant results in RCS water level less than a required lowerlimit for greater than or equal to 15 min. (Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Mode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):UNPLANNED-. A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient. The cause of the parameter changeor event may be known or unknown.CPNPP Basis:With the plant in Cold Shutdown, RCS water level is normally maintained above thepressurizer low level setpoint of 17% (ref. 1). However, if RCS level is being controlled belowthe pressurizer low level setpoint, or if level is being maintained in a designated band in thereactor vessel it is the inability to maintain level above the low end of the designated controlband due to a loss of inventory resulting from a leak in the RCS that is the concern.With the plant in Refueling mode, RCS water level is normally maintained at or above thereactor vessel flange (Technical Specification LCO 3.9.7 requires at least 23 ft. of water abovethe top of the reactor vessel flange in the refueling cavity during refueling operations) (ref. 2).The Reactor Vessel flange level is 834' 1/2" elevation or 132.5 in. above the upper core plate(top) (ref. 3).NEI 99-01 Basis:This IC addresses the inability to restore and maintain water level to a required minimum level(or the lower limit of a level band), or a loss of the ability to monitor .. .....-v-..-,IRCS {RPWR],r RPV- r[{BVR)-level concurrent with indications of coolant leakage. Either of these conditionsis considered to be a potential degradation of the level of safety of the plant.Refueling evolutions that decrease RCS-water inventory are carefully planned and controlled.An UNPLANNED event that results in water level decreasing below a procedurally requiredlimit warrants the declaration of an Unusual Event due to the reduced water inventory that isavailable to keep the core covered.This EAL-#4 recognizes that the minimum required [P.4'.] or RPV [BW,4..)level can change several times during the course of a refueling outage as different plantconfigurations and system lineups are implemented. This EAL is met if the minimum level,-Page 72 of 276 ATTACHMENT IEAL Basesspecified for the current plant conditions, cannot be maintained for 15 minutes or longer. Theminimum level is typically specified in the applicable operating procedure but may be specifiedin another controlling document.The 15-minute threshold duration allows sufficient time for prompt operator actions to restoreand maintain the expected water level. This criterion excludes transient conditions causing abrief lowering of water level.EAL o #2 addresses a condition ,,,here all means to determ.ine (reactoF vesse.R.S[PI/lRI or RPV [B3VARj) level have been lost. In tthis, cond-iqtion, operators may determinle that aninventory loss is occurring by ebseiR'ig changes1 inSu1mp anmd/ortank levels. Sump and/eFtank level chng Smut bhe evaluated against other potential sources of water flow to ensur~eth e" a re n di r-t:-; Zt ofp l11e.a ka ge fro m th e (re acGtor '.GeS P-VRCGS2 [P- tIP] orF RPV [BWR]).Continued loss of RCS inventory may result in escalation to the Alert emergency classificationlevel via either IC CAl or CA3.CPNPP Basis Reference(s):1. ALM-0052A/B Alarm Procedure u-ALB-5B3 (513-3.6)2. Technical Specification Section 3.9.7 Refueling Cavity Water Level3. IPO-Ol OAIB Reactor Coolant System Reduced Inventory Operations4. NEI 99-01 CUII Page 73 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:C -Cold Shutdown / Refueling System Malfunction1 -RCS LevelInitiating Condition: UNPLANNED loss of RCS inventory for 15 minutes or longerEAL:CUI.2 Unusual EventRCS water level cannot be monitoredAND EITHER* UNPLANNED increase in any Table C-1 sump/tank level due to loss of RCSinventory* Visual observation of UNISOLABLE RCS leakageTable C-I Sumps / Tanks* Containment Sump 1* Containment Sump 2* Reactor Cavity Sump* CCW Surge Tank A* CCW Surge Tank B* PRT.RCDTMode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally.UNPLANNED-. A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient. The cause of the parameter changeor event may be known or unknown.CPNPP Basis:In Cold Shutdown mode, the RCS will normally be intact and standard RCS level monitoringmeans are available.In this EAL, all water level indication is unavailable and the RCS inventory loss must bedetected by indirect leakage indications (Table C-1). Level increases must be evaluatedagainst other potential sources of leakage such as cooling water sources inside thecontainment to ensure they are indicative of RCS leakage. If the make-up rate to the RCSunexplainably rises above the pre-established rate, a loss of RCS inventory may be occurringeven if the source of the leakage cannot be immediately identified. Visual observation ofPage 74 of 276 ATTACHMENT 1EAL Basesleakage from systems connected to the RCS that cannot be isolated could also be indicative ofa loss of RCS inventory (ref. 1, 2, 3, 4).NEI 99-01 Basis:This IC addresses the inability to restore and maintain water level to a required minimum level(or the lower limit of a level band), or a loss of the ability to monitor (FeaGter-veseRCS [PWR]or RP, rl3WRYlevel concurrent with indications of coolant leakage. Either of these conditionsis considered to be a potential degradation of the level of safety of the plant.Refueling evolutions that decrease RCS water inventory are carefully planned and controlled.An UNPLANNED event that results in water level decreasing below a procedurally requiredlimit warrants the declaration of an Unusual Event due to the reduced water inventory that isavailable to keep the core covered.E-=ALI #1 recognizes that the mnumrquired (reactor vessel/RCS [PWRq or RPV[914qR) level can change several times during the course of a refueling outage as difflerent.plant configurations and system lineups are implemented. This EAL is met if the minimumlevel, specified forF the- current plant conditions, cannot be maint-ained- for 15 minutes Or longer.-The minimum lev~,el is typically specified in the apibloperating procedure but ma" bespecified in another controlling document.The 15 minute threshold duration allows sufficient time8 for prompt operator actions to restoreand maintain the-expected water level. Thi-s criterion excludes transient conditinsausn abrief lowering of water level.This EAL-#-2 addresses a condition where all means to determine (reactor vessel!RCS [PWR9F RPV {BWRI.) level have been lost. In this condition, operators may determine that aninventory loss is occurring by observing changes in sump and/or tank levels (Table C-1).Sump and/or tank level changes must be evaluated against other potential sources of waterflow to ensure they are indicative of leakage from the (.eat.. .ess RCS R F R[BWR]).Continued loss of RCS inventory may result in escalation to the Alert emergency classificationlevel via either IC CAl or CA3.CPNPP Basis Reference(s):1. 1PO-01OA/B Reactor Coolant System Reduced Inventory Operations2. SOP-I1NB/1 Reactor Coolant System3. ABN-103 Excessive Reactor Coolant Leakage4. ABN-108 Shutdown Loss of Coolant5. NEI 99-01 CUIPage 75 of 276 ATTACHMENT 1EAL BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 1 -RCS LevelInitiating Condition: Loss of RCS inventoryEAL:CA1.1 AlertLoss of RCS inventory as indicated by RCS level less than 48 in. above upper core plate(top)Mode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):NoneCPNPP Basis:When Reactor Vessel water level decreases to 48 in. above the upper core plate (top) (EL 827'0"), RHR pump cavitation may occur. RCS level can be monitored by one or more of thefollowing (ref. 1, 2, 3, 4, 5, 6):* RCS Level Wide Range u-LI-3615B" RCS Level Narrow Range u-LI-3615A* RCS Extended Wide Range u-LI-3615C* Mansell Level Monitor System u-LT-3619AIB/C-1, -2" Plant Computer* RVLIS* Ultrasonic Level monitoring (optional)NEI 99-01 Basis:This IC addresses conditions that are precursors to a loss of the ability to adequately coolirradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This conditionrepresents a potential substantial reduction in the level of plant safety.For this EAL-#-I-, a lowering of RCS water level below 48 in. above the upper core plate (top)(site specific level) ft. indicates that operator actions have not been successful in restoring andmaintaining RCS (reactor vesse!IRCS [PWrIRI or RPV,4' water level. The heat-up rate ofthe coolant will increase as the available water inventory is reduced. A continuing decrease inwater level will lead to core uncovery.Although related, this EAL-#4 is concerned with the loss of RCS inventory and not the potentialconcurrent effects on systems needed for decay heat removal (e.g., loss of a Residul- DecayHeat Removal suction point). An increase in RGS-RCS temperature caused by a loss of decayheat removal capability is evaluated under IC CA3.Page 76 of 276 ATTACHMENT 1EAL BasesFor EAL #2, the inability to mon.itor (reactor V.SSel.R. S [PV.R] or RPV [B'A'R]) levelmay be caused by instrumentation and/orF power failures, Or water level dropping below therange of available instFrumentation. If water level c3-annot bhe monitored, operators m~aydetermne that an invento' los is occurri. by obsering c.hanges in sump and/or tank levels.Sump and/or tank level changes mus t be evaluated against other potenti-al sou1-rces of waterflow to ensure they are iniaieof leakage from the (reactor vcssel/RCS [PWVRI or RPVThe 15 mninute duration for the loss of level indication was chosen because it is half of the EALduration specified in IC CSIIf RCS the (reactor vcssel/RCS [PWVR] or RPV~ [BVVR]) inventory water level continues tolower, then escalation to Site Area Emergency would be via IC OSi.CPNPP Basis Reference(s):1 .IPO-Ol OA/B Reactor Coolant System Reduced Inventory Operations2. INC-6269 Calibration of the Mansell RCS Measurement System3. SOP-l0lA/B Reactor Coolant System4. ABN-1 03 Excessive Reactor Coolant Leakage5. ABN-104 Residual Heat Removal System Malfunction6. ABN-108 Shutdown Loss of Coolant7. NEI 99-01 CAlPage 77 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:EAL:C -Cold Shutdown / Refueling System Malfunction1 -RCS LevelLoss of RCS inventoryCA1.2 AlertRCS water level cannot be monitored for greater than or equal to 15 min. (Note 1)AND EITHER" UNPLANNED increase in any Table C-1 sump/tank level due to loss of RCSinventory" Visual observation of UNISOLABLE RCS leakageNote 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has beenexceeded, or will likely be exceeded.Table C-1 Sumps / Tanks* Containment Sump 1* Containment Sump 2* Reactor Cavity Sump* CCW Surge TankA* CCW Surge Tank B* PRT* RCDTMode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally.UNPLANNED-. A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient. The cause of the parameter changeor event may be known or unknown.CPNPP Basis:In Cold Shutdown mode, the RCS will normally be intact and standard RCS level monitoringmeans are available.In the Refuel mode, the RCS is not intact and RPV level may be monitored by different means,including the ability to monitor level visually.In this EAL, all RCS water level indication would be unavailable for greater than 15 minutes,and the RCS inventory loss must be detected by indirect leakage indications (Table C-1).Page 78 of 276 ATTACHMENT 1EAL BasesSurveillance procedures provide instructions for calculating primary system leak rate bymanual or computer-based water inventory balances. Level increases must be evaluatedagainst other potential sources of leakage such as cooling water sources inside thecontainment to ensure they are indicative of RCS leakage. If the make-up rate to the RCSunexplainably rises above the pre-established rate, a loss of RCS inventory may be occurringeven if the source of the leakage cannot be immediately identified. Visual observation ofleakage from systems connected to the RCS that cannot be isolated could also be indicative ofa loss of RCS inventory (ref. 1, 2, 3).NEI 99-01 Basis:This IC addresses conditions that are precursors to a loss of the ability to adequately coolirradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This conditionrepresents a potential substantial reduction in the level of plant safety.PFo EAL #1, a lowering of water level below (site specific level) indicates, that operatorFactions have not been successful in restGorig and maintainfin~g (reactor vcsscl!RCS [PL4'R oeRPV wate level. The heat up rate Of the coolant will increase as the available winventory is reduced. A continuing decrease in water leve! will lead to core uncover;.Alth.ugh related, #1 is concerned with the loss of RCS inventor.' and not thepotential concurrent effects On systems needed for decay heat removal (e.g., loss of aResidual Heat Removal suction point). An inrGease in ROS temperature caused by a loss ofdecay heat removal capability is evaluated under IC CA33.For this EAL-#2-, the inability to monitor RCS (reactor ..essel.RS [PWý eor RPV [BWR]) levelmay be caused by instrumentation and/or power failures, or water level dropping below therange of available instrumentation. If water level cannot be monitored, operators maydetermine that an inventory loss is occurring by observing changes in sump and/or tank levels.Sump and/or tank level changes must be evaluated against other potential sources of waterflow to ensure they are indicative of leakage from the (eaeter-vesSeRCS [PWR] or RPVThe 15-minute duration for the loss of level indication was chosen because it is half of the EALduration specified in IC CS1_If the (reaGt O.ves....RCS [PW.r or RPV [B\R]) inventory level continues to lower, thenescalation to Site Area Emergency would be via IC CS1.CPNPP Basis Reference(s):1. ABN-1 03 Excessive Reactor Coolant Leakage2. ABN-108 Shutdown Loss of Coolant3. FSAR 5.2.5.24. NEI 99-01 CA1Page 79 of 276 ATTACHMENT 1EAL BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 1 -RCS LevelInitiating Condition: Loss of RCS inventory affecting core decay heat removal capabilityEAL:CS1.1 Site Area EmergencyWith CONTAINMENT CLOSURE not established, RCS level less than 27.3 in. aboveupper core plate (top)Mode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):CONTAINMENT CLOSURE -The procedurally defined actions taken to secure containmentand its associated structures, systems, and components as a functional barrier to fissionproduct release under shutdown conditions. Containment closure means that all potentialescape paths are closed or capable of being closed:A. All penetrations providing direct access from Containment atmosphere to outsideatmosphere are closed except:" Penetrations with automatic valves capable of being closed by an operable CVI" Penetrations under administrative controls (e.g., Control Room notified and designatedperson to close if required by fuel handling accident)B. Equipment hatch is closed and held in place by 4 bolts, or is capable of being closed andheld in place by 4 boltsC. One emergency airlock door is closedD. One personnel airlock door is capable of being closedCPNPP Basis:When Reactor Vessel water level decreases to 27.25 in. (rounded to 27.3 in. for instrumentreadability), 825'-3 11/4" elevation (ref. 1), water level is six inches below the elevation of thebottom of the RCS hot leg penetration. When Reactor Vessel water level drops significantlybelow the elevation of the bottom of the RCS hot leg penetration, all sources of RCS injectionhave failed or are incapable of making up for the inventory loss. RCS elevations are illustratedin Figure C-3. RCS level can be monitored by one or more of the following (ref. 1, 2, 3):* RCS Level Wide Range u-LI-3615B* RCS Level Narrow Range u-LI-3615A" RCS Extended Wide Range u-LI-3615C* Mansell Level Monitor System u-LT-3619A/B/C-1, -2" Plant Computer" RVLIS* Ultrasonic Level monitoring (optional)Page 80 of 276 ATTACHMENT 1EAL BasesIn Refueling mode, Reactor Vessel water level indication from RVLIS is likely unavailable butalternate means of level indication are normally installed (including visual observation) toassure that the ability to monitor water level will not be interrupted.The status of Containment closure is tracked if plant conditions change that could raise the riskof a fission product release as a result of a loss of decay heat removal (ref. 4, 5).NEI 99-01 Basis:This IC addresses a significant and prolonged loss of (reactor vessel/RGS-RCS_[PWRe] r RPVrf^W])- inventory control and makeup capability leading to IMMINENT fuel damage. The lostinventory may be due to a RCS component failure, a loss of configuration control or prolongedboiling of reactor coolant. These conditions entail major failures of plant functions needed forprotection of the public and thus warrant a Site Area Emergency declaration.Following an extended loss of core decay heat removal and inventory makeup, decay heat willcause reactor coolant boiling and a further reduction in reactor vessel level. If RGS-eaeteivessel RCS level cannot be restored, fuel damage is probable.Outage/shutdown contingency plans typically provide for re-establishing or verifyingCONTAINMENT CLOSURE following a loss of heat removal or RCS inventory controlfunctions. The difference in the specified RCS/reactor vessel levels of EALs 4-1CS1.1 and2-bCS2.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lowerprobability of a fission product release to the environment.in EAL 3.a, the 30 minute criterion is tied to a readily reco~gnizable event start time (i.e., thetetal loss of ability to monitor level), and allows sGfficient time to monitoF, a2s6se -and c iorrelreactor and plant co~nditions to determnine if core Uncovery has actually occurred (i.e., toacco)unt for vaiu accident progression and instFrumentation uncertainties). it also allowssufficient time for pcfo~rmancc of actions to terminate leakage, recover inventor;control/makeup equipment and~or restore leVel monitoring-.The inability to monitor RCS (reactor vesscl/RGS [PWVR] or RPV [BWVR]) level may be causedby instrumentation and/or power failures, or water level dropping below the range of availab~ins~trumentation. if water level cannot be monitored, operators ma" determine thatainventor; loss is Gccurring by observing chane in sup and/or tank levels. Sump and/oitank level changes must be evaluated against other potential sources of water flow to ensurethey are indicativ~e of leakage from theRGS (reactor vesseliRCS [PVVARI or RP'. [B'NRI).These-ThisEALs addresses concerns raised by Generic Letter 88-17, Loss of Decay HeatRemoval; SECY 91 -283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449,Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the UnitedStates; and NUMARO 91-06, Guidelines for Industry Actions to Assess ShutdownManagement.Escalation of the emergency classification level would be via IC CGI or AG4RG1CPNPP Basis Reference(s):1. IPO-01 OA/B Reactor Coolant System Reduced Inventory Operations2. INC-6269 Calibration of the Mansell RCS Measurement System3. SOP-l0lA/B Reactor Coolant System4. Technical Specifications 3.9.4I Page 81 of 276 ATTACHMENT 1EAL Bases5. OPT-408A/B Refueling Containment Penetration Verification6. NEI 99-01 CS1Page 82 of 276 ATTACHMENT 1EAL BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 1 -RCS LevelInitiating Condition: Loss of RCS inventory affecting core decay heat removal capabilityEAL:CS1.2 Site Area EmergencyWith CONTAINMENT CLOSURE established, RCS level less than or equal to 0 in. aboveupper core plate (top)Mode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):CONTAINMENT CLOSURE -The procedurally defined actions taken to secure containmentand its associated structures, systems, and components as a functional barrier to fissionproduct release under shutdown conditions. Containment closure means that all potentialescape paths are closed or capable of being closed:A. All penetrations providing direct access from Containment atmosphere to outsideatmosphere are closed except:* Penetrations with automatic valves capable of being closed by an operable CVI" Penetrations under administrative controls (e.g., Control Room notified and designatedperson to close if required by fuel handling accident)B. Equipment hatch is closed and held in place by 4 bolts, or is capable of being closed andheld in place by 4 boltsC. One emergency airlock door is closedD. One personnel airlock door is capable of being closedCPNPP Basis:When Reactor Vessel water level drops below 0 in. above upper core plate (top) 823'-0"elevation (ref. 1), core uncovery is about to occur. RCS level can be monitored by one or moreof the following (ref. 1, 2, 3):" RCS Level Wide Range u-LI-3615B* RCS Level Narrow Range uj-LI-3615A" RCS Extended Wide Range u-LI-3615C* Mansell Level Monitor System u-LT-3619A/B/C-1, -2" Plant Computer* RVLIS* Ultrasonic Level monitoring (optional)Under the conditions specified by this EAL, continued lowering of Reactor Vessel water level isindicative of a loss of inventory control. Inventory loss may be due to a vessel breach, RCSpressure boundary leakage or continued boiling in the Reactor Vessel. The magnitude of thisPage 83 of 276 ATTACHMENT 1EAL Basesloss of water indicates that makeup systems have not been effective and may not be capableof preventing further RCS or Reactor Vessel water level drop and potential core uncovery. Theinability to restore and maintain level after reaching this setpoint infers a failure of the RCSbarrier and Potential Loss of the Fuel Clad barrier.The status of Containment closure is tracked if plant conditions change that could raise the riskof a fission product release as a result of a loss of decay heat removal (ref. 4, 5).NEI 99-01 Basis:This IC addresses a significant and prolonged loss of (reactor vessel/RS-&RCS_[PWR]-r RPV[-3,WRI}-inventory control and makeup capability leading to IMMINENT fuel damage. The lostinventory may be due to a RCS component failure, a loss of configuration control or prolongedboiling of reactor coolant. These conditions entail major failures of plant functions needed forprotection of the public and thus warrant a Site Area Emergency declaration.Following an extended loss of core decay heat removal and inventory makeup, decay heat willcause reactor coolant boiling and a further reduction in reactor vessel level. If RGS~eaetorYesse4RCS level cannot be restored, fuel damage is probable.Outage/shutdown contingency plans typically provide for re-establishing or verifyingCONTAINMENT CLOSURE following a loss of heat removal or RCS inventory controlfunctions. The difference in the specified RCS/reactor vessel levels of EALs bCS1.1 and2-.bCS1.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lowerprobability of a fission product release to the environment.1in EAL 3.a, the 30 m~inu-te c-riterion is tied to a readily recognizable event start time (i.e., thetotal less of ability to m onGitor level), and .....uffici, rntime to monitor, assess and corelatereoac-tor and plant conditions to determine if coro uncover,' has actually occurred (i.e., tosuffiient time for perfoermance of ac-tions; to terminate leakage, recover inventor!control/makeup equipment and/or restore level moenitoring-.The inability to mon~itor RCS (reactorF vesscl/RGS [PWVR] or RPV~ [BWR]) level may be causedby instrumentation and/or power failures, or water level dropping belo thag F availableinstrum~entation. if water level cannot be monRitored, operatorns may dote rmine that aninventor,' losi ocuring by obser~ing changes in sump and/or tank levels. Sump and/ortank level changes must be evaluated against other potential sources of water flow to ensurethey are indicative of leakage fromn the (reactor vessel/RCS [PVWR] or RPV [BWRJ).These-This EALs addresses concerns raised by Generic Letter 88-17, Loss of Decay HeatRemoval; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449,Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the UnitedStates; and NUMARC 91-06, Guidelines for Industry Actions to Assess ShutdownManagement.Escalation of the emergency classification level would be via IC CG1 or AG--RG1CPNPP Basis Reference(s):1. IPO-01OA/B Reactor Coolant System Reduced Inventory Operations2. INC-6269 Calibration of the Mansell RCS Measurement System3. SOP-101A/B Reactor Coolant SystemPage 84 of 276 1 ATTACHMENT 1EAL Bases4. Technical Specifications 3.9.45. OPT-408A/B Refueling Containment Penetration Verification6. NEI 99-01 CS1Page 85 of 276 1 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:EAL:C -Cold Shutdown / Refueling System Malfunction1 -RCS LevelLoss of RCS inventory affecting core decay heat removal capabilityCS1.3 Site Area EmergencyRCS water level cannot be monitored for greater than or equal to 30 min. (Note 1)ANDCore uncovery is indicated by any of the following:" UNPLANNED increase in any Table C-1 sump/tank level of sufficient magnitude toindicate core uncovery* Erratic Source Range Monitor indication* greater than 20,000 R/hr on any of the following:-CTEu16, Containment HRRM (u-RE-6290A)-CTWu17, Containment HRRM (u-RE-6290B)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has beenexceeded, or will likely be exceeded.Table C-1 Sumps / Tanks* Containment Sump 1* Containment Sump 2* Reactor Cavity Sump* CCW Surge TankA* CCW Surge Tank B* PRT* RCDTMode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally.UNPLANNED-. A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient. The cause of the parameter changeor event may be known or unknown.CPNPP Basis:In Cold Shutdown mode, the RCS will normally be intact and standard RCS level monitoringmeans are available.Page 86 of 276 ATTACHMENT IEAL BasesIn the Refueling mode, the RCS is not intact and RCS level may be monitored by differentmeans, including the ability to monitor level visually.In this EAL, all RCS water level indication would be unavailable for greater than 30 minutes,and the RCS inventory loss must be detected by indirect leakage indications (Table C-1).Surveillance procedures provide instructions for calculating primary system leak rate bymanual or computer-based water inventory balances. Level increases must be evaluatedagainst other potential sources of leakage such as cooling water sources inside thecontainment to ensure they are indicative of RCS leakage. If the make-up rate to the RCSunexplainably rises above the pre-established rate, a loss of RCS inventory may be occurringeven if the source of the leakage cannot be immediately identified (ref. 1, 2).The RCS inventory loss may be detected by the Containment High Range Radiation Monitor(HRRM) or erratic Source Range Monitor indication. As water level in the Reactor Vessellowers, the dose rate above the core will rise. The dose rate due to this core shine shouldresult in Containment High Range Radiation Monitor indication greater than 20,000 R/hr (ref.3). The Containment HRRMs have a range of IE-1 to 1E+8 R/hr.Post-TMI accident studies indicated that the installed PWR nuclear instrumentation will operateerratically when the core is uncovered and that this should be used as a tool for making suchdeterminations (ref. 4, 5).NEI 99-01 Basis:This IC addresses a significant and prolonged loss of (reater- ves, .RCSfPWR-,-RPV.)-inventory control and makeup capability leading to IMMINENT fuel damage. The lostinventory may be due to a RCS component failure, a loss of configuration control or prolongedboiling of reactor coolant. These conditions entail major failures of plant functions needed forprotection of the public and thus warrant a Site Area Emergency declaration.Following an extended loss of core decay heat removal and inventory makeup, decay heat willcause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactervessel level cannot be restored, fuel damage is probable.CONTAINMENT CLOSURE follGoI g a loss of heat removal or RCS , controlfunctionS. The diffeMrene in the specified RGS/reactor vessel levels of EALS 1 .b and 2.b reflectthe fact that with CONTAINIMlENIT CGLOSURE established, ther Is a lower pro.bability of afission product release to the envieroment.In EAL 3.aa-The 30-minute criterion is tied to a readily recognizable event start time (i.e., thetotal loss of ability to monitor level), and allows sufficient time to monitor, assess and correlatereactor and plant conditions to determine if core uncovery has actually occurred (i.e., toaccount for various accident progression and instrumentation uncertainties). It also allowssufficient time for performance of actions to terminate leakage, recover inventorycontrol/makeup equipment and/or restore level monitoring.The inability to monitor RCS (reactor vossel/ROS [PWR] or RPV [BDD 'R]) level may be causedby instrumentation and/or power failures, or water level dropping below the range of availableinstrumentation. If water level cannot be monitored, operators may determine that aninventory loss is occurring by observing changes in sump and/or tank levels. Sump and/orPage 87 of 276 ATTACHMENT IEAL Basestank level changes must be evaluated against other potential sources of water flow to ensurethey are indicative of leakage from the RCS (rca.te. v...e...lRCS [PWR] or RPV [BWVVR]).These-This EALs addresses concerns raised by Generic Letter 88-17, Loss of Decay HeatRemoval; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449,Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the UnitedStates; and NUMARC 91-06, Guidelines for Industry Actions to Assess ShutdownManagement.Escalation of the emergency classification level would be via IC CG1 or AG-1-RG1CPNPP Basis Reference(s):1. ABN-103 Excessive Reactor Coolant Leakage2. ABN-108 Shutdown Loss of Coolant3. Engineering Handbook, Guidelines for Events Beyond Design Basis: Spent Fuel Pools,Figure D "Dose Rate at Elevation 860' above Stored Fuel vs. Water Level Depth in SFP"4. Severe Accident Management Guidance Technical Basis Report, Volume 1: CandidateHigh-Level Actions and Their Effects, pgs 2-18, 2-195. Nuclear Safety Analysis Center (NSAC), 1980, "Analysis of Three Mile Island -Unit 2Accident," NSAC-16. NEI 99-01 CS1I Page 88 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:C -Cold Shutdown / Refueling System MalfunctionI -RCS LevelLoss of RCS inventory affecting fuel clad integrity with containmentchallengedEAL:CGI.1 General EmergencyRCS level less than or equal to 0 in. above upper core plate (top) for _> 30 min. (Note 1)ANDAny Containment Challenge indication, Table C-2Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declarationof a General Emergency is not required.Table C-2 Containment Challenge Indications* CONTAINMENT CLOSURE not established(Note 6)* Containment hydrogen concentration greaterthan 4%* Unplanned rise greater than 1 psig inContainment pressureMode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):CONTAINMENT CLOSURE -The procedurally defined actions taken to secure containmentand its associated structures, systems, and components as a functional barrier to fissionproduct release under shutdown conditions. Containment closure means that all potentialescape paths are closed or capable of being closed:A. All penetrations providing direct access from Containment atmosphere to outsideatmosphere are closed except:" Penetrations with automatic valves capable of being closed by an operable CVI" Penetrations under administrative controls (e.g., Control Room notified and designatedperson to close if required by fuel handling accident)B. Equipment hatch is closed and held in place by 4 bolts, or is capable of being closed andheld in place by 4 boltsC. One emergency airlock door is closedD. One personnel airlock door is capable of being closedPage 89 of 276 ATTACHMENT 1EAL BasesCPNPP Basis:When Reactor Vessel water level drops below 0 in. above upper core plate (top) 823'-0"elevation (ref. 1), core uncovery is about to occur. RCS level can be monitored by one or moreof the following (ref. 1, 2, 3):" RCS Level Wide Range u-LI-3615B" RCS Level Narrow Range u-LI-3615A* RCS Extended Wide Range u-LI-3615C* Mansell Level Monitor System u-LT-3619A/B/C-1, -2" Plant Computer* RVLIS* Ultrasonic Level monitoring (optional)Under the conditions specified by this EAL, continued lowering of Reactor Vessel water level isindicative of a loss of inventory control. Inventory loss may be due to a vessel breach, RCSpressure boundary leakage or continued boiling in the Reactor Vessel. The magnitude of thisloss of water indicates that makeup systems have not been effective and may not be capableof preventing further RCS or Reactor Vessel water level drop and potential core uncovery. Theinability to restore and maintain level after reaching this setpoint infers a failure of the RCSbarrier and Potential Loss of the Fuel Clad barrier.Three conditions are associated with a challenge to Containment integrity:1. CONTAINMENT COSURE not established -The status of Containment closure istracked if plant conditions change that could raise the risk of a fission product release asa result of a loss of decay heat removal (ref. 4, 5). If containment closure is re-established prior to exceeding the 30 minute core uncovery time limit then escalation toGE would not occur.2. Containment hydrogen greater than 4% -The 4% hydrogen concentration threshold isgenerally considered the lower limit for hydrogen deflagrations. CPNPP is equippedwith a Hydrogen Control System (HCS) which serves to limit or reduce combustible gasconcentrations in the Containment. The plant has two hydrogen monitoring systems.Each monitoring system consists of four sensor modules and one microprocessoranalyzer. Two sensors from each Containment are coupled to one of the two hydrogenmicroprocessors located in the Control Room. Thus each microprocessor analyzer isshared by Units 1 and 2. The analyzer system has a range of 0-10% hydrogen byvolume. The detector modules are located on the 905', 873', and 860' elevations inContainment. A fourth detector is located on 832' level across from the loop roomentrance for loops 1 and 4. Hydrogen concentration is displayed in the Control Room onu-AI-5506A/B and u-AI-5506C/D. Hydrogen concentration can also be displayed on thePlant Computer. Alarms at -3% are provided for high hydrogen concentration,u-ALB-3A, window 3.7. If a hydrogen concentration value can not be obtained from thehydrogen monitoring system, a grab sample from the containment PIG radiation monitormay be used to determine the hydrogen concentration (ref. 6, 7, 8, 9).3. UNPLANNED rise in Containment pressure -An unplanned pressure rise incontainment while in cold Shutdown or Refueling modes can threaten ContainmentPage 90 of 276 ATTACHMENT 1EAL BasesClosure capability and thus Containment potentially cannot be relied upon as a barrierto fission product release.NEI 99-01 Basis:This IC addresses the inability to restore and maintain reactor vessel level above the top ofactive fuel with containment challenged. This condition represents actual or IMMINENTsubstantial core degradation or melting with potential for loss of containment integrity.Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for morethan the immediate site area.Following an extended loss of core decay heat removal and inventory makeup, decay heat willcause reactor coolant boiling and a further reduction in reactor vessel level. If RCSRCS/reactor vessel level cannot be restored, fuel damage is probable.With CONTAINMENT CLOSURE not established, there is a high potential for a direct andunmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a GeneralEmergency is not required.The existence of an explosive mixture means, at a minimum, that the containment atmospherichydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagrationlimit). A hydrogen burn will raise containment pressure and could result in collateral equipmentdamage leading to a loss of containment integrity. It therefore represents a challenge toContainment integrity.In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a coreuncovery could result in an explosive gas mixture in containment. If all installed hydrogen gasmonitors are out-of-service during an event leading to fuel cladding damage, it may not bepossible to obtain a containment hydrogen gas concentration reading as ambient conditionswithin the containment will preclude personnel access. During periods when installedcontainment hydrogen gas monitors are out-of-service, operators may use the other listedindications to assess whether or not containment is challenged.In 30-minute criterion is tied to a readily recognizable event start time (i.e., thetotal loss of ability to monitor level), and allows sufficient time to monitor, assess and correlatereactor and plant conditions to determine if core uncovery has actually occurred (i.e., toaccount for various accident progression and instrumentation uncertainties). It also allowssufficient time for performance of actions to terminate leakage, recover inventorycontrol/makeup equipment and/or restore level monitoring.The inability ton monitor (reactor [PAqrI or RPV RCSJr'I'RJ) level may be causedby itrmn atinad/or power failures, or water level dropping below the range of availab-leinstrum~entation. if water level cannot be monflitored, operators ma" determine that aninventor,' loss is occurring by obseR'ing changsinsm and/or tank levels. Sump and/ottank level changes must be evaluatedagaistothetl I al sources of water floewto the" are indicativ~e of leakage from the (reactor ':ossel! RCS [PW4R] or RPV [BIA'R]).Thisese EALs addresses concerns raised by Generic Letter 88-17, Loss of Decay HeatRemoval; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449,Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the UnitedStates; and NUMARC 91-06, Guidelines for Industry Actions to Assess ShutdownPage 91 of 276 ATTACHMENT IEAL BasesManagement.IPage 92 of 276 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):1. IPO-010A/B Reactor Coolant System Reduced Inventory Operations2. INC-6269 Calibration of the Mansell RCS Measurement System3. SOP-101A/B Reactor Coolant System4. Technical Specifications 3.9.45. OPT-408A/B Refueling Containment Penetration Verification6. FRC-0.1A/B Response to Inadequate Core Cooling, Attachment 57. FSAR Section 6.2.58. FSAR Table 7.5-7A9. CHM-1 11 Primary Chemistry Accident Assessment Sampling Program10. NEI 99-01 CS1I Page 93 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:C -Cold Shutdown / Refueling System Malfunction1 -RCS LevelLoss of RCS inventory affecting fuel clad integrity with containmentchallengedEAL:CG1.2 General EmergencyRCS level cannot be monitored for greater than or equal to 30 min. (Note 1)ANDCore uncovery is indicated by any of the following:" UNPLANNED increase in any Table C-1 sump/tank level of sufficient magnitude toindicate core uncovery* Erratic Source Range Monitor indication" Greater than 20,000 R/hr on any of the following:-CTEu16, Containment HRRM (u-RE-6290A)-CTWu17, Containment HRRM (u-RE-6290B)ANDAny Containment Challenge indication, Table C-2Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration ofa General Emergency is not required.Table C-1 Sumps / Tanks* Containment Sump I* Containment Sump 2* Reactor Cavity Sump* CCW Surge TankA* CCW Surge Tank B* PRT* RCDTPage 94 of 276 ATTACHMENT 1EAL BasesTable C-2 Containment Challenge Indications* CONTAINMENT CLOSURE not established(Note 6)" Containment hydrogen concentration greaterthan 4%* Unplanned rise greater than 1 psig inContainment pressureMode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):CONTAINMENT CLOSURE -The procedurally defined actions taken to secure containmentand its associated structures, systems, and components as a functional barrier to fissionproduct release under shutdown conditions. Containment closure means that all potentialescape paths are closed or capable of being closed:A. All penetrations providing direct access from Containment atmosphere to outsideatmosphere are closed except:" Penetrations with automatic valves capable of being closed by an operable CVI" Penetrations under administrative controls (e.g., Control Room notified and designatedperson to close if required by fuel handling accident)B. Equipment hatch is closed and held in place by 4 bolts, or is capable of being closed andheld in place by 4 boltsC. One emergency airlock door is closedD. One personnel airlock door is capable of being closedUNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally.UNPLANNED-. A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient. The cause of the parameter changeor event may be known or unknown.CPNPP Basis:In Cold Shutdown mode, the RCS will normally be intact and standard RCS level monitoringmeans are available.In the Refueling mode, the RCS is not intact and RCS level may be monitored by differentmeans, including the ability to monitor level visually.In this EAL, all RCS water level indication would be unavailable for greater than 30 minutes,and the RCS inventory loss must be detected by indirect leakage indications (Table C-1).Surveillance procedures provide instructions for calculating primary system leak rate bymanual or computer-based water inventory balances. Level increases must be evaluatedagainst other potential sources of leakage such as cooling water sources inside thecontainment to ensure they are indicative of RCS leakage. If the make-up rate to the RCSPage 95 of 276 ATTACHMENT 1EAL Basesunexplainably rises above the pre-established rate, a loss of RCS inventory may be occurringeven if the source of the leakage cannot be immediately identified (ref. 1, 2).The RCS inventory loss may be detected by the Containment High Range Radiation Monitor(HRRM) or erratic Source Range Monitor indication. As water level in the Reactor Vessellowers, the dose rate above the core will rise. The dose rate due to this core shine shouldresult in Containment High Range Radiation Monitor indication greater than 20,000 R/hr (ref.3). The Containment HRRMs have a range of 1 E-1 to 1 E+8 R/hr.Post-TMI accident studies indicated that the installed PWR nuclear instrumentation will operateerratically when the core is uncovered and that this should be used as a tool for making suchdeterminations (ref. 4, 5).Three conditions are associated with a challenge to Containment integrity:1. CONTAINMENT COSURE not established -The status of Containment closure istracked if plant conditions change that could raise the risk of a fission product release asa result of a loss of decay heat removal (ref. 6, 7). If containment closure is re-established prior to exceeding the 30 minute core uncovery time limit then escalation toGE would not occur.2. Containment hydrogen greater than 4% -The 4% hydrogen concentration threshold isgenerally considered the lower limit for hydrogen deflagrations. CPNPP is equippedwith a Hydrogen Control System (HCS) which serves to limit or reduce combustible gasconcentrations in the Containment. The plant has two hydrogen monitoring systems.Each monitoring system consists of four sensor modules and one microprocessoranalyzer. Two sensors from each Containment are coupled to one of the two hydrogenmicroprocessors located in the Control Room. Thus each microprocessor analyzer isshared by Units 1 and 2. The analyzer system has a range of 0-10% hydrogen byvolume. The detector modules are located on the 905', 873', and 860' elevations inContainment. A fourth detector is located on 832' level across from the loop roomentrance for loops 1 and 4. Hydrogen concentration is displayed in the Control Room onu-AI-5506A/B and u-AI-5506C/D. Hydrogen concentration can also be displayed on thePlant Computer. Alarms at -3% are provided for high hydrogen concentration,u-ALB-3A, window 3.7. If a hydrogen concentration value can not be obtained from thehydrogen monitoring system, a grab sample from the containment PIG radiation monitormay be used to determine the hydrogen concentration (ref. 8, 9, 10, 11).3. UNPLANNED rise in Containment pressure -An unplanned pressure rise incontainment while in cold Shutdown or Refueling modes can threaten ContainmentClosure capability and thus Containment potentially cannot be relied upon as a barrierto fission product release.NEI 99-01 Basis:This IC addresses the inability to restore and maintain reactor vessel level above the top ofactive fuel with containment challenged. This condition represents actual or IMMINENTsubstantial core degradation or melting with potential for loss of containment integrity.Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for morethan the immediate site area.Following an extended loss of core decay heat removal and inventory makeup, decay heat willcause reactor coolant boiling and a further reduction in reactor vessel level. IfRCSPage 96 of 276 ATTACHMENT 1EAL Bases9RCS/Fractor Vosel level cannot be restored, fuel damage is probable.With CONTAINMENT CLOSURE not established, there is a high potential for a direct andunmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a GeneralEmergency is not required.The existence of an explosive mixture means, at a minimum, that the containment atmospherichydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagrationlimit). A hydrogen burn will raise containment pressure and could result in collateral equipmentdamage leading to a loss of containment integrity. It therefore represents a challenge toContainment integrity.In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a coreuncovery could result in an explosive gas mixture in containment. If all installed hydrogen gasmonitors are out-of-service during an event leading to fuel cladding damage, it may not bepossible to obtain a containment hydrogen gas concentration reading as ambient conditionswithin the containment will preclude personnel access. During periods when installedcontainment hydrogen gas monitors are out-of-service, operators may use the other listedindications to assess whether or not containment is challenged.in -=n-lA 2.bt-The 30-minute criterion is tied to a readily recognizable event start time (i.e., thetotal loss of ability to monitor level), and allows sufficient time to monitor, assess and correlatereactor and plant conditions to determine if core uncovery has actually occurred (i.e., toaccount for various accident progression and instrumentation uncertainties). It also allowssufficient time for performance of actions to terminate leakage, recover inventorycontrol/makeup equipment and/or restore level monitoring.The inability to monitor (reactor ..esscl,.RGS [Pr4'R] or RPV-RCS.BrK,) level may be causedby instrumentation and/or power failures, or water level dropping below the range of availableinstrumentation. If water level cannot be monitored, operators may determine that aninventory loss is occurring by observing changes in sump and/or tank levels. Sump and/ortank level changes must be evaluated against other potential sources of water flow to ensurethey are indicative of leakage from the (reaGte .vesse./RNCS [P-K/ or RPV [BI\'R,).Thisese EALs addresses concerns raised by Generic Letter 88-17, Loss of Decay HeatRemoval; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1 449,Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the UnitedStates; and NUMARC 91-06, Guidelines for Industry Actions to Assess ShutdownManagement.Page 97 of 276 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):1. ABN-1 03 Excessive Reactor Coolant Leakage2. ABN-108 Shutdown Loss of Coolant3. Engineering Handbook, Guidelines for Events Beyond Design Basis: Spent Fuel Pools,Figure D "Dose Rate at Elevation 860' above Stored Fuel vs. Water Level Depth in SFP"4. Severe Accident Management Guidance Technical Basis Report, Volume 1: CandidateHigh-Level Actions and Their Effects, pgs 2-18, 2-195. Nuclear Safety Analysis Center (NSAC), 1980, "Analysis of Three Mile Island -Unit 2Accident," NSAC-16. Technical Specifications 3.9.47. OPT-408A/B Refueling Containment Penetration Verification8. FRC-O.1A/B Response to Inadequate Core Cooling, Attachment 59. FSAR Section 6.2.510. FSAR Table 7.5-7A11. CHM-1 II Primary Chemistry Accident Assessment Sampling Program12. NEI 99-01 CG1Page 98 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:C -Cold Shutdown / Refueling System Malfunction2 -Loss of Emergency AC PowerLoss of all but one AC power source to safeguard buses for 15minutes or longerEAL:CU2.1 Unusual EventAC power capability, Table C-3, to 6.9 KV safeguard buses uEA1 and uEA2 reduced to asingle power source for greater than or equal to 15 min. (Note 1)ANDAny additional single Table C-3 power source failure will result in loss of all AC power toSAFETY SYSTEMSNote 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Table C-3 AC Power SourcesOffsite:* 138 KV switchyard circuit* 345 KV switchyard circuitOnsite:" uEG1* uEG2Mode Applicability:5 -Cold Shutdown, 6- Refueling, D -DefueledDefinition(s):SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/orplacing it in the cold shutdown condition, including the ECCS. These are typically systemsclassified as safety-related (as defined in 10CFR50.2):Those structures, systems and components that are relied upon to remain functional duringand 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 could resultin potential offsite exposures.Page 99 of 276 ATTACHMENT 1EAL BasesCPNPP Basis:For emergency classification purposes, "capability" means that an offsite AC power source(s)is available to the emergency buses, whether or not the buses are powered from it.The condition indicated by this EAL is the degradation of the offsite and onsite power sourcessuch that any additional single failure would result in a loss of all AC power to the emergencybuses.The safeguards AC distribution system power sources consist of the preferred and alternateoffsite power sources, and the onsite standby emergency diesel generators uEGI and uEG2.Offsite power is supplied to the plant switchyards from the transmission network by five 345 KVand two 138 KV transmission lines. From the switchyards, two electrically and physicallyseparated circuits provide AC power through step down startup transformers, to the 6.9 kVsafeguard buses. The 138 kV switchyard circuit is the preferred source for Unit 2 and alternatesource for Unit 1. The 345 KV circuit is the preferred source for Unit 1 and alternate source forUnit 2. The onsite AC distribution system is divided into redundant trains so that the loss of anyone load group does not prevent the minimum safety functions from being performed. Eachtrain has connections to two offsite power sources and a dedicated diesel generator. Eachoffsite circuit can supply the Unit 1 and Unit 2 6.9 KV safeguard buses. (ref. 1, 2, 3, 4).This cold condition EAL is equivalent to the hot condition EAL SAI.1.NEI 99-01 Basis:This IC describes a significant degradation of offsite and onsite AC power sources such thatany additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. Inthis condition, the sole AC power source may be powering one, or more than one, train ofsafety-related equipment.When in the cold shutdown, refueling, or defueled mode, this condition is not classified as anAlert because of the increased time available to restore another power source to service.Additional time is available due to the reduced core decay heat load, and the lowertemperatures and pressures in various plant systems. Thus, when in these modes, thiscondition is considered to be a potential degradation of the level of safety of the plant.An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplyingrequired power to an essential bus. Some examples of this condition are presented below.* A loss of all offsite power with a concurrent failure of all but one emergency powersource (e.g., an onsite diesel generator)." A loss of all offsite power and loss of all emergency power sources (e.g., onsite dieselgenerators) with a single train of emergency buses being back-fed from the unit maingenerator.* A loss of emergency power sources (e.g., onsite diesel generators) with a single train ofemergency buses being back-fed from an offsite power source.Fifteen minutes was selected as a threshold to exclude transient or momentary losses ofpower.The subsequent loss of the remaining single power source would escalate the event to an Alertin accordance with IC CA2.Page 100 of 276 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):1.2.3.4.5.6.7.FSAR Figure 8.3-1FSAR Section 8.2FSAR Section 8.3Technical Specifications B3.8.1ABN-601 Response to a 138/345 KV System MalfunctionABN-602 Response to a 6900/480V System MalfunctionNEI 99-01 CU2Page 101 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:C -Cold Shutdown / Refueling System Malfunction2 -Loss of Emergency AC PowerLoss of all offsite and all onsite AC power to safeguard buses for 15minutes or longerEAL:CA2.1 AlertLoss of all offsite and all onsite AC power capability, Table C-3, to 6.9 KV safeguardbuses uEA1 and uEA2 for greater than or equal to 15 min. (Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has beenexceeded, or will likely be exceeded.Table C-3 AC Power SourcesOffsite:* 138 KV switchyard circuit* 345 KV switchyard circuitOnsite:" uEG1* uEG2Mode Applicability:5 -Cold Shutdown, 6 -Refueling, D -DefueledCPNPP Basis:For emergency classification purposes, "capability" means that an offsite AC power source(s)is available to the emergency buses, whether or not the buses are powered from it.The safeguards AC distribution system power sources consist of the preferred and alternateoffsite power sources, and the onsite standby emergency diesel generators uEGI and uEG2.Offsite power is supplied to the plant switchyards from the transmission network by five 345 KVand two 138 KV transmission lines. From the switchyards, two electrically and physicallyseparated circuits provide AC power through step down startup transformers, to the 6.9 kVsafeguard buses. The 138 kV switchyard circuit is the preferred source for Unit 2 and alternatesource for Unit 1. The 345 KV circuit is the preferred source for Unit I and alternate source forUnit 2. The onsite AC distribution system is divided into redundant trains so that the loss of anyone load group does not prevent the minimum safety functions from being performed. Eachtrain has connections to two offsite power sources and a dedicated diesel generator. Eachoffsite circuit can supply the Unit 1 and Unit 2 6.9 KV safeguard buses. (ref. 1, 2, 3, 4)This cold condition EAL is equivalent to the hot condition loss of all offsite AC power EALSS1.1.I Page 102 of 276 1 ATTACHMENT 1EAL BasesNEI 99-01 Basis:This IC addresses a total loss of AC power that compromises the performance of all SAFETYSYSTEMS requiring electric power including those necessary for emergency core cooling,containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.When in the cold shutdown, refueling, or defueled mode, this condition is not classified as aSite Area Emergency because of the increased time available to restore an emergency bus toservice. Additional time is available due to the reduced core decay heat load, and the lowertemperatures and pressures in various plant systems. Thus, when in these modes, thiscondition represents an actual or potential substantial degradation of the level of safety of theplant.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Escalation of the emergency classification level would be via IC CS1 or AS4-RS1.CPNPP Basis Reference(s):1. FSAR Figure 8.3-12. FSAR Section 8.23. FSAR Section 8.34. Technical Specifications B3.8.15. ABN-601 Response to a 138/345 KV System Malfunction6. ABN-602 Response to a 6900/480V System Malfunction7. NEI 99-01 CA2Page 103 of 276 1 ATTACHMENT 1EAL BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 3 -RCS TemperatureInitiating Condition: UNPLANNED increase in RCS temperatureEAL:CU3.1 Unusual EventUNPLANNED increase in RCS temperature to greater than 200OF due to loss of decayheat removal capability (Note 9)Note 9: Begin monitoring hot condition EALs concurrently.Mode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):UNPLANNED-. A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient. The cause of the parameter changeor event may be known or unknown.CPNPP Basis:Several instruments are capable of providing indication of RCS temperature with respect to theTechnical Specification cold shutdown temperature limit (200'F, ref. 1). These include loop Thot(u-TR-413A/23A, u -TR-433A/43A, u -TI-413A, u -TI-423A) and, if no RCPs are operating, theCore Exit Thermocouples (TCs). The most limiting temperature indication should be used. Forexample, during heatup, the highest reading temperature indication should be used; duringcooldown, the lowest (ref. 2, 3, 4, 5).In the absence of reliable RCS temperature indication caused by the loss of decay heatremoval capability, classification should be based on the RCS pressure increase criteria whenthe RCS is intact in Mode 5 or based on time to boil data when in Mode 6 or the RCS is notintact in Mode 5.The note is a reminder that any temperature increase above 2001F is an operating modechange from cold to hot conditions. Since each EAL is associated with operating modeapplicability, the set of EALs that must be monitored must now include EALs associated withhot condition operating modes.NEI 99-01 Basis:This IC addresses an UNPLANNED increase in RCS temperature above the TechnicalSpecification cold shutdown temperature limit.,or the inability to determinc RCS t.mpc.atur.aadlevel-and represents a potential degradation of the level of safety of the plant. If the RGSRCS-is not intact and CONTAINMENT CLOSURE is not established during this event, theEmergency Director Emerqency Coordinator should also refer to IC CA3.A momentary UNPLANNED excursion above the Technical Specification cold shutdowntemperature limit when the heat removal function is available does not warrant a classification.Page 104 of 276 ATTACHMENT 1EAL BasesI EAL-#4This EAL involves a loss of decay heat removal capability, or an addition of heat to theRCS in excess of that which can currently be removed, such that reactor coolant temperaturecannot be maintained below the cold shutdown temperature limit specified in TechnicalSpecifications. During this condition, there is no immediate threat of fuel damage because thecore decay heat load has been reduced since the cessation of power operation.During an outage, the level in the reactor vessel will normally be maintained at or above thereactor vessel flange. Refueling evolutions that lower water level below the reactor vesselflange are carefully planned and controlled. A loss of forced decay heat removal at reducedinventory may result in a rapid increase in reactor coolant temperature depending on the timeafter shutdown.EAL #2 reflects a onldition whore there has been a Significant loss orf instrmentationcapability ecessary t on moritor RCS pcoditions and operators would be unable to moaitor keypa1.reters neessarp' to assure core decay heat removal. During this condition, there is noMimmednate threat of fuel damnage beause the core decay heat load has been reduced incethe cessation of power operatien.Fifteen mninutes was selected as a threshold to exclude tr-anSient or moementar,' loesof IRdeation-Escalation to Alert would be via IC CAl based on an inventory loss or IC CA3 based onexceeding plant configuration-specific time criteria.CPNPP Basis Reference(s):1 .Technical Specifications Table 1.1-12. IPO-OO5N/B Plant Cooldown From Hot Standby To Cold Shutdown3. Technical Specifications 3.4.34. OPT-407 RCS Pressure and Temperature Verification5. IPO-01 OA/B Reactor Coolant System Reduced Inventory Condition6. NEI 99-01 CU3Page 105 of 276 1 ATTACHMENT 1EAL BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 3 -RCS TemperatureInitiating Condition: UNPLANNED increase in RCS temperatureEAL:CU3.2 Unusual EventLoss of all RCS temperature and RCS level indication for greater than or equal to 15 min.(Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Mode Applicability:5 -Cold Shutdown, 6- RefuelingDefinition(s):NoneCPNPP Basis:RCS level can be monitored by one or more of the following (ref. 2, 3, 4):" RCS Level Wide Range u-LI-3615B* RCS Level Narrow Range u-LI-3615A* RCS Extended Wide Range u-LI-3615C* Mansell Level Monitor System u-LT-3619A/B/C-1, -2* Plant Computer* RVLIS* Ultrasonic Level monitoring (optional)Several instruments are capable of providing indication of RCS temperature with respect to theTechnical Specification cold shutdown temperature limit (200'F, ref. 1). These include loop Thot(u-TR-413A/23A, u -TR-433A/43A, u -TI-413A, u -TI-423A) and, if no RCPs are operating, theCore Exit Thermocouples (TCs). The most limiting temperature indication should be used. Forexample, during heatup, the highest reading temperature indication should be used; duringcooldown, the lowest (ref. 5, 6, 7, 8).In the absence of reliable RCS temperature indication caused by the loss of decay heatremoval capability, classification should be based on heat up rate data or additionally if inMode 5 with RCS intact on pressure increase.NEI 99-01 Basis:This IG-EAL addresses an UNPLANNED i en RS temperature abo'... the TechniGalSpecific3tio c.ld. shutdown temperature limit, or the inability to determine RCS temperaturePage 106 of 276 ATTACHMENT 1EAL Basesand level, and represents a potential degradation of the level of safety of the plant. If the RCSis not intact and CONTAINMENT CLOSURE is not established during this event, theEmergency DiroctorEmerqency Coordinator should also refer to IC CA3.A tar,' UNPLANNED ,Xcur"sin above the Rpecification cold shutd-n,+ nnr 4..Ir i n-,. *.,kn 4 k kn 4- .-rn ,-,I.v 4n4.a H-. ,-I.k1 AlC +in i M, +;4E=AL #1 involves a loss of decay heat remoeval capability, or an addition of heat to theRCS in exGess ef that Which can currntly b8 rFemroved, such tha-t coolart temperatuFecannot be mnaintained below the coldG sh-utdown40 temperature limit specified in Technic~alSpecifications. DUFiRg this conditinR, there is no immediate threat of fuel damage because thecore decay heat lead has been reduced s the cessatio of power opeFatinr.Duarig an outage, the levol in the reactor ve-s-sel will nrmally be Maintained above thereactor vessel flange. Refueling evoluitions that lower water level below the reactor vesselflange are carefully planned and controlled. A loss- o-f forc-ed decay heat Frnemval at reducedinventei' may result in a rapid increase ireactor coolant temperature depending en the timeafter shutdown.EAl= #2This EAL reflects a condition where there has been a significant loss of instrumentationcapability necessary to monitor RCS conditions and operators would be unable to monitor keyparameters necessary to assure core decay heat removal. During this condition, there is noimmediate threat of fuel damage because the core decay heat load has been reduced sincethe cessation of power operation.Fifteen minutes was selected as a threshold to exclude transient or momentary losses ofindication.Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based onexceeding plant configuration-specific time criteria.CPNPP. Basis Reference(s):1. Technical Specifications Table 1 .1-12. IPO-01OA/B Reactor Coolant System Reduced Inventory Operations3. INC-6269 Calibration of the Mansell RCS Measurement System4. SOP-101A/B Reactor Coolant System5. IPO-005A/B Plant Cooldown From Hot Standby To Cold Shutdown6. Technical Specifications 3.4.37. OPT-407 RCS Pressure and Temperature Verification8. IPO-01OA/B Reactor Coolant System Reduced Inventory Condition9. NEI 99-01 CU3Page 107 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:C -Cold Shutdown / Refueling System Malfunction3 -RCS TemperatureInitiating Condition: Inability to maintain plant in cold shutdownEAL:CA3.1 AlertUNPLANNED increase in RCS temperature to greater than 200OF for greater than TableC-4 duration(Notes 1, 9)ORUNPLANNED RCS pressure increase greater than 10 psig due to a loss of RCS cooling(This EAL does not apply during water-solid plant conditions)Note 1: The Emergency Coordinator should declare the event promptly upon determining that the applicabletime has been exceeded, or will likely be exceeded.Note 9: Begin monitoring hot condition EALs concurrently.Table C-4: RCS Heat-up Duration ThresholdsCONTAINMENTRCS Status CLOSURE Status Heat-up DurationIntact (but notREDUCED N/A 60 min.*INVENTORY)Not intact Established 20 min.*ORREDUCED INVENTORY Not established 0 min.* If an RCS heat removal system is in operation within this time frame and RCS temperature isbeing reduced, the EAL is not applicable.Mode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):CONTAINMENT CLOSURE -The procedurally defined actions taken to secure containmentand its associated structures, systems, and components as a functional barrier to fissionproduct release under shutdown conditions. Containment closure means that all potentialescape paths are closed or capable of being closed:A. All penetrations providing direct access from Containment atmosphere to outsideatmosphere are closed except:" Penetrations with automatic valves capable of being closed by an operable CVI" Penetrations under administrative controls (e.g., Control Room notified and designatedperson to close if required by fuel handling accident)Page 108 of 276 ATTACHMENT 1EAL BasesB. Equipment hatch is closed and held in place by 4 bolts, or is capable of being closed andheld in place by 4 boltsC. One emergency airlock door is closedD. One personnel airlock door is capable of being closedUNPLANNED -. A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient. The cause of the parameter changeor event may be known or unknown.REDUCED INVENTORY -Plant condition when fuel is in the reactor vessel and ReactorCoolant System level is < 80 inches above core plate (829'8").CPNPP Basis:Several instruments are capable of providing indication of RCS temperature with respect to theTechnical Specification cold shutdown temperature limit (200'F, ref. 1). These include loop Thot(u-TR-413A/23A, u -TR-433A/43A, u -TI-413A, u -TI-423A) and, if no RCPs are operating, theCore Exit Thermocouples (TCs). The most limiting temperature indication should be used. Forexample, during heatup, the highest reading temperature indication should be used; duringcooldown, the lowest (ref. 2, 3, 4, 5).In the absence of reliable RCS temperature indication caused by the loss of decay heatremoval capability, classification should be based on heat up rate data or additionally if inMode 5 with RCS intact on pressure increase.A 10 psig RCS pressure increase can be monitored on u-PI-403A and computer pointsP6498A and P6499A (ref. 9, 10).The status of Containment closure is tracked if plant conditions change that could raise the riskof a fission product release as a result of a loss of decay heat removal (ref. 6, 7).The note is a reminder that any temperature increase above 200OF is an operating modechange from cold to hot conditions. Since each EAL is associated with operating modeapplicability, the set of EALs that must be monitored must now include EALs associated withhot condition operating modes.NEI 99-01 Basis:This IC addresses conditions involving a loss of decay heat removal capability or an addition ofheat to the RCS in excess of that which can currently be removed. Either condition representsan actual or potential substantial degradation of the level of safety of the plant.A momentary UNPLANNED excursion above the Technical Specification cold shutdowntemperature limit when the heat removal function is available does not warrant a classification.The RGS-RCSHeat-up Duration Thresholds table addresses an increase in RCS temperaturewhen CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS inventory isreduced (e.g., mid-loop operation i-R PWRs). The 20-minute criterion was included to allowtime for operator action to address the temperature increase.The RCS Heat-up Duration Thresholds table also addresses an increase in RCS temperaturewith the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this conditionsince the intact RCS is providing a high pressure barrier to a fission product release. The 60-Page 109 of 276 ATTACHMENT 1EAL Basesminute time frame should allow sufficient time to address the temperature increase without asubstantial degradation in plant safety.Finally, in the case where there is an increase in RCS temperature, the RCS is not intact or isat reduced inventory-fP-WR4, and CONTAINMENT CLOSURE is not established, no heat-upduration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may bereleased directly into the containment atmosphere and subsequently to the environment, and2) there is reduced reactor coolant inventory above the top of irradiated fuel.EAL4#2The RCS pressure increase threshold provides a pressure-based indication of RCSheat-up in the absence of RCS temperature monitoring capability.Escalation of the emergency classification level would be via IC CS1 or A-S4RS1.CPNPP Basis Reference(s):1. Technical Specifications Table 1.1-12. IPO-005A/B Plant Cooldown From Hot Standby To Cold Shutdown3. Technical Specifications 3.4.34. OPT-407 RCS Pressure and Temperature Verification5. IPO-01OA/B Reactor Coolant System Reduced Inventory Condition6. Technical Specifications 3.9.47. OPT-408A/B Refueling Containment Penetration Verification8. IPO-010A/B Reactor Coolant System Reduced Inventory Operations9. IPO-005AIB Plant Cooldown From Hot Standby To Cold Shutdown10. SOP-101A/B Reactor Coolant System Reduced Inventory11. NEI 99-01 CA3I Page 110 of 276 ATTACHMENT 1EAL BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 4 -Loss of Vital DC PowerInitiating Condition: Loss of vital DC power for 15 minutes or longerEAL:CU4.1 Unusual EventLess than 105 VDC bus voltage indications on Technical Specification required 125 VDCbuses (uED1, uED2, uED3, uED4) for greater than or equal to 15 min. (Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Mode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):NoneCPNPP Basis:The purpose of this EAL is to recognize a loss of DC power compromising the ability to monitorand control the removal of decay heat during cold shutdown or refueling operations. This EALis intended to be anticipatory in as much as the operating crew may not have necessaryindication and control of equipment needed to respond to the loss. The fifteen minute intervalis intended to exclude transient or momentary power losses.The safeguards 125 VDC buses are the Class 1E buses uED1, uED2, uED3 and uED4 (ref. 1).The 125 VDC safeguard distribution system is illustrated in Figure C-2 (ref. 2, 3).Each redundant safeguards 125 VDC system consists of two independent batteries eachhaving one main distribution bus, two static battery chargers (one spare), and local distributionpanels. For Unit 1, batteries BT1 ED1 and BT1 ED3 feed all train A load requirements, whilebatteries BT1 ED2 and BT1 ED4 supply train B load requirements.For Unit 2, batteries BT2ED1 and BT2ED3 feed all train A load requirements, while batteriesBT2ED2 and BT2ED4 supply train B load requirements. There are no bus ties or sharing ofpower supplies between redundant trains (ref. 1).Minimum DC bus voltage is 105 VDC (ref. 4). Bus voltage may be monitored from the followingindications (ref. 6):Control Room Panel CP-10 Annunciator u--ALB-10B Plant ComputerV-1ED1, 125VDC SWITCH PNL 1ED1 VOLT 1.13 V6501A BATT BT1ED1 VOLTV-1 ED2, 125VDC SWITCH PNL 1ED2 VOLT 2.13 V6502A BATT BT1ED2 VOLTV-1ED3, 125VDC SWITCH PNL 1ED3 VOLT 1.9 V6503A BATT BT1ED3 VOLTV-1 ED4, 125VDC SWITCH PNL 1ED4 VOLT 3.9 V6504A BATT BT1ED4 VOLTThis EAL is the cold condition equivalent of the hot condition loss of DC power EAL SS7.1.Page 111 of 276 ATTACHMENT 1EAL BasesNEI 99-01 BasisThis IC addresses a loss of vital DC power which compromises the ability to monitor andcontrol operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode.In these modes, the core decay heat load has been significantly reduced, and coolant systemtemperatures and pressures are lower; these conditions increase the time available to restorea vital DC bus to service. Thus, this condition is considered to be a potential degradation ofthe level of safety of the plant.As used in this EAL, "required" means the vital DC buses necessary to support operation ofthe in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, ifTrain A is out-of-service (inoperable) for scheduled outage maintenance work and Train B isin-service (operable), then a loss of Vital DC power affecting Train B would require thedeclaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant anemergency classification.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Depending upon the event, escalation of the emergency classification level would be via ICCA1 or CA3, or an IC in Recognition Category AR.CPNPP Basis Reference(s):1. FSAR 8.3.22. FSAR Figure 8.3-143. FSAR Figure 8.3-14A4. ECA-O.OA/B Loss of All AC Power5. SOP-605ANB 125 VDC Switchgear and Distribution Systems, Batteries and BatteryChargers6. ALM-0102A/B Alarm Procedures Manual, u-ALB-1OB, nos. 1.9, 1.13, 2.13, 3.87. NEI 99-01 CU4Page 112 of 276 ATTACHMENT IEAL BasesCategory:Subcategory:Initiating Condition:EAL:C -Cold Shutdown / Refueling System Malfunction5 -Loss of CommunicationsLoss of all onsite or offsjte communications capabilitiesCU5.1 Unusual EventLoss of all Table C-5 onsite communication methodsORLoss of all Table C-5 offsite communication methodsORLoss of all Table C-5 NRC communication methodsTable C-5 Communication MethodsSystem Onsite Offsite NRCGai-tronics Page/Party (PA) XPlant Radios XPABX X X XPublic Telephone X X XFederal Telephone System (FTS) X XMode Applicability:5 -Cold Shutdown, 6 -Refueling, D -DefueledDefinition(s):NoneCPNPP Basis:Onsite/offsite communications include one or more of the systems listed in Table C-5 (ref. 1,2).This EAL is the cold condition equivalent of the hot condition EAL SU7.1.NEI 99-01 Basis:This IC addresses a significant loss of on-site or offsite communications capabilities. While nota direct challenge to plant or personnel safety, this event warrants prompt notifications toOROs and the NRC.Page 113 of 276 ATTACHMENT 1EAL BasesThis IC should be assessed only when extraordinary means are being utilized to makecommunications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent tooffsite locations, etc.).I EAL-#4The first EAL condition addresses a total loss of the communications methods used insupport of routine plant operations.EAL #2The second EAL condition addresses a total loss of the communications methods usedto notify all OROs of an emergency declaration. The offsite (OROs. referred to here are-(seeDevelGpe.-N.tee) the State Department of Public Safety, Somervell and Hood County EOCs7EAL4#3The third EAL addresses a total loss of the communications methods used to notify theNRC of an emergency declaration.CPNPP Basis Reference(s):1. FSAR 9.5.22. DBD-EE-048 Communication System3. NEI 99-01 CU5I Page 114 of 276 1 ATTACHMENT IEAL BasesCategory:Subcategory:Initiating Condition:C -Cold Shutdown / Refueling System Malfunction6 -Hazardous Event Affecting Safety SystemsHazardous event affecting a SAFETY SYSTEM needed for the currentoperating modeEAL:CA6.1 AlertThe occurrence of any Table C-6 hazardous eventAND EITHER:" Event damage has caused indications of degraded performance in at least one trainof a SAFETY SYSTEM needed for the current operating mode* The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component orstructure needed for the current operating modeTable C-6 Hazardous Events* Seismic event (earthquake)* Internal or external FLOODING event* High winds or tornado strike" FIRE* EXPLOSION" Other events with similar hazard characteristicsas determined by the Emergency CoordinatorMode Applicability:5 -Cold Shutdown, 6 -RefuelingDefinition(s):EXPLOSION -A rapid, violent and catastrophic failure of a piece of equipment due tocombustion, chemical reaction or overpressurization. A release of steam (from high energylines or components) or an electrical component failure (caused by short circuits, grounding,arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drivebelts or overheated electrical equipment do not constitute fires. Observation of flame ispreferred but is NOT required if large quantities of smoke and heat are observed.Page 115 of 276 ATTACHMENT 1EAL BasesFLOODING -A condition where water is entering a room or area faster than installedequipment is capable of removal, resulting in a rise of water level within the room or area.SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/orplacing it in the cold shutdown condition, including the ECCS. These are typically systemsclassified as safety-related (as defined in 10CFR50.2):Those structures, systems and components that are relied upon to remain functional duringand 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 could resultin potential offsite exposures.VISIBLE DAMAGE -Damage to a component or structure that is readily observable withoutmeasurements, testing, or analysis. The visual impact of the damage is sufficient to causeconcern regarding the operability or reliability of the affected component or structure.CPNPP Basis:" The significance of seismic events are discussed under EAL HU2.1 (ref. 1).* Internal FLOODING may be caused by events such as component failures, equipmentmisalignment, or outage activity mishaps (ref. 2).* External flooding may be due to high lake level (ref. 3, 4).* Seismic Category I structures are analyzed to withstand a sustained, design wind velocityof at least 80 mph. (ref. 5).* Areas containing functions and systems required for safe shutdown of the plant areidentified by fire area (ref. 6, 7).* An explosion that degrades the performance of a SAFETY SYSTEM train or visiblydamages a SAFETY SYSTEM component or structure would be classified under this EAL.NEI 99-01 Basis:This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or astructure containing SAFETY SYSTEM components, needed for the current operating mode.This condition significantly reduces the margin to a loss or potential loss of a fission productbarrier, and therefore represents an actual or potential substantial degradation of the level ofsafety of the plant.EAL 4.b.'The first conditional addresses damage to a SAFETY SYSTEM train that is inservice/operation since indications for it will be readily available. The indications of degradedperformance should be significant enough to cause concern regarding the operability orreliability of the SAFETY SYSTEM train.Page 116 of 276 ATTACHMENT 1EAL BasesEAL l1b.2The second conditional addresses damage to a SAFETY SYSTEM component thatis not in service/operation or readily apparent through indications alone, or to a structurecontaining SAFETY SYSTEM components. Operators will make this determination based onthe totality of available event and damage report information. This is intended to be a briefassessment not requiring lengthy analysis or quantification of the damage.Escalation of the emergency classification level would be via IC CS1 or A-S--RSI.CPNPP Basis Reference(s):1. ABN-907 Acts of Nature2. CPNPP PRA Accident Sequence Analysis "Internal Flooding Sequences"3. FSAR Section 2.4.3.7 Flood Evaluations for Safe Shutdown Impoundment4. DBD-CS-071 Maximum Probable Flood5. FSAR Section 3.3.1.1 Wind Loadings6. CPNPP Fire Protection Report, Section 5.0 "Fire Safe Shutdown Equipment List"7. FSAR Section 7.4 Systems Required for Safe Shutdown8. NEI 99-01 CA6Page 117 of 276 ATTACHMENT 1EAL BasesCategory H -Hazards and Other Conditions Affecting Plant SafetyEAL Group: ANY (EALs in this category are applicable to any plantcondition, hot or cold.)Hazards are non-plant, system-related events that can directly or indirectly affect plantoperation, reactor plant safety or personnel safety.1. SecurityUnauthorized entry attempts into the Protected Area, bomb threats, sabotage attempts, andactual security compromises threatening loss of physical control of the plant.2. Seismic EventNatural events such as earthquakes have potential to cause plant structure or equipmentdamage of sufficient magnitude to threaten personnel or plant safety.3. Natural or Technolo-gy HazardOther natural and non-naturally occurring events that can cause damage to plant facilitiesinclude tornados, FLOODING, hazardous material releases and events restricting siteaccess warranting classification.4. FireFires can pose significant hazards to personnel and reactor safety. Appropriate forclassification are fires within the site Protected Area or which may affect operability ofequipment needed for safe shutdown5. Hazardous GasToxic, corrosive, asphyxiant or flammable gas leaks can affect normal plant operations orpreclude access to plant areas required to safely shutdown the plant.6. Control Room EvacuationEvents that are indicative of loss of Control Room habitability. If the Control Room must beevacuated, additional support for monitoring and controlling plant functions is necessarythrough the emergency response facilities.Page 118 of 276 1 ATTACHMENT IEAL Bases7. Emeraencv Coordinator JudQmentThe EALs defined in other categories specify the predetermined symptoms or events thatare 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 emergencies basedon operator/management experience and judgment is still necessary. The EALs of thiscategory provide the Emergency Coordinator the latitude to classify emergency conditionsconsistent with the established classification criteria based upon Emergency Coordinatorjudgment.Page 119 of 276 ATTACHMENT 1EAL BasesCategory: H -HazardsSubcategory: 1 -SecurityInitiating Condition: Confirmed SECURITY CONDITION or threatEAL:HUI.1 Unusual EventA SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by theSecurity Shift SupervisorMode Applicability:AllDefinition(s):SECURITY CONDITION -Any security event as listed in the approved security contingencyplan that constitutes a threat/compromise to site security, threat/risk to site personnel, or apotential degradation to the level of safety of the plant. A security condition does not involve ahostile action.HOSTILE ACTION -An act toward CPNPP or its personnel that includes the use of violentforce to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end.This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or otherdevices used to deliver destructive force. Other acts that satisfy the overall intent may beincluded. Hostile action should not be construed to include acts of civil disobedience orfelonious acts that are not part of a concerted attack on CPNPP. Non-terrorism-based EALsshould be used to address such activities (i.e., this may include violent acts betweenindividuals in the owner controlled area).CPNPP Basis:The security shift supervision is defined as the Security Shift Supervisor.This EAL is based on the CPNPP Safeguards Contingency Plan (ref. 1).NEI 99-01 Basis:This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEMequipment, and thus represent a potential degradation in the level of plant safety. Securityevents which do not meet one of these EALs are adequately addressed by the requirements of10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS areclassifiable under ICs HA1, HS1 and HGI.Timely and accurate communications between Security Shift Supervision and the ControlRoom is essential for proper classification of a security-related event. Classification of theseevents will initiate appropriate threat-related notifications to plant personnel and OffsiteResponse Orqanizations.Security plans and terminology are based on the guidance provided by NEI 03-12, Templatefor the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [aP4ndoondPnt Spent Fuel Storage Installation Se276iy P-rogram.Page 120 of 276 ATTACHMENT 1EAL Bases&AL--#-This EAL references (site speGifiG-the seGU-ity-_shift Shift SecuritySupervisor because these are the individuals trained to confirm that a security event isoccurring or has occurred. Training on security event confirmation and classification iscontrolled due to the nature of Safeguards and 10 CFR § 2.39 information.Emergency plans and implementing procedures are public documents; therefore, EALs shouldnot incorporate Security-sensitive information. This includes information that may beadvantageous to a potential adversary, such as the particulars concerning a specific threat orthreat location. Security-sensitive information should be contained in non-public documentssuch as the CPNPP Safeguards Contingency Plan (ref. 1).Escalation of the emergency classification level would be via IC HAl.CPNPP Basis Reference(s):1. CPNPP Safeguards Contingency Plan (Safeguards)2. NEI 99-01 HU1Page 121 of 276 i ATTACHMENT 1EAL BasesCategory: H -HazardsSubcategory: 1 -SecurityInitiating Condition: Confirmed SECURITY CONDITION or threatEAL:HUI.2 Unusual EventNotification of a credible security threat directed at the siteMode Applicability:AllDefinition(s):SECURITY CONDITION -Any security event as listed in the approved security contingencyplan that constitutes a threat/compromise to site security, threat/risk to site personnel, or apotential degradation to the level of safety of the plant. A security condition does not involve ahostile action.HOSTILE ACTION- An act toward CPNPP or its personnel that includes the use of violentforce to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end.This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or otherdevices used to deliver destructive force. Other acts that satisfy the overall intent may beincluded. Hostile action should not be construed to include acts of civil disobedience orfelonious acts that are not part of a concerted attack on CPNPP. Non-terrorism-based EALsshould be used to address such activities (i.e., this may include violent acts betweenindividuals in the owner controlled area).CPNPP Basis:The security shift supervision is defined as the Security Shift Supervisor.This EAL is based on the CPNPP Safeguards Contingency Plan (ref. 1).NEI 99-01 Basis:This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEMequipment, and thus represent a potential degradation in the level of plant safety. Securityevents which do not meet one of these EALs are adequately addressed by the requirements of10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS areclassifiable under ICs HA1, HS1 and HGI.Timely and accurate communications between Security Shift Supervision and the ControlRoom is essential for proper classification of a security-related event. Classification of theseevents will initiate appropriate threat-related notifications to plant personnel and OffsiteResponse Orqanizations.Security plans and terminology are based on the guidance provided by NEI 03-12, Templatefor the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [aRdndpcndent Spent Fuel Storage installation Seur.ty Po.gra., .EAL #1 references (sitespecific security Shift supeiion) because these are the i d trained to cnR thatsecurity event is ocurner has occGurred. Trainin on security event confirmation andPage 122 of 276 ATTACHMENT 1EAL Basesis cRntrlled due to the nature of Saf.guards and 10 CIFR § 2.39 EAL #2 This EAL addresses the receipt of a credible security threat. The credibility of thethreat is assessed in accordance with (site-specific procedure).EA6L #3 addresses the threat from thoe of an aircraft on the plant. The NRI mIeadguartors Operations Officer (HOO) Will commUnicate to the licensee if the threat innvoVesan aircraft. The status and size of thc plane mnay also be provided by NORA.D through theNRC. Valid-ation of the threat is performed in accordance with (site specific proceduro).Emergency plans and implementing procedures are public documents; therefore, EALs shouldnot incorporate Security-sensitive information. This includes information that may beadvantageous to a potential adversary, such as the particulars concerning a specific threat orthreat location. Security-sensitive information should be contained in non-public documentssuch as the CPNPP Safequards Contingrency Plan (ref. 1).Escalation of the emergency classification level would be via IC HA1.CPNPP Basis Reference(s):1. CPNPP Safeguards Contingency Plan (Safeguards)2. NEI 99-01 HU1IPage 123 of 276 ATTACHMENT 1EAL BasesCategory: H -HazardsSubcategory: I -SecurityInitiating Condition: Confirmed SECURITY CONDITION or threatEAL:HUl.3 Unusual EventA validated notification from the NRC providing information of an aircraft threatMode Applicability:AllDefinition(s):NoneCPNPP Basis:This EAL is based on the CPNPP Safeguards Contingency Plan (ref. 1).NEI 99-01 Basis:This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEMequipment, and thus represent a potential degradation in the level of plant safety. Securityevents which do not meet one of these EALs are adequately addressed by the requirements of10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS areclassifiable under ICs HA1, HS1 and HGI.Timely and accurate communications between Security Shift Supervision and the ControlRoom is essential for proper classification of a security-related event. Classification of theseevents will initiate appropriate threat-related notifications to plant personnel and OffsiteResponse Orqanizations.Security plans and terminology are based on the guidance provided by NEI 03-12, Templatefor the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [a&REAL-#3This EAL addresses the threat from the impact of an aircraft on the plant. The NRCHeadquarters Operations Officer (HOO) will communicate to the licensee if the threat involvesan aircraft. The status and size of the plane may also be provided by NORAD through theNRC. Validation of the threat is performed in accordance with (site-specific procedure).Emergency plans and -implementing procedures are public documents; therefore, EALs shouldnot incorporate Security-sensitive information. This includes information that may beadvantageous to a potential adversary, such as the particulars concerning a specific threat orthreat location. Security-sensitive information should be contained in non-public documentssuch as the CPNPP Safeguards Continqency Plan (ref. 1).Escalation of the emergency classification level would be via IC HAl.CPNPP Basis Reference(s):1. CPNPP Safeguards Contingency Plan (Safeguards)Page 124 of 276 ATTACHMENT 1EAL Bases2. NEI 99-01 HU1Category: H -HazardsSubcategory: 1 -SecurityInitiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA orairborne attack threat within 30 minutesEAL:HAI.1 AlertA HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLEDAREA as reported by the Security Shift SupervisorMode Applicability:AllDefinition(s):HOSTILE ACTION -An act toward CPNPP or its personnel that includes the use of violentforce to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end.This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or otherdevices used to deliver destructive force. Other acts that satisfy the overall intent may beincluded. Hostile action should not be construed to include acts of civil disobedience orfelonious acts that are not part of a concerted attack on CPNPP. Non-terrorism-based EALsshould be used to address such activities (i.e., this may include violent acts betweenindividuals in the owner controlled area).OWNER CONTROLLED AREA -As shown in CPNPP Emergency Plan Appendix E, Complexand Owner Controlled Area.CPNPP Basis:The security shift supervision is defined as the Security Shift Supervisor.NEI 99-01 Basis:This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLEDAREA or notification of an aircraft attack thrcFat. This event will require rapid response andassistance due to the possibility of the attack progressing to the PROTECTED AREA, or theneed to prepare the plant and staff for a potential aircraft impact.Timely and accurate communications between the Security Shift uperiAsioR -Supervisor andthe Control Room is essential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Templatefor the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan-[aiwdIndopondont Spent Fu~el Storage Installation Se G1rFtY P-ro grFam.As time and conditions allow, these events require a heightened state of readiness by the plantstaff and implementation of onsite protective measures (e.g., evacuation, dispersal orsheltering). The Alert declaration will also heighten the awareness of Offsite ResponseI Organizations (OROs), allowing them to be better prepared should it be necessary to considerfurther actions.Page 125 of 276 ATTACHMENT 1EAL BasesThis IC does not apply to incidents that are accidental events, acts of civil disobedience, orotherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples includethe crash of a small aircraft, shots from hunters, physical disputes between employees, etc.Reporting of these types of events is adequately addressed by other EALs, or therequirements of 10 CFR § 73.71 or 10 CFR § 50.72.EAL-#!This EAL is applicable for any HOSTILE ACTION occurring, or that has occurred, in theOWNER CONTROLLED AREA. This includes any action directed against an ISFSI that islocated outside the plant PROTECTED AREA.EAL #taddresses the threat f9Rom the imnpact of an aircraft on the plant, and the anticipatednotific-';ations arc made in a timely mRannr so that plant personnRel and OROsr are inheightened state of readireSS. This EAL is met when the threat related iRnformation has beenva-lid,-afted on aG,,,,G l-Jn,'aR w^ith (site fifa The NRC Headquarters Operations Officer (HO00) Will communicate to the licensee if thethreat involves an aircraft. The status and size of the plane may be provided by NORAIDthr.ugh the NRC.Inn Sme cases, it t be readily apparent if an aircraft impat within the OWNERCONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). it is expected, although notcertain, that notification A ' by an appropriate Federal agency to the site would .larif,' this point.÷ Inthis ease, the appropriate federal agency is intended to be NORAID, FBI, FA. or NRC. Theemnergency declar-ation, including onRe -b-ased on other l~siFEALs, should not be unduly delayewhile awaiting notification by a Federal agcncy.Emergency plans and implementingprocedures are public documents; therefore, EALs should not incorporate Security-sensitiveinformation. This includes information that may be advantageous to a potential adversary,such as the particulars concerning a specific threat or threat location. Security-sensitiveinformation should be contained in non-public documents such as the CPNPP SafeguardsContingiency Plan (ref. 1).CPNPP Basis Reference(s):1. CPNPP Safeguards Contingency Plan (Safeguards)2. NEI 99-01 HA1IPage 126 of 276 ATTACHMENT 1EAL BasesCategory: H -HazardsSubcategory: 1 -SecurityInitiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA orairborne attack threat within 30 minutesEAL:HA1.2 AlertA validated notification from NRC of an aircraft attack threat within 30 min. of the siteMode Applicability:AllDefinition(s):NoneCPNPP Basis:NoneNEI 99-01 Basis:This IC addresses the occurrence of a HOSTILE ACTION withiR the OWNER CONTROLLEDAREA or notification of an aircraft attack threat. This event will require rapid response andassistance due to the possibility of the attack progressing to the PROTECTED AREA, or theneed to prepare the plant and staff for a potential aircraft impact.I Timely and accurate communications between the Security Shift Su pe iSieR-Supervisor andthe Control Room is essential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Templatefor the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan-[andIndependent Spent Fue! Storage Installation SocuritY P-rogram]ý1.As time and conditions allow, these events require a heightened state of readiness by the plantstaff and implementation of onsite protective measures (e.g., evacuation, dispersal orsheltering). The Alert declaration will also heighten the awareness of Offsite ResponseOrganizations (OROs), allowing them to be better prepared should it be necessary to considerfurther actions.This IC does not apply to incidents that are accidental events, acts of civil disobedience, orotherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples includethe crash of a small aircraft, shots from hunters, physical disputes between employees, etc.Reporting of these types of events is adequately addressed by other EALs, or therequirements of 10 CFR § 73.71 or 10 CFR § 50.72.EAL #1 is applicable for any H"OSTILn EACTPI"ION oc-curring, or that has -occured, in the OWNEA R N E D AR EA This inql- cludepan',' action against an ISSI- that is loc-Gated outside the plant PROTECTED [AREAEAL-#2This EAL addresses the threat from the impact of an aircraft on the plant, and theanticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in aPage 127 of 276 ATTACHMENT 1EAL Basesheightened state of readiness. This EAL is met when the threat-related information has beenvalidated in accordance with (site-specific security procedures).The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if thethreat involves an aircraft. The status and size of the plane may be provided by NORADthrough the NRC.In some cases, it may not be readily apparent if an aircraft impact within the OWNERCONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although notcertain, that notification by an appropriate Federal agency to the site would clarify this point. Inthis case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. Theemergency declaration, including one based on other ICs/EALs, should not be unduly delayedwhile awaiting notification by a Federal agency.Emergency plans and implementing procedures are public documents; therefore, EALs shouldnot incorporate Security-sensitive information. This includes information that may beadvantageous to a potential adversary, such as the particulars concerning a specific threat orthreat location. Security-sensitive information should be contained in non-public documentssuch as the CPNPP Safeguards Contingency Plan (ref. 1).CPNPP Basis Reference(s):1. CPNPP Safeguards Contingency Plan (Safeguards)2. NEI 99-01 HAIPage 128 of 276 ATTACHMENT 1EAL BasesCategory: H -HazardsSubcategory: 1 -SecurityInitiating Condition: HOSTILE ACTION within the PROTECTED AREAEAL:HSI.1 Site Area EmergencyA HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA asreported by the Security Shift SupervisorMode Applicability:AllDefinition(s):HOSTILE ACTION -,An act toward CPNPP or its personnel that includes the use of violentforce to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end.This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or otherdevices used to deliver destructive force. Other acts that satisfy the overall intent may beincluded. Hostile action should not be construed to include acts of civil disobedience orfelonious acts that are not part of a concerted attack on CPNPP. Non-terrorism-based EALsshould be used to address such activities (i.e., this may include violent acts betweenindividuals in the owner controlled area).PROTECTED AREA -An area encompassed by physical barriers and to which access iscontrolled. The Protected Area refers to the designated security area around the processbuildings and is depicted in FSAR Figure 1.2-1 Plot Plan.CPNPP Basis:The security shift supervision is defined as the Security Shift Supervisor.These individuals are the designated on-site personnel qualified and trained to confirm that asecurity event is occurring or has occurred. Training on security event classificationconfirmation is closely controlled due to the strict secrecy controls placed on the CPNPPSafeguards Contingency Plan (Safeguards) information. (ref. 1)NEI 99-01 Basis:This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA.This event will require rapid response and assistance due to the possibility for damage to plantequipment.Timely and accurate communications between Security Shift Supervision and the ControlRoom is essential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Templatefor the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [a&RIndependent Spent Fue! Storage. n'stallatioAn Security Pregram].As time and conditions allow, these events require a heightened state of readiness by the plantstaff and implementation of onsite protective measures (e.g., evacuation, dispersal orPage 129 of 276 ATTACHMENT 1EAL Basessheltering). The Site Area Emergency declaration will mobilize Offsite Response Organization(ORO) resources and have them available to develop and implement public protective actionsin the unlikely event that the attack is successful in impairing multiple safety functions.This IC does not apply to a HOSTILE ACTION directed at an ISFSI PROTECTED AREAlocated outside the plant PROTECTED AREA; such an attack should be assessed using ICHAl. It also does not apply to incidents that are accidental events, acts of civil disobedience,or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examplesinclude the crash of a small aircraft, shots from hunters, physical disputes between employees,etc. Reporting of these types of events is adequately addressed by other EALs, or therequirements of 10 CFR § 73.71 or 10 CFR § 50.72.Emergency plans and implementing procedures are public documents; therefore, EALs shouldnot incorporate Security-sensitive information. This includes information that may beadvantageous to a potential adversary, such as the particulars concerning a specific threat orthreat location. Security-sensitive information should be contained in non-public documentssuch as the CPNPP Safeguards Contingency Plan (ref. 1).Escalation of the emergency classification level would be via IC HG1.CPNPP Basis Reference(s):1. CPNPP Safeguards Contingency Plan (Safeguards)2. NEI 99-01 HS1Page 130 of 276 ATTACHMENT 1EAL BasesCategory:H -HazardsSubcategory: 1 -SecurityInitiating Condition: HOSTILE ACTION resulting in loss of physical control of the facilityEAL:HGI.1 General EmergencyA HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA asreported by the Security Shift SupervisorAND EITHER of the following has occurred:" One or more of the following safety functions cannot be controlled or maintained-Reactivity control-Core cooling-RCS heat removalOR" Damage to spent fuel has occurred or is IMMINENTMode Applicability:AllDefinition(s):HOSTILE ACTION -An act toward CPNPP or its personnel that includes the use of violentforce to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end.This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or otherdevices used to deliver destructive force. Other acts that satisfy the overall intent may beincluded. Hostile action should not be construed to include acts of civil disobedience orfelonious acts that are not part of a concerted attack on CPNPP. Non-terrorism-based EALsshould be used to address such activities (i.e., this may include violent acts betweenindividuals in the owner controlled area).IMMINENT -The trajectory of events or conditions is such that an EAL will be met within arelatively short period of time regardless of mitigation or corrective actionsPROTECTED AREA -An area encompassed by physical barriers and to which access iscontrolled. The Protected Area refers to the designated security area around the processbuildings and is depicted in FSAR Figure 1.2-1 Plot Plan.CPNPP Basis:The security shift supervision is defined as the Security Shift Supervisor.NEI 99-01 Basis:This IC addresses an event in which a HOSTILE FORCE has taken physical control of thefacility to the extent that the plant staff can no longer operate equipment necessary to maintainkey safety functions. It also addresses a HOSTILE ACTION leading to a loss of physicalcontrol that results in actual or IMMINENT damage to spent fuel due to 1) damage to a spentPage 131 of 276 I ATTACHMENT 1EAL Basesfuel pool cooling system (e.g., pumps, heat exchangers, controls, etc.) or, 2) loss of spent fuelpool integrity such that sufficient water level cannot be maintained.Timely and accurate communications between Security Shift Supervision and the ControlRoom is essential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Templatefor the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [andIndependent Spent Fu4el Storage Installation Socbrity P-rogram].Emergency plans and implementing procedures are public documents; therefore, EALs shouldnot incorporate Security-sensitive information. This includes information that may beadvantageous to a potential adversary, such as the particulars concerning a specific threat orthreat location. Security-sensitive information should be contained in non-public documentssuch as the CPNPP Safeguards Contingency Plan (ref.1).CPNPP Basis Reference(s):1. CPNPP Safeguards Contingency Plan (Safeguards)2. NEI 99-01 HG1Page 132 of 276 ATTACHMENT 1EAL BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 2 -Seismic EventInitiating Condition: Seismic event greater than OBE levelEAL:HU2.1 Unusual EventSeismic event greater than OBE as indicated by annunciator 2A-3.1, OBE EXCEEDED, oryellow OBE light on Seismic Monitoring system panelMode Applicability:AllDefinition(s):NoneCPNPP Basis:Seismic events of this magnitude can result in a areas needed for safe shutdown beingsubjected to forces beyond design limits, and thus damage may be assumed to have occurredto plant safety systems.A conservative Safe Shutdown Earthquake (SSE) having a peak horizontal groundacceleration at the top of bedrock of 0.12 g has been selected for design (FSAR Section2.5.2.6). The Operating Basis Earthquake (OBE) is equal to 1/2 the SSE (ref. 1).When the seismic recorder indicates that the OBE has been exceeded, System Engineeringmust evaluate and determine whether the reactor must be shut down and remain shutdownuntil inspection of the facility shows that no damage has been incurred which would jeopardizesafe operation of the facility or until such damage is repaired. CPNPP was designed such that,for ground motion less than the OBE, the features of the plant necessary for continuedoperation without undue risk to the health and safety of the public will remain functional. Anyground motion in excess of this results in an uncertainty as to the extent of the damage whichmust be resolved before continued operation can be considered safe (ref. 2).The seismic trigger, CP1-SIATAS-03, is set to activate the strong motion recording system atan acceleration level slightly above normal ambient vibrations (0.01g) and well below thepostulated OBE "free field" ground acceleration (0.06g horizontal). This causes an alarm in thecontrol room to alert the operator. (ref. 2, 3) The seismic recorders (strong motionaccelerators) monitor earth vibration and, when triggered, store data in the recorder. TriaxialSMAs are installed at appropriate locations to provide data on the frequency, amplitude, andphase relationship of the seismic response of the containment structure and the seismic inputto other seismic Category I structures, systems, and components. The Seismic Instrumentationconsists of strong motion accelerograph (triaxial time history accelerograph system), triaxialpeak accelerograph recorders, passive response spectrum recorders, a response spectrumswitch, and a seismic switch. Except for sensors for the active instrumentation, all electronicsfor processing and storage of the seismic data are located in the seismic instrumentation panelCPX-ECPRCV-1 1 in the control room. There is no additional seismic instrumentation requiredfor Unit 2. However, alarms from seismic instrumentation in Unit 1 are duplicated in Unit 2. ThePage 133 of 276 ATTACHMENT 1EAL Basestime history accelerograph system is fully operational within 0.1 second after the seismictrigger is actuated.It will operate continuously during the period in which the earthquake exceeds the seismictrigger threshold (0.01g) plus 5 seconds (minimum) beyond the last seismic trigger signal.ABN-907 Acts of Nature provides the guidance for determining if the OBE earthquakethreshold is exceeded and any required response actions. (ref. 2)To avoid inappropriate emergency classification resulting from spurious actuation of theseismic instrumentation or felt motion not attributable to seismic activity, an offsite agency(USGS, National Earthquake Information Center) can confirm that an earthquake has occurredin the area of the plant. Such confirmation should not, however, preclude a timely emergencydeclaration based on receipt of the OBE alarm. The NEIC can be contacted by calling (303)273-8500. Select option #1 and inform the analyst you wish to confirm recent seismic activityin the vicinity of CPNPP. Alternatively, near real-time seismic activity can be accessed via theNEIC website:http://earthquake.usgs.gov/earthquakes/dyfi/archives.phpNEI 99-01 Basis:This IC addresses a seismic event that results in accelerations at the plant site greater thanthose specified for an Operating Basis Earthquake (OBE). An earthquake greater than anOBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact onsafety-related systems, structures and components; however, some time may be required forthe plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections). Given the time necessary to perform walk-downs andinspections, and fully understand any impacts, this event represents a potential degradation ofthe level of safety of the plant.Event verification with external sources should not be necessary during or following an OBE.Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as aseismic event (e.g., lateral accelerations in excess of 0.08g). The Shift Manager orEmergency IDEeeter-Coordinator may seek external verification if deemed appropriate (e.g., acall to the USGS, check internet news sources, etc.); however, the verification action must notpreclude a timely emergency declaration.Depending upon the plant mode at the time of the event, escalation of the emergencyclassification level would be via IC CA6 or SA9.CPNPP Basis Reference(s):1. FSAR Section 2.5.4.9 Earthquake Design Basis2. ABN-907 Acts of Nature3. DBD-EE-077 Seismic Instrumentation4. 1, 2-ALB-2A-3.1 OBE EXCEEDED5. DBD-ME-028 Classification of Structures, Systems and Components6. NEI 99-01 HU2Page 134 of 276 ATTACHMENT 1EAL BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 3 -Natural or Technology HazardInitiating Condition: Hazardous eventEAL:HU3.1 Unusual EventA tornado strike within the PROTECTED AREAMode Applicability:AllDefinition(s):PROTECTED AREA -An area encompassed by physical barriers and to which access iscontrolled. The Protected Area refers to the designated security area around the processbuildings and is depicted in FSAR Figure 1.2-1 Plot Plan.CPNPP Basis:Response actions associated with a tornado onsite is provided in ABN-907 Acts of Nature (ref.1).If damage is confirmed visually or by other in-plant indications, the event may be escalated toan Alert under EAL CA6.1 or SA9.1.A tornado striking (touching down) within the PROTECTED AREA warrants declaration of anUnusual Event regardless of the measured wind speed at the meteorological tower. A tornadois defined as a violently rotating column of air in contact with the ground and extending fromthe base of a thunderstorm.NEI 99-01 Basis:This IC addresses hazardous events that are considered to represent a potential degradationof the level of safety of the plant.I EAL #1EAL HU3.1 addresses a tornado striking (touching down) within the PROTECTEDAREA.EAL #2 addresses flooding of a building roomn or area that results in operators isolating powcrto a SAFETY SYSTEM compnenRGt du-e to wgater level oer o-ther wetting cARonerns. Cla~ssificloatio~n.is also rcguircd if the water level or rclated Wetting causcs an automatic isolation of a SAFEPSYSTEM comnponent from its power source (e.g., a breaker or relay trip). To9 warrnclassification, operability of the affected component must be reguircd by TcehnicalSpecifications for the current operating moede.-EAL #3 addresses a hadosmtrlsevent originating at an eff-sitc loc-ation and ofsufiietmagnitude to impede the moevement Of peF8rRonnl Within t-hc PROTEC=GTEFD AREA.EAL #4 addrcsscs a hazardous event that c~auses an on site impedimcnet to Vehicle moVmentand Significant enough topo ibthe plant staff from accesSing the site usngprsoalvehcis.Exape ofsuch. an. event include site flooding caused by -a huMrricamne, hcaVY rains,Page 135 of 276 ATTACHMENT 1EAL Basesup rier wter releases, damn failure, etc., Or an on site train derailment blocking the accessThis EAL is inteRded apply to routine impediments such as fog, snow, i.e, or Vehi-lebreakdowns or accidents, but rather to mrnee significant conditions such as the HurricaneAndrew strike on Turkey Point in 1992, the flooding around the Cooper Station during theMidw~est floods of 1993, or the flooding aroun~d Ft. Calhoun Station in 2011.EAL #5 addresses (site specific description).Escalation of the emergency classification level would be based on ICs in RecognitionCategories AR, F, S or C.CPNPP Basis Reference(s):1. ABN-907 Acts of Nature2. NEI 99-01 HU3I Page 136 of 276 ATTACHMENT 1EAL BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 3 -Natural or Technology HazardInitiating Condition: Hazardous eventEAL:HU3.2 Unusual EventInternal room or area FLOODING of a magnitude sufficient to require manual or automaticelectrical isolation of a SAFETY SYSTEM component needed for the current operatingmodeMode Applicability:AllDefinition(s):FLOODING -A condition where water is entering a room or area faster than installedequipment is capable of removal, resulting in a rise of water level within the room or area.SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/orplacing it in the cold shutdown condition, including the ECCS. These are typically systemsclassified as safety-related (as defined in 10CFR50.2):Those structures, systems and components that are relied upon to remain functional duringand 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 could resultin potential offsite exposures.CPNPP Basis:The internal flooding areas of concern are the Safeguards Building and Turbine Building(ref.1). Refer to EAL CA6.1 for internal flooding affecting one or more SAFETY SYSTEMtrains.NEI 99-01 Basis:This IC addresses hazardous events that are considered to represent a potential degradationof the level of safety of the plant.EAL #1 addresses a tornado striking (touching doWn) Within tho_ P2ROTECQTED ARE=A.This EAL addresses FLOODING of a building room or area that results in operators isolatingpower to a SAFETY SYSTEM component due to water level or other wetting concerns.Classification is also required if the water level or related wetting causes an automatic isolationof a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). Towarrant classification, operability of the affected component must be required by TechnicalSpecifications for the current operating mode.Page 137 of 276 ATTACHMENT 1EAL BasesEAL #3 address-sc-es aa materials evPent rigiRnating at an f-fsite lcatio;n and ofSUfficieRt magnitude to impede the movement of pcrSOnncl Within thc PROTECTED rAREA.AE:AI #1 addresses a h.zArdou e.....vent that c.au.ses an onsite impediment to vehicle movementand significant cnough to prohibit the plant staff fro~m acc-pessing the site using personRalvehicles. Examples oif such an event include site flooding caused by a heavy rains&,Iup rier wter releases, dam failure, etc., or an on site train derailment blocking the accessThis EAL intended apply to routine impediments .uch as fog, snow, iGe, or vehiclebhre-akdoAwns or accidents, but rather to imopree significant conditions such As6 the- HurricaneAndrew strike on Turkey Point in 1992, the floodin;g around the Cooper Station during theMidwest flo-dis o-f 1Q993, or the flooding around Ft. C-alhoun61 Station in 20-11.EAL #5 addresses (site specific description).Escalation of the emergency classification level would be based on ICs in Recognition Categories AR,F, SorC.CPNPP Basis Reference(s):1. CPNPP PRA Accident Sequence Analysis "Internal Flooding Sequences"2. NEI 99-01 HU3Page 138 of 276 ATTACHMENT 1EAL BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 3 -Natural or Technology HazardInitiating Condition: Hazardous eventEAL:HU3.3 Unusual EventMovement of personnel within the PROTECTED AREA is IMPEDED due to an offsiteevent involving hazardous materials (e.g., an offsite chemical spill or toxic gas release)Mode Applicability:AllDefinition(s):IMPEDE(D) -Personnel access to a room or area is hindered to an extent that extraordinarymeasures are necessary to facilitate entry of personnel into the affected room/area(e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).PROTECTED AREA -An area encompassed by physical barriers and to which access iscontrolled. The Protected Area refers to the designated security area around the processbuildings and is depicted in FSAR Figure 1.2-1 Plot Plan.CPNPP Basis:As used here, the term "offsite" is meant to be areas external to the CPNPP PROTECTEDAREA.NEI 99-01 Basis:This IC addresses hazardous events that are considered to represent a potential degradationof the level of safety of the plant.EAL 11 addresses a terade stFriking (touching dwnr) within the PROTECTED AREA.This EA'l addresses flooding of a buildi'g room or area that .esults in operators isolatirgpower to a SAFETY SYSTEM co4r-mponont due to water level or other wetting conceFrn.Classificatioen is also required if the water level or related wetting causes an automatic isolatioof a SAFETY SYSTEM compoenet froM its power sourcGe (o.g., a breaker or relay trip). ToGwarrant classification, operability of the affected component must be required by Technicalpecifiations for the Gcurent operating mode.EAL-#3This EAL addresses a hazardous materials event originating at an offsite location andof sufficient magnitude to impede the movement of personnel within the PROTECTED AREA.EAL #4 addresses a hazardous event that c~auses an on site impediment to vehicle movemenand significant enough to prohibit the plant staff from accessing the Site using personalvehicloes. Examples Of such an event include site flooding caused by a hurricane, heav' rains&,up rior wter releases, dam failure, etc., or an on site train derailment blocking the accessPage 139 of 276 ATTACHMENT 1EAL BasesThis EAL is notntended apply to routineim;pediments 6uch as fog, snow, ice, or vehiclebre~akdoGwns or accidcnts, but rather to) mor)e Significant conditionRs such as the HurcnAndrew strike.on Turkey Point in 1992, the flooding around the .r. Qtati-"n during theMidwest floods of 1993, Or the flooding aroun~d Ft. Calhoun Station in 2011.E=AL #f5 addresses (site specific description).Escalation of the emergency classification level would be based on ICs in RecognitionCategories AR, F, S or C.CPNPP Basis Reference(s):1. NEI 99-01 HU3Page 140 of 276 ATTACHMENT 1EAL BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 3 -Natural or Technology HazardInitiating Condition: Hazardous eventEAL:HU3.4 Unusual EventA hazardous event that results in on-site conditions sufficient to prohibit the plant staff fromaccessing the site via personal vehicles (Note 7)Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns oraccidents.Mode Applicability:AllDefinition(s):NoneCPNPP Basis:NoneNEI 99-01 Basis:This IC addresses hazardous events that are considered to represent a potential degradationof the level of safety of the plant.EAL #1 -,,e,4 a t .... ,,euh. @ down) ,i;hi14the PROTECTED AREA.This EAL addresses flooding of a building reem o area that results in operatorsi ipovee to a SAFETY SYSTEM comproent due to water level or other wetting concerns.lassificatine is also required if the water level or related wetting causes an automatic isolationef a SAFETY SYSTEM comrponent fom its powf r seouce (e.g., a breaker or relay trip). Towarrant classification, operability of the affected compoenet must be required by TechnicalSpecfcations for the current operating mode.EAL #3 addresses a hazardous materials event originating at an offsite location and ofsufficient magnitude to impede the mo'erncent of personnel w~ithin the PROTECTED ,AREA.ELAL44This EAL addresses a hazardous event that causes an on-site impediment to vehiclemovement and significant enough to prohibit the plant staff from accessing the site usingpersonal vehicles. Examples of such an event include site FLOODING caused by a hurricane,heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blockingthe access road.This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehiclebreakdowns or accidents, but rather to more significant conditions such as the HurricaneAndrew strike on Turkey Point in 1992, the flooding around the Cooper Station during theMidwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011 .EAL #5 addresses(site specific description)Page 141 of 276 ATTACHMENT 1EAL BasesEscalation of the emergency classification level would be based on ICs in RecognitionI Categories AR, F, S or C.CPNPP Basis Reference(s):1. NEI 99-01 HU3Page 142 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:H -Hazards and Other Conditions Affecting Plant Safety4- FireInitiating Condition: FIRE potentially degrading the level of safety of the plantEAL:HU4.1 Unusual EventA FIRE is not extinguished within 15 min. of any of the following FIRE detectionindications (Note 1):" Report from the field (i.e., visual observation)* Receipt of multiple (more than 1) fire alarms or indications" Field verification of a single fire alarmANDThe FIRE is located within any Table H-1 areaNote 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Table H-1 Fire Areas* u-Containment* u-Safeguards Building* X-Auxiliary Building* X-Electrical & Control Building* X-Fuel Building.X-Service Water Intake Structure* u-Diesel Generator Building* u-Normal Switchgear Rooms* u-CST* u-RWSTMode Applicability:AllDefinition(s):FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drivebelts or overheated electrical equipment do not constitute fires. Observation of flame ispreferred but is NOT required if large quantities of smoke and heat are observed.CPNPP Basis:The 15 minute requirement begins with a credible notification that a fire is occurring, or receiptof multiple valid fire detection system alarms or field validation of a single fire alarm. The alarmis to be validated using available Control Room indications or alarms to prove that it is notspurious, or by reports from the field. Actual field reports must be made within the 15 minutePage 143 of 276 ATTACHMENT 1EAL Basestime limit or a classification must be made. If a fire is verified to be occurring by field report, the15 minute time limit is from the original receipt of the fire detection alarm.Table H-1 applies to buildings and areas housing equipment needed for safe shutdown(SAFETY SYSTEMS) (ref. 1, 2).NEI 99-01 Basis:This IC addresses the magnitude and extent of FIRES that may be indicative of a potentialdegradation of the level of safety of the plant.EAL-#4-The-For EAL HU4.1 the intent of the 15-minute duration is to size the FIRE and to discriminateagainst small FIRES that are readily extinguished (e.g., smoldering waste paper basket). Inaddition to alarms, other indications of a FIRE could be a drop in fire main pressure, automaticactivation of a suppression system, etc.Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm,indication, or report. For EAL assessment purposes, the emergency declaration clock starts atthe time that the initial alarm, indication, or report was received, and not the time that asubsequent verification action was performed. Similarly, the fire duration clock also starts atthe time of receipt of the initial alarm, indication or report.EAL #2This EAL addresses rece.pt of a single fire alarm, and thc existence of a FPRE is not verified(i.e., proyed or disproved) within 30 minutes of the alarm. UJpon receipt, operators will takeprompt actions to confirm the validity of a single fire alarm. For E=AL assessment purposes, th~e30 Minute clock starts at the time that the initial -alar~m w~as received, and not the time that asubsequent verification action was performed.A single fire alarm, absent other iniainsef a FIRE, may be indicative of equipment failuror I .a.puius activation, and not an actual FIRE. Fer this reaseo, additional time is allowed toverif;, the validity of the alarm. The 30 minute period is a reasonable amount of time_ todetermiRe if an actual FIRE exists; however, after that time, and absent information to thecontrary, it is assumed that an actual FIRE is in progress.if an a.tual FIRE is verified by a report from the field, then EAL #11 is immediately applicable1and the emergency must be declared if the FIRE is not extinguished within 15-minutes of threporFt. if the alarm is verified to be due to an equipmnent failure or a Gpurious activation, andthis .occur within 30 minutes of the receipt of the alarm, then this EAL is netappicable aRd no emeFrgency declaratin is warranted.in addition to a FIRE addressed by, EAL #i or EAL #t2, a FIRE within the plant PROTECTEDAREA not extinguished within 60 minutes may also potentially degrade the level of plantsafety. This basis extends to a FIRE occurring within the PROTECTED AREA of an ISFSilocated outside the plant PROTEC-TED ,AREA. [Sentence for plants with an ISFSI outside th~eplant Protec-ted Area]P::AI if a-FIRE= within the o IOF I for,, plants with an ISF=I outside the plant Protected Area]PROTECiGTED AREA is of sufficient size to requie a re se by an offeite firefighting agencyPage 144 of 276 [ ATTACHMENT 1EAL Bases(e.g., a local tovn Fire Department.), then the lev~,el of plant safety is potentially degraded. Theis needed to actively SUPPort firefighting efforts because the fire i beyond the capability of thoFire BrFigade to extinguiSh. Declaration is not neesar if--- th 1 gnc resources arc placed 9Rstand by, or supporting post extinguishment reor;r investigation actiolns.Basis DRel+nA 0 ... ..r.m.. A .n...ixv 0Appendix R to 10 CFR 50, states in part:Criterion 3 of Appendix A to this part specifies that "Structures, systems, andcomFponents important to safety shall be designed and located to minimize, consistentwith other safety requirements, the probability and effect of fires adepoin.WheR considering the effects of fire, those systems associated with achieving aRdmaintaining safe shutdown conditions assume major importance to safety becausedamage to themB can lead to core damage resulting from loss of coolant through boil off.Because fire man" affect safe shutdown systems aRd because the loss of functifn Ofsystems used to mitigate the consequences of design basis accidents under postfrco~nditions does not per se impact public safety, the need to limit fire damage to systemsrequired to -achieve and mnaintain safe shutdown conditions is greater than the need to-limit fire damage to those systems required to mitigate the consequences Of desig.nbasis accidents.In addition, Appendix R to 10 .FR 50, requires, among other considerations, the use of 1 hourfire barriers for the enclosure of cable and equipment and associated non safety cirut of one-redundant train (G.2.G). As used in EAL #2, the 30 minutes to verify a single alarm is wellwithin this wor-st ease 1 hour time period.Depending upon the plant mode at the time of the event, escalation of the emergencyclassification level would be via IC CA6 or SA9.CPNPP Basis Reference(s):1. CPNPP Fire Protection Report, Section 5.0 "Fire Safe Shutdown Equipment List"2. FSAR Section 7.4 Systems Required for Safe Shutdown3. NEI 99-01 HU4Page 145 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:H -Hazards and Other Conditions Affecting Plant Safety4- FireInitiating Condition: FIRE potentially degrading the level of safety of the plantEAL:HU4.2 Unusual EventReceipt of a single fire alarm (i.e., no other indications of a FIRE)ANDThe fire alarm is indicating a FIRE within any Table H-1 areaANDThe existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Table H-1 Fire Areas* u-Containment* u-Safeguards Building* X-Auxiliary Building* X-Electrical & Control Building" X-Fuel Building* X-Service Water Intake Structure" u-Diesel Generator Building* u-Normal Switchgear Rooms" u-CST* u-RWSTMode Applicability:AllDefinition(s):FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drivebelts or overheated electrical equipment do not constitute fires. Observation of flame ispreferred but is NOT required if large quantities of smoke and heat are observed.CPNPP Basis:The 30 minute requirement begins upon receipt of a single valid fire detection system alarm.The alarm is to be validated using available Control Room indications or alarms to prove that itis not spurious, or by reports from the field. Actual field reports must be made within the 30minute time limit or a classification must be made. If a fire is verified to be occurring by fieldreport, classification shall be made based on EAL HU4.1.Page 146 of 276 ATTACHMENT 1EAL BasesTable H-1 applies to buildings and areas housing equipment needed for safe shutdown(SAFETY SYSTEMS) (ref. 1, 2).NEI 99-01 Basis:This IC addresses the magnitude and extent of FIRES that may be indicative of a potentialdegradation of the level of safety of the plant.E-AL-#-The intent of the 15 minute duration is to size the FIRE and to discrimninatc against smallFIRES that are readily extinguished (c.g., smoldering waste papcr basket). In addition toala~rms, other indle~ations of a FIRE could be a drop in fire mnain pressure, automatic activatioof a suppression system, etc.Upon receipt, operators will take prompt actionS to cofirm the validity of an initial fire alarm,indication, or report. For EAL assessment purposes, the emergency declaration clock starts at.the time that the initial alarm, indication, Or report was received, and not the time thasubsequent verification action was performed. Similarly, the fire duration clock alsoe starts at.the time of receipt of the initial alarm, indication or report.EAL-#2-This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified(i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will takeprompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the30-minute clock starts at the time that the initial alarm was received, and not the time that asubsequent verification action was performed.A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failureor a spurious activation, and not an actual FIRE. For this reason, additional time is allowed toverify the validity of the alarm. The 30-minute period is a reasonable amount of time todetermine if an actual FIRE exists; however, after that time, and absent information to thecontrary, it is assumed that an actual FIRE is in progress.IIf an actual FIRE is verified by a report from the field, then EAL #1 HU4.1 is immediatelyapplicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spuriousactivation, and this verification occurs within 30-minutes of the receipt of the alarm, then thisEAL is not applicable and no emergency declaration is warranted.EALI-#-3In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant PROTECTEDAREA not extinguished within 60- minutes may also potentially degrade the level of plantsafety. Thi' s basis extends to a FIRE= oGccurrg within the PROTECTEDý AREA of an !S1located outside the plant P-ROTECTED AREA4. [Senten-Ge for plants with an ISFSI outside theplant Pr-etected Area]EAI-#4Basis-Related Requirements from Appendix RAppendix R to 10 CFR 50, states in part:Page 147 of 276 ATTACHMENT 1EAL BasesCriterion 3 of Appendix A to this part specifies that "Structures, systems, andcomponents important to safety shall be designed and located to minimize, consistentwith other safety requirements, the probability and effect of fires and explosions."When considering the effects of fire, those systems associated with achieving andmaintaining safe shutdown conditions assume major importance to safety becausedamage to them can lead to core damage resulting from loss of coolant through boil-off.Because fire may affect safe shutdown systems and because the loss of function ofsystems used to mitigate the consequences of design basis accidents under post-fireconditions does not per se impact public safety, the need to limit fire damage to systemsrequired to achieve and maintain safe shutdown conditions is greater than the need tolimit fire damage to those systems required to mitigate the consequences of designbasis accidents.In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hourfire barriers for the enclosure of cable and equipment and associated non-safety circuits of oneredundant train (G.2.c). As used in EAL-#2HU4.2, the 30-minutes to verify a single alarm iswell within this worst-case 1-hour time period.Depending upon the plant mode at the time of the event, escalation of the emergencyclassification level would be via IC CA6 or SA9.CPNPP Basis Reference(s):1. CPNPP Fire Protection Report, Section 5.0 "Fire Safe Shutdown Equipment List"2. FSAR Section 7.4 Systems Required for Safe Shutdown3. NEI 99-01 HU4I Page 148 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:EAL:H -Hazards and Other Conditions Affecting Plant Safety4 -FireFIRE potentially degrading the level of safety of the plantHU4.3 Unusual EventA FIRE within the ISFSI or plant PROTECTED AREA not extinguished within 60 min. ofthe initial report, alarm or indication (Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Mode Applicability:AllDefinition(s):FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drivebelts or overheated electrical equipment do not constitute fires. Observation of flame ispreferred but is NOT required if large quantities of smoke and heat are observed.PROTECTED AREA -An area encompassed by physical barriers and to which access iscontrolled. The Protected Area refers to the designated security area around the processbuildings and is depicted in FSAR Figure 1.2-1 Plot Plan.CPNPP Basis:NoneNEI 99-01 Basis:This IC addresses the magnitude and extent of FIRES that may be indicative of a potentialdegradation of the level of safety of the plant.The intent of the 15 minute duration is to size the FIRE and to discrhiminate against smallFIRES that are readily (e.g., smoldering waste papcr basket). in addition toalarms, othc .indications of a FIRE could be a drop infir. main pr e..ure, automatic activationof a suppression system+, et+.UJpon receipt, oper-ators will take prompt actions to co4nfr the validity of an initial fire alarm,indicatiEn, or repeot. For EAL assessment purpoSes, the em.ergency declaration clock starts atthe time that the initial alarm, indicatin, Or report was rece.ived, and not the time that asubsequent verification action was perormed. Similarly, the fire duration clock also starts atthe time of receipt of the initial alarm, indication or report.This EA- addresses receipt of a single fire alarm, and the existenRce of FIRE is net Verified(i.e., proved Or disproved) within 30 of the alarm. Upon receipt, opeFators will takep.ropt to confirm the validity of a single fire alarm. For EAL assessment purposes,hPage 149 of 276 ATTACHMENT 1EAL Bases30 minute clock starts at the time that the initial! alarm was received, and not the time that asubsequent verfification action was performed.A single fire alarmn, absen-t o-t-her indication(s) of a F=IRE, may be indicative of equipment failror a spurio~us activation, and not an actuial FIRE. ForF this reason, additional time is allowed toverif' the validity of the alarm. The 30 minute period is a reasonable amount Of tim~e todetermnine if an actual F=IRE= exists; however, after that time, and absent informnation to thecontrar',, it is assumed that an a-tual IRE is in progress.If an actual FIRE is verified by a repod from the field, then .AL #1 is immediately applicable,and the emergency must be declared if the FIRE is not extinguished within 15 m-iutes ef therepodt. if the alarmn is verified to be due to an equipment failure or a spurious activation, and-this verification occurs Within 30 mFinutes of the receipt of the alarm, then this EAL= is notapplicable and no) emergency declaration is warranted.EAL-#3In addition to a FIRE addressed by EAL #4-HU4.1 or EAL #2HU4.2, a FIRE within the plantPROTECTED AREA not extinguished within 60-minutes may also potentially degrade the levelof plant safety. This basis extends to a FIRE occurring within the PROTECTED AREA of anISFSI located outside the plant PROTECTED AREA. [SentenGe foAplants with an !SFSIout-side the plaRt Pr-tected Ara]eAI"/Aif a FIRE within the [for plants with an ISFS outsido the Prote.etd Are.a]PROTECTED AREA is of sufficient size to require a nse by an offsite firefighting agency(e.g., a localI twn1 Fire Depadtment), then the level of plant safet is potentially degraded. Theis6 needed to actively suppedt firefighting effod-s because the fire is beyond the capability of theFire Brigade to extinguish. Declaration is not necessar,' if the agency resources are placed ostand by, or suppoding post- extinguishment recve ;orivestigation actions.Basis Related 0 Re.. {ir .. ...+ ,.,m A nn.R. ;,d .Appendix R to 10 CF=R 50, states in pad:-Criterion 3 of Appendix A to this padt specifies that "Structures, systems, andcomponents impodtant to safet shall be designed and located to minimize cnIstent4with other safety requirements, the probability and effect of fires and explio,-nos."When considering the effects of fire, those systemns associated with achieving andmaintaining safe shutdown conditionRs assumne mjo impedtance to safet becauseedamage to themR can lead to core damage resulting fromn loss of coolant through boil off.Beause fire may affect safe shutdown systems and because the loSs Of funcItion ofsystemns used to mitigate the consequences of design basis accidents under postfrconditions does not per se impact public safety, the need to limit fire damage to systemnsrequired to achieve and maintain safe shutdown conditions is greater than the need tolimit fire damage to those systems required to mitigate the consequences of designbasis accidents%.in addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1 hourfire barriers for the enclosure of cablean-d e-quipment and associated non safety iruits of. onePage 150 of 276 ATTACHMENT IEAL Basesred-undiant train (G.2.c). As used in EAL #2, the 30 minutes to verif' a single alarm is wellwithin this wor.st case 1 hour time .periOd.Depending upon the plant mode at the time of the event, escalation of the emergencyclassification level would be via IC CA6 or SA9.CPNPP Basis Reference(s):1. NEI 99-01 HU4Page 151 of 276 ATTACHMENT 1EAL BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 4- FireInitiating Condition: FIRE potentially degrading the level of safety of the plantEAL:HU4.4 Unusual EventA FIRE within the ISFSI or plant PROTECTED AREA that requires firefighting support byan offsite fire response agency to extinguishMode Applicability:AllDefinition(s):FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drivebelts or overheated electrical equipment do not constitute fires. Observation of flame ispreferred but is NOT required if large quantities of smoke and heat are observed.PROTECTED AREA -An area encompassed by physical barriers and to which access iscontrolled. The Protected Area refers to the designated security area around the processbuildings and is depicted in FSAR Figure 1.2-1 Plot Plan.CPNPP Basis:NoneNEI 99-01 Basis:This IC addresses the magnitude and extent of FIRES that may be indicative of a potentialdegradation of the level of safety of the plant.EIL-#!The intent of the 15 minute duration is to size the FIRE and to against smallFIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition toalarms, other indc;ations of a FIRE could be a drop in fire main pr essure, automatic activatioof a 6uppression system, etc.Upon receipt, operateos Will take prompt actoRns to- co'Rfirm the validity of an initial fire alarm,'indication, or report. For EAL assessment purposes, the emergency starts at.the time that the initial alarmn, indication, or report was rec~eived, and not the time that akn n,~, + ; fnln +; +; IfArrr Q; 1- InI1 +1,. F; Ar Cl, ,rn !n drr in -,c 4. + 4- -4I a Vtttl'J* VtaI ~*I .A ),lcthe time of receipt of the initial alarm, indication Or reporFt.EIAL-4t2This EAL addresses receipt of a single fire alarmF, and the existence of a FIRE is no~t Verifiedr rnnA rj r rlcý rrnxfpA\ iAcn~n2 oilhjtpC r ri 14c of +kc',nm nc. ranr +nf pnrp+r..rq %.At;! +nLp.promnpt actions to confirmn the validity of a single fire alarm. ForF E=AL assessmnent purposes,th30 _minuteý clock starts at the time that the initial alarm was received, and not the time that asubsequent verific-ation action Was pe~ore~fld.I Page 152 of 276 1 ATTACHMENT 1EAL BasesA single fire alarm, absent other indication(s) of a FIRE, mnay be indicativ~e of equipment failuror sprius activation, and not an actuai PRE~. tr-e 1IS reasGn, auuRIuuai umne is aluoweu toerif, the ,l;4idity of the alaFrm. The 30 minute peFred is a reasonable amou.nt of time todetermie if .an atual FIRE exists; however, after that time, and absent to theGontrar,', it is -assu~med that an actual FIRE is i rgesIf an actual FIRE is erified by a repoFt from the field, then EAL imm ediatel applicable,and- the emergency must be declredAp-- if the- FIRE is not extingui-shed- 1Aithin 15 m-inutes o~f thereport. if the alarm is verified to be due to an equipment failure or a spurious activation, and-this verification occurs- Within 0miue of the receipt of the alaFrm, then this EAL is notapplicable and no emergency declaration is warranted.In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant PROTECTEDAREA not extinguished xvithin 60 minutes may also potentially degrade the love! of plantsafety. This basis extends to a FIRE occUrring within the P-ROTECTEiD AREA of aý IS/Slocated outside the plant P-ROTEC TED9 AREA. [Sentence for plants w.'th anQ 1SFSI obutsd theplant P-rotected Area]E-AL-#4If a FIRE within the plant or ISFSI [forplants with an I-SFS! outSide the plant PDotected, Ar..a]PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency(e.g., a local town Fire Department), then the level of plant safety is potentially degraded. Thedispatch of an offsite firefighting agency to the site requires an emergency declaration only if itis needed to actively support firefighting efforts because the fire is beyond the capability of theFire Brigade to extinguish. Declaration is not necessary if the agency resources are placed onstand-by, or supporting post-extinguishment recovery or investigation actions.Basi Related~~nr Reau'rcrnts afro(mn AooeRnrJdiAppendix R to 10 CFR 50, states n part--.Criterion 3 of Appendix A to this part specifies that "Structures, systems, andiriir uui ^npnr m[ ni ruel n:i II Iii Q:1e' ý-; 1:1 [ me nu- ýu InnL nnn iii nri:i~ Ii 11111 n r-Ti-rt-r I ---J -..- .- .-----with other safety requirements, the probability ard effect of fires When onnsideiRng the effects of fire, those systems ass"oiated with achieving am, ,inrin f ... ..cndtn5 ~ m m iri nr hnet 'ftx ~~~5e.--.-. .----..-.---.--. ..... .. c.....wn .damage to them. can lead to core damage resulting fro. lss of colant through boil o.ff.fire may affect safe shutdcOWRsystems and because the lo9s Of fLuncrtion Ofsystems used to mitigate the consequences of design basis accidents under post frconditions does not per se impact public safety, the need to limnit fire damnage to systemsrequired to achieve and mnaintain safe Shutdown conditions is greater than the need toli.it fire damage to those systems required to .itigate the .cnsequen~esof design_basis accidents.In addition, Appendix R to 10 CPR 50, aFmGog other cnsiderations, the use of 1 haoufire barriers for the encloGsure Of cable and equipment and associated non safetywcircuits; of onePage 153 of 276 ATTACHMENT 1EAL Bases,redundant train (G.2.c;). As used in EAL #2, the 30 to ve,,;' a single alarm i wellwithin this WorSt caSe 1 hour time period.Depending upon the plant mode at the time of the event, escalation of the emergencyclassification level would be via IC CA6 or SA9.CPNPP Basis Reference(s):1. NEI 99-01 HU4Page 154 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:H -Hazards and Other Conditions Affecting Plant Safety5 -Hazardous GasesGaseous release IMPEDING access to equipment necessary fornormal plant operations, cooldown or shutdownEAL:HA5.1 AlertRelease of a toxic, corrosive, asphyxiant or flammable gas into any Table H-2 rooms orareasANDEntry into the room or area is prohibited or IMPEDED (Note 5)Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, thenno emergency classification is warranted.Table H-2 Safe Operation & Shutdown Rooms/AreasRoom/Area Mode ApplicabilityCharging Pump Rooms 1, 2, 3, 4, 5, 6CVCS Valve Rooms 1, 2, 3, 4, 5, 61 E Switchgear Rooms AllRHR Pump Rooms 4, 5, 6Mode Applicability:AllDefinition(s):IMPEDE(D) -Personnel access to a room or area is hindered to an extent that extraordinarymeasures are necessary to facilitate entry of personnel into the affected room/area(e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).CPNPP Basis:If the equipment in the listed room or area was already inoperable, or out-of-service, before theevent occurred, then no emergency should be declared since the event will have no adverseimpact beyond that already allowed by Technical Specifications at the time of the event.The list of plant rooms or areas with entry-related mode applicability identified specify thoserooms or areas that contain equipment which require a manual/local action as specified inoperating procedures used for normal plant operation, cooldown and shutdown. Rooms orareas in which actions of a contingent or emergency nature would be performed (e.g., anaction to address an off-normal or emergency condition such as emergency repairs, correctivemeasures or emergency operations) are not included. In addition, the list specifies the plantmode(s) during which entry would be required for each room or area (ref. 1).I Page 155 of 276 ATTACHMENT 1EAL BasesNEI 99-01 Basis:This IC addresses an event involving a release of a hazardous gas that precludes or impedesaccess to equipment necessary to maintain normal plant operation, or required for a normalplant cooldown and shutdown. This condition represents an actual or potential substantialdegradation of the level of safety of the plant.An Alert declaration is warranted if entry into the affected room/area is, or may be, procedurallyrequired during the plant operating mode in effect at the time of the gaseous release. Theemergency classification is not contingent upon whether entry is actually necessary at the timeof the release.Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the,,nEergencY Emergency Coordinator's judgment that the gas concentration in theaffected room/area is sufficient to preclude or significantly impede procedurally requiredaccess. This judgment may be based on a variety of factors including an existing job hazardanalysis, report of ill effects on personnel, advice from a subject matter expert or operatingexperience with the same or similar hazards. Access should be considered as impeded ifextraordinary measures are necessary to facilitate entry of personnel into the affectedroom/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinelyemployed).An emergency declaration is not warranted if any of the following conditions apply:" The plant is in an operating mode different than the mode specified for the affectedroom/area (i.e., entry is not required during the operating mode in effect at the time of thegaseous release). For example, the plant is in Mode 1 when the gaseous release occurs,and the procedures used for normal operation, cooldown and shutdown do not requireentry into the affected room until Mode 4." The gas release is a planned activity that includes compensatory measures which addressthe temporary inaccessibility of a room or area (e.g., fire suppression system testing)." The action for which room/area entry is required is of an administrative or record keepingnature (e.g., normal rounds or routine inspections).* The access control measures are of a conservative or precautionary nature, and would notactually prevent or impede a required action.An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerouslevels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment.This reduces the concentration of oxygen below the normal level of around 19%, which canlead to breathing difficulties, unconsciousness or even death.This EAL does not apply to firefighting activities that automatically or manually activate a firesuppression system in an area, or to intentional in"rting of etainment.. (BVWR -ony).Escalation of the emergency classification level would be via Recognition Category A R, C or FICs.CPNPP Basis Reference(s):1. Attachment 3 Safe Operation & Shutdown Areas Tables R-3 & H-2 Bases2. NEI 99-01 HA5Page 156 of 276 ATTACHMENT 1EAL BasesI Page 157 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:H -Hazards and Other Conditions Affecting Plant Safety6 -Control Room EvacuationControl Room evacuation resulting in transfer of plant control toalternate locationsEAL:HA6.1 AlertAn event has resulted in plant control being transferred from the Control Room to theRemote Shutdown Panel (RSP)Mode Applicability:AllDefinition(s):NoneCPNPP Basis:Upon evacuation of the Control Room plant control is established at the Remote ShutdownPanel (RSP). ABN-905A/B "Loss of Control Room Habitability" and ABN-803A/B "Response toa Fire in the Control Room or Cable Spreading Room" provide the instructions for tripping theunit, and maintaining RCS inventory and Hot Shutdown conditions from outside the ControlRoom. The Shift Manager (SM) determines if the Control Room is inoperable and requiresevacuation. Control Room inhabitability may be caused by fire, dense smoke, noxious fumes,bomb threat in or adjacent to the Control Room, or other life threatening conditions. (Ref. 1, 2,3, 4, 5).Inability to establish plant control from outside the Control Room escalates this event to a SiteArea Emergency per EAL HS6.1.NEI 99-01 Basis:This IC addresses an evacuation of the Control Room that results in transfer of plant control toalternate locations outside the Control Room. The loss of the ability to control the plant fromthe Control Room is considered to be a potential substantial degradation in the level of plantsafety.Following a Control Room evacuation, control of the plant will be transferred to alternateshutdown locations. The necessity to control a plant shutdown from outside the Control Room,in addition to responding to the event that required the evacuation of the Control Room, willpresent challenges to plant operators and other on-shift personnel. Activation of the ERO andemergency response facilities will assist in respondingto these challenges.Escalation of the emergency classification level would be via IC HS6.Page 158 of 276 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):1. DBD-ME-003 Control Room Habitability2. ABN-905A Loss of Control Room Habitability3. ABN-905B Loss of Control Room Habitability4. ABN-803A Response to a Fire in the Control Room or Cable Spreading Room5. ABN-803B Response to a Fire in the Control Room or Cable Spreading Room6. NEI 99-01 HA6Page 159 of 276 ATTACHMENT 1EAL BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 6 -Control Room EvacuationInitiating Condition: Inability to control a key safety function from outside the Control RoomEAL:HS6.1 Site Area EmergencyAn event has resulted in plant control being transferred from the Control Room to theRemote Shutdown Panel (RSP)ANDControl of any of the following key safety functions is not re-established within 15 min.(Note 1):" Reactivity" Core Cooling" RCS heat removalNote 1: The Emergency Coordinator should declare the event promptly upon determining that.time limit hasbeen exceeded, or will likely be exceeded.Mode Applicability:AllDefinition(s):NoneCPNPP Basis:Upon evacuation of the Control Room plant control is established at the Remote ShutdownPanel (RSP). ABN-905A/B "Loss of Control Room Habitability" and ABN-803A/B "Response toa Fire in the Control Room or Cable Spreading Room" provide the instructions for tripping theunit, and maintaining RCS inventory and Hot Shutdown conditions from outside the ControlRoom. The Shift Manager (SM) determines if the Control Room is inoperable and requiresevacuation. Control Room inhabitability may be caused by fire, dense smoke, noxious fumes,bomb threat in or adjacent to the Control Room, or other life threatening conditions. (Ref. 1, 2,3, 4, 5).The intent of this EAL is to capture events in which control of the plant cannot be reestablishedin a timely manner. The fifteen minute time for transfer starts when the Control Room begins tobe evacuated (not when the ABN is entered). The time interval is based on how quickly controlmust be reestablished without core uncovery and/or core damage. The determination ofwhether or not control is established from outside the Control Room is based on EmergencyCoordinator judgment. The Emergency Coordinator is expected to make a reasonable,informed judgment that control of the plant from outside the Control Room cannot beestablished within the fifteen minute interval.Once the Control Room is evacuated, the objective is to establish control of important plantequipment and maintain knowledge of important plant parameters in a timely manner. PrimaryPage 160 of 276 ATTACHMENT 1EAL Basesemphasis should be placed on components and instruments that supply protection for andinformation about safety functions. Typically, these safety functions are reactivity control(ability to shutdown the reactor and maintain it shutdown), RCS inventory (ability to cool thecore), and secondary heat removal (ability to maintain a heat sink).NEI 99-01 Basis:This IC addresses an evacuation of the Control Room that results in transfer of plant control toalternate locations, and the control of a key safety function cannot be reestablished in a timelymanner. The failure to gain control of a key safety function following a transfer of plant controlto alternate locations is a precursor to a challenge to one or more fission product barrierswithin a relatively short period of time.The determination of whether or not "control" is established at the remote safe shutdownlocation(s) is based on Emergency Directer-Coordinator judgment. The Emergency DiieetOrCoordinator is expected to make a reasonable, informed judgment within (the site ,pecific timefeF1t4ansfeF)15 minutes whether or not the operating staff has control of key safety functionsfrom the remote safe shutdown location(s).Escalation of the emergency classification level would be via IC FG1 or CG1CPNPP Basis Reference(s):1. DBD-ME-003 Control Room Habitability2. ABN-905A Loss of Control Room Habitability3. ABN-905B Loss of Control Room Habitability4. ABN-803A Response to a Fire in the Control Room or Cable Spreading Room5. ABN-803B Response to a Fire in the Control Room or Cable Spreading Room6. NEI 99-01 HS6I Page 161 of 276 ATTACHMENT 1EAL BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 7 -Emergency Coordinator JudgmentInitiating Condition: Other conditions existing that in the judgment of the EmergencyCoordinator warrant declaration of a UEEAL:HU7.1 Unusual EventOther conditions exist which in the judgment of the Emergency Coordinator 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.No releases of radioactive material requiring offsite response or monitoring are expectedunless further degradation of SAFETY SYSTEMS occurs.Mode Applicability:AllDefinition(s):SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/orplacing it in the cold shutdown condition, including the ECCS. These are typically systemsclassified as safety-related (as defined in 10CFR50.2):Those structures, systems and components that are relied upon to remain functional duringand 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 could resultin potential offsite exposures.CPNPP Basis:The Emergency Coordinator is the designated onsite individual having the responsibility andauthority for implementing the CPNPP Radiological Emergency Response Plan. The ShiftManager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions asoutlined in the Emergency Plan implementing procedures. If required by the emergencyclassification or if deemed appropriate by the Emergency Coordinator, emergency responsepersonnel are notified and instructed to report to their emergency response locations. In thismanner, the individual usually in charge of activities in the Control[Room is responsible forinitiating the necessary emergency response, but Plant Management is expected to managethe emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1).NEI 99-01 Basis:This IC addresses unanticipated conditions-not addressed explicitly elsewhere but that warrantdeclaration of an emergency because conditions exist which are believed by the EmergencyPage 162 of 276 ATTACHMENT 1EAL BasesDireeteF-Coordinator to fall under the emergency classification level description for anNQUEUnusual Event.Page 163 of 276 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):1. CPNPP Radiological Emergency Response Plan section 1.1.2 Response2. NEI 99-01 HU7Page 164 of 276 ATTACHMENT 1EAL BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 7 -Emergency Coordinator JudgmentInitiating Condition: Other conditions exist that in the judgment of the EmergencyCoordinator warrant declaration of an AlertEAL:HA7.1 AlertOther conditions exist which, in the judgment of the Emergency Coordinator, 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.Mode Applicability:AllDefinition(s):HOSTILE ACTION -An act toward CPNPP or its personnel that includes the use of violentforce to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end.This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or otherdevices used to deliver destructive force. Other acts that satisfy the overall intent may beincluded. Hostile action should not be construed to include acts of civil disobedience orfelonious acts that are not part of a concerted attack on CPNPP. Non-terrorism-based EALsshould be used to address such activities (i.e., this may include violent acts betweenindividuals in the owner controlled area).CPNPP Basis:The Emergency Coordinator is the designated onsite individual having the responsibility andauthority for implementing the CPNPP Radiological Emergency Response Plan. The ShiftManager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions asoutlined in the Emergency Plan implementing procedures. If required by the emergencyclassification or if deemed appropriate by the Emergency Coordinator, emergency responsepersonnel are notified and instructed to report to their emergency response locations. In thismanner, the individual usually in charge of activities in the Control Room is responsible forinitiating the necessary emergency response, but Plant Management is expected to managethe emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1).NEI 99-01 Basis:This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrantdeclaration of an emergency because conditions exist which are believed by the EmergencyI Direotor-Coordinator to fall under the emergency classification level description for an Alert.I Page 165 of 276 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):1. CPNPP Radiological Emergency Response Plan section 1.1.2 Response2. NEI 99-01 HA7Page 166 of 276 ATTACHMENT 1EAL BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 7 -Emergency Coordinator JudgmentInitiating Condition: Other conditions existing that in the judgment of the EmergencyCoordinator warrant declaration of a Site Area EmergencyEAL:HS7.1 Site Area EmergencyOther conditions exist which in the judgment of the Emergency Coordinator 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 could leadto 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 beyond the EXCLUSION AREABOUNDARYMode Applicability:AllDefinition(s):EXCLUSION AREA BOUNDARY -Exclusion Area Boundary is a synonymous term for SiteBoundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion AreaBoundary. This boundary is used for establishing effluent release limits with respect to therequirements of 10CFR20. See also CPNPP Emergency Plan Appendix E, Complex andOwner Controlled Area and CCNPP ODCM Section 5.0 Design Features.HOSTILE ACTION -An act toward CPNPP or its personnel that includes the use of violentforce to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end.This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or otherdevices used to deliver destructive force. Other acts that satisfy the overall intent may beincluded. Hostile action should not be construed to include acts of civil disobedience orfelonious acts that are not part of a concerted attack on CPNPP. Non-terrorism-based EALsshould be used to address such activities (i.e., this may include violent acts betweenindividuals in the owner controlled area)CPNPP Basis:The Emergency Coordinator is the designated onsite individual having the responsibility andauthority for implementing the CPNPP Radiological Emergency Response Plan. The ShiftManager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions asoutlined in the Emergency Plan implementing procedures. If required by the emergencyclassification or if deemed appropriate by the Emergency Coordinator, emergency responsepersonnel are notified and instructed to report to their emergency response locations. In thismanner, the individual usually in charge of activities in the Control Room is responsible forinitiating the necessary emergency response, but Plant Management is expected to managethe emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1).Page 167 of 276 ATTACHMENT 1EAL BasesNEI 99-01 Basis:This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrantdeclaration of an emergency because conditions exist which are believed by the EmergencyDir-eeter-Coordinator to fall under the emergency classification level description for a Site AreaEmergency.CPNPP Basis Reference(s):1. CPNPP Radiological Emergency Response Plan section 1.1.2 ResponseI 2. NEI 99-01 HS7Page 168 of 276 ATTACHMENT 1EAL BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 7 -Emergency Coordinator JudgmentInitiating Condition: Other conditions exist which in the judgment of the EmergencyCoordinator warrant declaration of a General EmergencyEAL:HG7.1 General EmergencyOther conditions exist which in the judgment of the Emergency Coordinator 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 offsite formore than the immediate site areaMode Applicability:AllDefinition(s):HOSTILE ACTION -An act toward CPNPP or its personnel that includes the use of violentforce to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end.This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or otherdevices used to deliver destructive force. Other acts that satisfy the overall intent may beincluded. Hostile action should not be construed to include acts of civil disobedience orfelonious acts that are not part of a concerted attack on CPNPP. Non-terrorism-based EALsshould be used to address such activities (i.e., this may include violent acts betweenindividuals in the owner controlled area).IMMINENT -The trajectory of events or conditions is such that an EAL will be met within arelatively short period of time regardless of mitigation or corrective actions.CPNPP Basis:The Emergency Coordinator is the designated onsite individual having the responsibility andauthority for implementing the CPNPP Radiological Emergency Response Plan. The ShiftManager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions asoutlined in the Emergency Plan implementing procedures. If required by the emergencyclassification or if deemed appropriate by the Emergency Coordinator, emergency responsepersonnel are notified and instructed to report to their emergency response locations. In thismanner, the individual usually in charge of activities in the Control Room is responsible forinitiating the necessary emergency response, but Plant Management is expected to managethe emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1).Releases can reasonably be expected to exceed EPA PAG plume exposure levels outside theSite Boundary.Page 169 of 276 ATTACHMENT 1EAL BasesNEI 99-01 Basis:This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrantdeclaration of an emergency because conditions exist which are believed by the EmergencyDi-eeter-Coordinator to fall under the emergency classification level description for a GeneralEmergency.CPNPP Basis Reference(s):1. CPNPP Radiological Emergency Response Plan section 1.1.2 Response2. NEI 99-01 HG7L Page 170 of 276 ATTACHMENT 1EAL BasesCategory S -System MalfunctionEAL Group: Hot Conditions (RCS temperature greater than 2001F);EALs in this category are applicable only in one or morehot operating modes.Numerous system-related equipment failure events that warrant emergency classification havebeen identified in this category. They may pose actual or potential threats to plant safety.The events of this category pertain to the following subcategories:1. Loss of Emergency AC PowerLoss of emergency electrical power can compromise plant safety system operabilityincluding decay heat removal and emergency core cooling systems which may benecessary to ensure fission product barrier integrity. This category includes loss of onsiteand offsite sources for 6.9KV AC safeguard buses.2. Loss of Vital DC PowerLoss of emergency electrical power can compromise plant safety system operabilityincluding decay heat removal and emergency core cooling systems which may benecessary to ensure fission product barrier integrity. This category includes loss of vitalplant 125 VDC power sources.3. Loss of Control Room IndicationsCertain events that degrade plant operator ability to effectively assess plant conditionswithin the plant warrant emergency classification. Losses of indicators are in thissubcategory.4. RCS ActivityDuring 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 increase fromthese base-line levels (2% -5% clad failures) is indicative of fuel failures and is coveredunder the Fission Product Barrier Degradation category. However, lesser amounts of claddamage may result in coolant activity exceeding Technical Specification limits. Thesefission products will be circulated with the reactor coolant and can be detected by coolantsampling.5. RCS LeakageThe reactor vessel provides a volume for the coolant that covers the reactor core. Thereactor pressure 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 indicatespotential pipe cracks that may propagate to an extent threatening fuel clad, RCS andcontainment integrity.6. RPS FailureThis subcategory includes events related to failure of the Reactor Protection System (RPS)to initiate and complete reactor trips. In the plant licensing basis, postulated failures of theRPS to complete a reactor trip comprise a specific set of analyzed events referred to asPage 171 of 276 ATTACHMENT IEAL BasesAnticipated Transient Without Scram (ATWS) events. For EAL classification, however,ATWS is intended to mean any trip failure event that does not achieve reactor shutdown. IfRPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk andcould cause a threat to fuel clad, RCS and containment integrity.7. Loss of CommunicationsCertain events that degrade plant operator ability to effectively communicate with essentialpersonnel within or external to the plant warrant emergency classification.8. Containment FailureFailure of containment isolation capability (under conditions in which the containment is notcurrently challenged) warrants emergency classification. Failure of containment pressurecontrol capability also warrants emergency classification.9. Hazardous Event Affectinq Safety SystemsVarious natural and technological events that result in degraded plant safety systemperformance or significant visible damage warrant emergency classification under thissubcategory.Page 172 of 276 1 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:S -System Malfunction1 -Loss of Emergency AC PowerLoss of all offsite AC power capability to safeguard buses for 15minutes or longerEAL:SUI.1 Unusual EventLoss of all offsite AC power capability, Table S-1, to 6.9 KV safeguard buses uEA1 anduEA2 for greater than or equal to 15 min. (Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Table S-1 AC Power SourcesOffsite:* 138 KV switchyard circuit* 345 KV switchyard circuitOnsite:* uEG1" uEG2Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):NoneBasis:CPNPP Basis:For emergency classification purposes, "capability" means that an offsite AC power source(s)is available to the safeguard buses, whether or not the buses are powered from it.The safeguards AC distribution system power sources consist of the preferred and alternateoffsite power sources, and the onsite standby emergency diesel generators uEG1 and uEG2.Offsite power is supplied to the plant switchyards from the transmission network by five 345 KVand two 138 KV transmission lines. From the switchyards, two electrically and physicallyseparated circuits provide AC power through step down startup transformers, to the 6.9 kVsafeguard buses. The 138 kV switchyard circuit is the preferred source for Unit 2 and alternatesource for-Unit 1. The 345 KV circuit is the preferred source for Unit 1 and alternate source forUnit 2. The onsite AC distribution system is divided into redundant trains so that the loss of anyone load group does not prevent the minimum safety functions from being performed. Eachtrain has connections to two offsite power sources and a dedicated diesel generator. Eachoffsite circuit can supply the Unit 1 and Unit 2 6.9 KV safeguard buses. (ref. 1, 2, 3, 4)Page 173 of 276 ATTACHMENT 1EAL BasesThe 15-minute interval was selected as a threshold to exclude transient or momentary powerlosses.NEI 99-01 Basis:This IC addresses a prolonged loss of offsite power. The loss of offsite power sources rendersthe plant more vulnerable to a complete loss of power to AC emergency buses. This conditionrepresents a potential reduction in the level of safety of the plant.For emergency classification purposes, "capability" means that an offsite AC power source(s)is available to the emergency buses, whether or not the buses are powered from it.Fifteen minutes was selected as a threshold to exclude transient or momentary losses ofoffsite power.Escalation of the emergency classification level would be via IC SAI.CPNPP Basis Reference(s):1. FSAR Figure 8.3-12. FSAR Section 8.23. FSAR Section 8.34. Technical Specifications B3.8.15. ABN-601 Response to a 138/345 KV System Malfunction6. ABN-602 Response to a 6900/480V System Malfunction7. NEI 99-01 SUWPage 174 of 276 1 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:S -System Malfunction1 -Loss of Emergency AC PowerLoss of all but one AC power source to safeguard buses for 15minutes or longerEAL:SAI.1 AlertAC power capability, Table S-1, to 6.9 KV safeguard buses uEA1 and uEA2 reduced to asingle power source for greater than or equal to 15 min. (Note 1)ANDAny additional single power source failure will result in loss of all AC power to SAFETYSYSTEMSNote 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Table S-1 AC Power SourcesOffsite:* 138 KV switchyard circuit* 345 KV switchyard circuitOnsite:* uEG1" uEG2Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/orplacing it in the cold shutdown condition, including the ECCS. These are typically systemsclassified as safety-related (as defined in 1OCFR50.2):Those structures, systems and components that are relied upon to remain functional duringand 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.Basis:CPNPP Basis:Page 175 of 276 ATTACHMENT 1EAL BasesFor emergency classification purposes, "capability" means that an offsite AC power source(s)is available to the emergency buses, whether or not the buses are powered from it.The condition indicated by this EAL is the degradation of the offsite and onsite power sourcessuch that any additional single failure would result in a loss of all AC power to the safeguardbuses.The safeguards AC distribution system power sources consist of the preferred and alternateoffsite power sources, and the onsite standby emergency diesel generators uEG1 and uEG2.Offsite power is supplied to the plant switchyards from the transmission network by five 345 KVand two 138 KV transmission lines. From the switchyards, two electrically and physicallyseparated circuits provide AC power through step down startup transformers, to the 6.9 kVsafeguard buses. The 138 kV switchyard circuit is the preferred source for Unit 2 and alternatesource for Unit 1. The 345 KV circuit is the preferred source for Unit 1 and alternate source forUnit 2. The onsite AC distribution system is divided into redundant trains so that the loss of anyone load group does not prevent the minimum safety functions from being performed. Eachtrain has connections to two offsite power sources and a dedicated diesel generator. Eachoffsite circuit can supply the Unit 1 and Unit 2 6.9 KV safeguard buses. (ref. 1, 2, 3, 4).The 15-minute interval was selected as a threshold to exclude transient or momentary powerlosses. If the capability of a second source of emergency bus power is not restored within 15minutes, an Alert is declared under this EAL.This hot condition EAL is equivalent to the cold condition EAL CU2. 1.NEI 99-01 Basis:This IC describes a significant degradation of offsite and onsite AC power sources such thatany additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. Inthis condition, the sole AC power source may be powering one, or more than one, train ofsafety-related equipment. This IC provides an escalation path from IC SUI.An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplyingrequired power to an emergency bus. Some examples of this condition are presented below." A loss of all offsite power with a concurrent failure of all but one emergency powersource (e.g., an onsite diesel generator)." A loss of all offsite power and loss of all emergency power sources (e.g., onsite dieselgenerators) with a single train of emergency buses being back-fed from the unit maingenerator." A loss of emergency power sources (e.g., onsite diesel generators) with a single train ofemergency buses being ba-k-fed from an offsite power source.Fifteen minutes was selected as a threshold to exclude transient or momentary losses ofpower.Escalation of the emergency classification level would be via IC SSI.I Page 176 of 276 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):1. FSAR Figure 8.3-12. FSAR Section 8.23. FSAR Section 8.34. Technical Specifications B3.8.15. ABN-601 Response to a 138/345 KV System Malfunction6. ABN-602 Response to a 6900/480V System Malfunction7. NEI 99-01 SAlI Page 177 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:S -System MalfunctionI -Loss of Emergency AC PowerLoss of all offsite power and all onsite AC power to safeguard busesfor 15 minutes or longerEAL:EAL:SS1.1 Site Area EmergencyLoss of all offsite and all onsite AC power capability, Table S-1, to 6.9 KV safeguard busesuEA1 and uEA2 for greater than or equal to 15 min. (Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Table S-1 AC Power SourcesOffsite:* 138 KV switchyard circuit* 345 KV switchyard circuitOnsite:" uEGI* uEG2Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):NoneCPNPP Basis:For emergency classification purposes, "capability" means that an AC power source isavailable to the safeguard buses, whether or not the buses are powered from it.The safeguards AC distribution system power sources consist of the preferred and alternateoffsite power sources, and the onsite standby emergency diesel generators uEG1 and uEG2.Offsite power is supplied to the plant switchyards from the transmission network by five 345 KVand two 138 KV transmission lines. From the switchyards, two electrically and physicallyseparated circuits provide AC power through step down startup transformers, to the 6.9 kVsafeguard buses. The 138 kV switchyard circuit is the preferred source for Unit 2 and alternatesource for Unit 1. The 345 KV circuit is the preferred source for Unit 1 and alternate source forUnit 2. The onsite AC distribution system is divided into redundant trains so that the loss of anyone load group does not prevent the minimum safety functions from being performed. Eachtrain has connections to two offsite power sources and a dedicated diesel generator. Eachoffsite circuit can supply the Unit I and Unit 2 6.9 KV safeguard buses. (ref. 1, 2, 3, 4).The 15-minute interval was selected as a threshold to exclude transient or momentary powerlosses. The interval begins when both offsite and onsite AC power capability are lost.Page 178 of 276 ATTACHMENT 1EAL BasesNEI 99-01 Basis:This IC addresses a total loss of AC power that compromises the performance of all SAFETYSYSTEMS requiring electric power including those necessary for emergency core cooling,containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.In addition, fission product barrier monitoring capabilities may be degraded under theseconditions. This IC represents a condition that involves actual or likely major failures of plantfunctions needed for the protection of the public.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Escalation of the emergency classification level would be via ICs AG-1-RG1, FG1 or SGI.CPNPP Basis Reference(s):1.2.3.4.5.6.7.FSAR Figure 8.3-1FSAR Section 8.2FSAR Section 8.3Technical Specifications B3.8.1ABN-601 Response to a 138/345 KV System MalfunctionABN-602 Response to a 6900/480V System MalfunctionNEI 99-01 SS1Page 179 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:S -System Malfunction1 -Loss of Emergency AC PowerProlonged loss of all offsite and all onsite AC power to safeguardbusesEAL:SGI.1 General EmergencyLoss of all offsite and all onsite AC power capability, Table S-I, to 6.9 KV safeguard busesuEA1 and uEA2AND EITHER:* Restoration of at least one safeguard bus from a Table S-1 source or APDG in lessthan 4 hours is not likely (Note 1)* CSFST Core Cooling RED Path conditions metNote 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Table S-1 AC Power SourcesOffsite:* 138 KV switchyard circuit* 345 KV switchyard circuitOnsite:* uEG1* uEG2Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):NoneCPNPP Basis:This EAL is indicated by the extended loss of all offsite and onsite AC power capability to 6.9KV safeguard buses uEA1 and uEA2 either for greater then the CPNPP Station Blackout(SBO) coping analysis time (4 hrs.) (ref. 7) or that has resulted in indications of an actual lossof adequate core cooling.Indication of continuing core cooling degradation is manifested by CSFST Core Cooling REDPath conditions being met. (ref. 8).For emergency classification purposes, "capability" means that an AC power source isavailable to the emergency buses, whether or not the buses are powered from it.The safeguards AC distribution system power sources consist of the preferred and alternateoffsite power sources, and the onsite standby emergency diesel generators uEG1 and uEG2.Page 180 of 276 ATTACHMENT 1EAL BasesOffsite power is supplied to the plant switchyards from the transmission network by five 345 KVand two 138 KV transmission lines. From the switchyards, two electrically and physicallyseparated circuits provide AC power through step down startup transformers, to the 6.9 kVsafeguard buses. The 138 kV switchyard circuit is the preferred source for Unit 2 and alternatesource for Unit 1. The 345 KV circuit is the preferred source for Unit 1 and alternate source forUnit 2. The onsite AC distribution system is divided into redundant trains so that the loss of anyone load group does not prevent the minimum safety functions from being performed. Eachtrain has connections to two offsite power sources and a dedicated diesel generator. Eachoffsite circuit can supply the Unit 1 and Unit 2 6.9 KV safeguard buses. (ref. 1, 2, 3, 4).CPNPP has also provided a set of non-safety related Alternate Power Diesel Generators(APDGs) for each unit with the capability to connect to a safeguards bus one at a time toprovide defense-in-depth for safe shutdown of a unit during outages or during extendedduration of an inoperable offsite circuit on occurrence of concurrent loss of offsite power andfailure of EDGs. The APDGs can provide 3450 kVA to provide long term cooling of each unit(ref. 3).Four hours is the station blackout coping time (ref 7).Indication of continuing core cooling degradation must be based on fission product barriermonitoring with particular emphasis on Emergency Coordinator judgment as it relates toimminent loss of fission product barriers and degraded ability to monitor fission productbarriers. Indication of continuing core cooling degradation is manifested by CSFST CoreCooling RED path conditions being met (ref. 8). Critical Safety Function Status Tree (CSFST)Core Cooling-RED path indicates significant core exit superheating and core uncovery. (ref. 3).NEI-9901 Basis:This IC addresses a prolonged loss of all power sources to AC emergency buses. A loss of allAC power compromises the performance of all SAFETY SYSTEMS requiring electric powerincluding those necessary for emergency core cooling, containment heat removal/pressurecontrol, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buseswill lead to a loss of one or more fission product barriers. In addition, fission product barriermonitoring capabilities may be degraded under these conditions.The EAL should require declaration of a General Emergency prior to meeting the thresholdsfor IC FGI. This will allow additional time for implementation of offsite protective actions.Escalation of the emergency classification from Site Area Emergency will occur if it is projectedthat power cannot be restored to at least one AC emergency bus by the end of the analyzedstation blackout coping period. Beyond this time, plant responses and event trajectory aresubject to greater uncertainty, and there is an increased likelihood of challenges to multiplefission product barriers.The estimate for restoring at least one emergency bus should be based on a realistic appraisalof the situation. Mitigation actions with a low probability of success should not be used as abasis for delaying a classification upgrade. The goal is to maximize the time available toprepare for, and implement, protective actions for the public.The EAL will also require a General Emergency declaration if the loss of AC power results inparameters that indicate an inability to adequately remove decay heat from the core.Page 181 of 276 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):1. FSAR Figure 8.3-12. FSAR Section 8.23. FSAR Section 8.34. Technical Specifications B3.8.15. ABN-601 Response to a 138/345 KV System Malfunction6. ABN-602 Response to a 6900/480V System Malfunction7. FSAR Section 8B Station Blackout8. FRC-O.1A/B Response to Inadequate Core Cooling9. NEI 99-01 SG1Page 182 of 276 1 ATTACHMENT 1EAL BasesCategory:Subcategory:S -System Malfunction1 -Loss of Emergency AC PowerInitiating Condition: Loss of all AC and vital DC power sources for 15 minutes or longerEAL:SG1.2 General EmergencyLoss of all offsite and all onsite AC power capability, Table S-1, to 6.9 KV safeguard busesuEA1 and uEA2 for greater than or equal to 15 min.ANDLess than 105 VDC on all 125 VDC safeguard buses uED1, uED2, uED3 and uED4 forgreater than or equal to 15 min.(Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Table S-1 AC Power SourcesOffsite:0 138 KV switchyard circuit* 345 KV switchyard circuitOnsite:" uEG1* uEG2Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):NoneCPNPP Basis:This EAL is indicated by the loss of all offsite and onsite emergency AC power capability to 6.9KV safeguard buses uEA1 and uEA2 for greater than 15 minutes in combination with degradedvital DC power voltage. This EAL addresses operating experience from the March 2011accident at Fukushima Daiichi.For emergency classification purposes, "capability" means that an AC power source isavailable to the emergency buses, whether or not the buses are powered from it.The safeguards AC distribution system power sources consist of the preferred and alternateoffsite power sources, and the onsite standby emergency diesel generators uEG1 and uEG2.Offsite power is supplied to the plant switchyards from the transmission network by five 345 KVand two 138 KV transmission lines. From the switchyards, two electrically and physicallyPage 183 of 276 ATTACHMENT 1EAL Basesseparated circuits provide AC power through step down startup transformers, to the 6.9 kVsafeguard buses. The 138 kV switchyard circuit is the preferred source for Unit 2 and alternatesource for Unit 1. The 345 KV circuit is the preferred source for Unit 1 and alternate source forUnit 2. The onsite AC distribution system is divided into redundant trains so that the loss of anyone load group does not prevent the minimum safety functions from being performed. Eachtrain has connections to two offsite power sources and a dedicated diesel generator. Eachoffsite circuit can supply the Unit 1 and Unit 2 6.9 KV safeguard buses. (ref. 1, 2, 3, 4).The safeguards 125 VDC buses are the Class 1E buses uED1, uED2, uED3 and uED4 (ref. 7,8,9).Each redundant safeguards 125 VDC system consists of two independent batteries eachhaving one main distribution bus, two static battery chargers (one spare), and local distributionpanels. For Unit 1, batteries BT1 ED1 and BT1 ED3 feed all train A load requirements, whilebatteries BT1 ED2 and BT1 ED4 supply train B load requirements.For Unit 2, batteries BT2ED1 and BT2ED3 feed all train A load requirements, while batteriesBT2ED2 and BT2ED4 supply train B load requirements. There are no bus ties or sharing ofpower supplies between redundant trains (ref. 7).Minimum DC bus voltage is 105 VDC (ref. 10). Bus voltage may be monitored from thefollowing indications (ref. 12):Control Room Panel CP-10 Annunciator u--ALB-10B Plant ComputerV-IED1, 125VDC SWITCH PNL lED1 VOLT 1.13 V6501A BATT BT1ED1 VOLTV-1 ED2, 125VDC SWITCH PNL 1ED2 VOLT 2.13 V6502A BATT BT1ED2 VOLTV-I ED3, 125VDC SWITCH PNL IED3 VOLT 1.9 noneV-i ED4, 125VDC SWITCH PNL IED4 VOLT 3.9 V6504A BATT BT1ED4 VOLTNEI-9901 Basis:This IC addresses a concurrent and prolonged loss of both emergency AC and Vital DCpower. A loss of all emergency AC power compromises the performance of all SAFETYSYSTEMS requiring electric power including those necessary for emergency core cooling,containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.A loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS.A sustained loss of both emergency AC and vital DC power will lead to multiple challenges tofission product barriers.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.The 15-minute emergency declaration clock begins at the point when both EAL thresholds aremet.Page 184 of 276 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):1. FSAR Figure 8.3-12. FSAR Section 8.23. PSAR Section 8.34. Technical Specifications B3.8.15. ABN-601 Response to a 138/345 KV System Malfunction6. ABN-602 Response to a 6900/480V System Malfunction7. FSAR 8.3.28. FSAR Figure 8.3-149. PSAR Figure 8.3-14A10. ECA-O.OA/B Loss of All AC Power11. SOP-605A/B 125 VDC Switchgear and Distribution Systems, Batteries and BatteryChargers12.ALM-0102A/B Alarm Procedures Manual, u-ALB-1OB, nos. 1.9, 1.13, 2.13, 3.813.NEI 99-01 SG8Page 185 of 276 ATTACHMENT 1EAL BasesCategory: S -System MalfunctionSubcategory: 2 -Loss of Vital DC PowerInitiating Condition: Loss of all vital DC power for 15 minutes or longerEAL:SS2.1 Site Area EmergencyLess than 105 VDC on all 125 VDC safeguard buses uED1, uED2, uED3 and uED4 forgreater than or equal to 15 min. (Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):NoneCPNPP Basis:The safeguards 125 VDC buses are the Class 1E buses uED1, uED2, uED3 and uED4 (ref. 1,2,3).Each redundant safeguards 125 VDC system consists of two independent batteries eachhaving one main distribution bus, two static battery chargers (one spare), and local distributionpanels. For Unit 1, batteries BT1 ED1 and BT1 ED3 feed all train A load requirements, whilebatteries BT1 ED2 and BT1 ED4 supply train B load requirements.For Unit 2, batteries BT2ED1 and BT2ED3 feed all train A load requirements, while batteriesBT2ED2 and BT2ED4 supply train B load requirements. There are no bus ties or sharing ofpower supplies between redundant trains (ref. 1).Minimum DC bus voltage is 105 VDC (ref. 4). Bus voltage may be monitored from the followingindications (ref. 6):Control Room Panel CP-10 Annunciator u--ALB-1OB Plant ComputerV-IED1, 125VDC SWITCH PNL IED1 VOLT 1.13 V6501A BATT BT1ED1 VOLTV-1ED2, 125VDC SWITCH PNL 1ED2 VOLT 2.13 V6502A BATT BT1ED2 VOLTV-1 ED3, 125VDC SWITCH PNL IED3 VOLT 1.9 V6503A BATT BT1ED3 VOLTV-I ED4, 125VDC SWITCH PNL 1ED4 VOLT 3.9 V6504A BATT BT1ED4 VOLTPage 186 of 276 ATTACHMENT 1EAL BasesNEI 99-01 Basis:This IC addresses a loss of vital DC power which compromises the ability to monitor andcontrol SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a majorfailure of plant functions needed for the protection of the public.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Escalation of the emergency classification level would be via ICs AG-1RGI, FG1 or SG8SGI.CPNPP Basis Reference(s):1. FSAR 8.3.22. FSAR Figure 8.3-143. FSAR Figure 8.3-14A4. ECA-O.OA/B Loss of All AC Power5. SOP-605A/B 125 VDC Switchgear and Distribution Systems, Batteries and BatteryChargers6. ALM-0102A/B Alarm Procedures Manual, u-ALB-1OB, nos. 1.9, 1.13, 2.13, 3.87. NEI 99-01 SS8I Page 187 of 276 1 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:S -System Malfunction3 -Loss of Control Room IndicationsUNPLANNED loss of Control Room indications for 15 minutes orlongerEAL:SU3.1 Unusual EventAn UNPLANNED event results in the inability to monitor one or more Table S-2parameters from within the Control Room for greater than or equal to 15 min. (Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Table S-2 Safety System Parameters" Reactor power" RCS level* RCS pressure" Core Exit T/C temperature" Level in at least one SG" Auxiliary or emergency feed flow inat least one SGMode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):UNPLANNED -A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient. The cause of the parameter changeor event may be known or unknown.CPNPP Basis:SAFETY SYSTEM parameters listed in Table S-2 are monitored in the Control Room througha combination of hard control panel indicators as well as computer based information systems.The Plant Process Computer, which displays SPDS required information, serves as aredundant compensatory indicator which may be utilized in lieu of normal Control Roomindicators (ref. 1, 2, 3, 4).Page 188 of 276 ATTACHMENT 1EAL BasesNEI 99-01 Basis:This IC addresses the difficulty associated with monitoring normal plant conditions without theability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition isa precursor to a more significant event and represents a potential degradation in the level ofsafety of the plant.As used in this EAL, an "inability to monitor" means that values for one or more of the listedparameters cannot be determined from within the Control Room. This situation would requirea loss of all of the Control Room sources for the given parameter(s). For example, the reactorpower level cannot be determined from any analog, digital and recorder source within theControl Room.An event involving a loss of plant indications, annunciators and/or display systems is evaluatedin accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine ifan NRC event report is required. The event would be reported if it significantly impaired thecapability to perform emergency assessments. In particular, emergency assessmentsnecessary to implement abnormal operating procedures, emergency operating procedures,and emergency plan implementing procedures addressing emergency classification, accidentassessment, or protective action decision-making.This EAL is focused on a selected subset of plant parameters associated with the key safetyfunctions of reactivity control, core cooling [PWR] /i RPV level and RCS heat removal.The loss of the ability to determine one or more of these parameters from within the ControlRoom is considered to be more significant than simply a reportable condition. In addition, if allindication sources for one or more of the listed parameters are lost, then the ability todetermine the values of other SAFETY SYSTEM parameters may be impacted as well. Forexample, if the value for reactor vessel level [PWRI / RPV water level [.WR] cannot bedetermined from the indications and recorders on a main control board, the SPDS or the plantcomputer, the availability of other parameter values may be compromised as well.Fifteen minutes was selected as a threshold to exclude transient or momentary losses ofindication.Escalation of the emergency classification level would be via IC SA2SA3.CPNPP Basis Reference(s):1. FSAR Section 7.52. DBD-EE-033 Detailed Control Room Design, 5.1.2, Figure 13. SOP 906 Plant Process Computer System Guidelines4. ABN 906 Plant Process Computer System Malfunction5. NEI 99-01 SU2Page 189 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:EAL:S -System Malfunction3 -Loss of Control Room IndicationsUNPLANNED loss of Control Room indications for 15 minutes orlonger with a significant transient in progressSA3.1 AlertAn UNPLANNED event results in the inability to monitor one or more Table S-2parameters from within the Control Room for greater than or equal to 15 min. (Note 1)ANDAny significant transient is in progress, Table S-3Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Table S-2 Safety System Parameters* Reactor power* RCS level* RCS pressure" Core Exit T/C temperature* Level in at least one SG* Auxiliary or emergency feed flow inat least one SGTable S-3 Significant Transients" Reactor trip* Runback greater than or equal to25% thermal power* Electrical load rejection greater than25% electrical load* ECCS actuationMode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):1JNPLANNED -A parameter change or an event that is not 1) the result of an intendedtolution or 2) an expected plant response to a transient. The cause of the parameter change93vent may be known or unknown.Page 190 of 276 1 ATTACHMENT 1EAL BasesCPNPP Basis:SAFETY SYSTEM parameters listed in Table S-1 are monitored in the Control Room througha combination of hard control panel indicators as well as computer based information systems.The Plant Computer, which displays SPDS required information, serves as a redundantcompensatory indicator which may be utilized in lieu of normal Control Room indicators (ref. 1,2,3,4).Significant transients are listed in Table S-2 and include response to automatic or manuallyinitiated functions such as reactor trips, runbacks involving greater than or equal to 25%thermal power change, electrical load rejections of greater than 25% full electrical load orECCS (SI) injection actuations.NEI 99-01 Basis:This IC addresses the difficulty associated with monitoring rapidly changing plant conditionsduring a transient without the ability to obtain SAFETY SYSTEM parameters from within theControl Room. During this condition, the margin to a potential fission product barrier challengeis reduced. It thus represents a potential substantial degradation in the level of safety of theplant.As used in this EAL, an "inability to monitor" means that values for one or more of the listedparameters cannot be determined from within the Control Room. This situation would requirea loss of all of the Control Room sources for the given parameter(s). For example, the reactorpower level cannot be determined from any analog, digital and recorder source within theControl Room.An event involving a loss of plant indications, annunciators and/or display systems is evaluatedin accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine ifan NRC event report is required. The event would be reported if it significantly impaired thecapability to perform emergency assessments. In particular, emergency assessmentsnecessary to implement abnormal operating procedures, emergency operating procedures,and emergency plan implementing procedures addressing emergency classification, accidentassessment, or protective action decision-making.This EAL is focused on a selected subset of plant parameters associated with the key safetyfunctions of reactivity control, core cooling [PuD4R] / RPV lev.,el [BWR] and RCS heat removal.The loss of the ability to determine one or more of these parameters from within the ControlRoom is considered to be more significant than simply a reportable condition. In addition, if allindication sources for one or more of the listed parameters are lost, then the ability todetermine the values of other SAFETY SYSTEM parameters may be impacted as well. Forexample, if the value for reactor vessel level [PWR] i RPV water level [.1r314RI cannot bedetermined from the indications and recorders on a main control board, the SPDS or the plantcomputer, the availability of other parameter values may be compromised as well.Fifteen minutes was selected as a threshold to exclude transient or momentary losses ofindication.Escalation of the emergency classification level would be via ICs FS1 or IC AS-I-RS1Page 191 of 276 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):1.2.3.4.5.FSAR Section 7.5DBD-EE-033 Detailed Control Room Design, 5.1.2, Figure 1SOP 906 Plant Process Computer System GuidelinesABN 906 Plant Process Computer System MalfunctionNEI 99-01 SA2Page 192 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:EAL:S -System Malfunction4 -RCS ActivityReactor coolant activity greater than Technical Specification allowablelimitsSU4.1 Unusual EventReactor coolant Dose Equivalent 1-131 specific activity greater than 60 pCi/gmORReactor coolant Dose Equivalent XE-133 specific activity greater than 500 pCi/gmMode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):NoneCPNPP Basis:This EAL addresses reactor coolant samples exceeding Technical Specification LCOs3.4.16.A and 3.4.16.B which are applicable in Modes 1, 2, and 3 and 4 (ref. 1). The TechnicalSpecification limits accommodate an iodine spike phenomenon that may occur followingchanges in thermal power. The Technical Specification LCO limits are established to minimizethe offsite radioactivity dose consequences in the event of a steam generator tube rupture(SGTR) accident (ref. 2).NEI 99-01 Basis:This IC addresses a reactor coolant activity value that exceeds an allowable limit specified inTechnical Specifications. This condition is a precursor to a more significant event andrepresents a potential degradation of the level of safety of the plant.Escalation of the emergency classification level would be via ICs FA1 or the RecognitionCategory A-R ICs.CPNPP Basis Reference(s):1. Technical Specifications Section 3.4.162. Technical Specifications Section B3.4.163. NEI 99-01 SU3Page 193 of 276 ATTACHMENT IEAL BasesCategory: S -System MalfunctionSubcategory: 4 -RCS ActivityInitiating Condition: Reactor coolant activity greater than Technical Specification allowablelimitsEAL:SU4.2 Unusual EventGross Failed Fuel Monitor, FFLu60 (u-RE-0406), High Alarm (RED)Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):NoneCPNPP Basis:This EAL addresses reactor coolant letdown line radiation levels sensed by FFLu60 (u-RE-0406) in excess of Technical Specification allowable limits. The High Alarm (RED) setpoint isbased on the Technical Specifications maximum allowable concentration of radioactivity in thereactor coolant (ref. 1, 2, 3). A Geiger-Mueller tube is mounted on the reactor coolant letdownline after the letdown heat exchanger to monitor fission-product activity. Detection of increasedsystem activity may be indicative of failed fuel. The monitor initiates Alert and High alarms inthe Control Room (PC-1 1 and Plant Computer) (ref. 3, 4, 5, 6, 7, 8).FFLu60 (u-RE-0406) has a range of 1 E-2 -1 E+7 pCi/ml.NEI 99-01 Basis:This IC addresses a reactor coolant activity value that exceeds an allowable limit specified inTechnical Specifications. This condition is a precursor to a more significant event andrepresents a potential degradation of the level of safety of the plant.Escalation of the emergency classification level would be via ICs FAI or the RecognitionCategory A-R ICs.CPNPP Basis Reference(s):1. Technical Specifications Section 3.4.162. ALM-3200 Alarm Procedure DRMS, Channel in High Alarm (RED), pg 543. DBD-EE-023 Radiation Monitoring System4. SWEC-NU(S)-174 Radiation Monitor Alarm Concentrations for Failed Fuel Monitors 1-RE-406 & 2-RE-4065. ABN-102 High Coolant Activity6. FSAR Section 11.5.2.7.117. FSAR Table 11.5-18. CHM-1 11 Primary Chemistry Accident Assessment Sampling ProgramPage 194 of 276 ATTACHMENT 1EAL Bases9. DBD-EE-023 Radiation Monitoring System10.NEI 99-01 SU3I Page 195 of 276 ATTACHMENT 1EAL BasesCategory: S -System MalfunctionSubcategory: 5 -RCS LeakageInitiating Condition: RCS leakage for 15 minutes or longerEAL:SU5.1 Unusual EventRCS unidentified or pressure boundary leakage greater than 10 gpm for greater than orequal to 15 min.ORRCS identified leakage greater than 25 gpm for greater than or equal to 15 min.ORUNISOLABLE leakage from the RCS to a location outside containment greater than 25gpm for greater than or equal to 15 min.(Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):NoneCPNPP Basis:RCS leakage outside of the containment that is not considered identified or unidentifiedleakage per Technical Specifications includes leakage via interfacing systems such as RCS tothe Component Cooling Water, or systems that directly see RCS pressure outside containmentsuch as Chemical & Volume Control System, Nuclear Sampling system and Residual HeatRemoval system (when in the shutdown cooling mode) (ref. 3, 6, 8)Isolating letdown is a standard abnormal operating procedure action and may preventunnecessary classification when a non-RCS leakage path, such as a CVCS leak, exists.Unidentified leakage and identified leakage are determined by performance of the RCS waterinventory balance. Pressure boundary leakage would first appear as unidentified leakage andcan only be positively identified by inspection (ref. 1). OPT-303 (ref. 1) is used to ensure RCSleakage is within Technical Specification limits (ref. 2). ABN-103 Attachments 1 and 3 (ref. 3)are used for excessive RCS leakage.Technical Specifications (ref. 4) defines RCS leakage as follows:Identified Leakage:o Leakage such as that from pump seals or valve packing (except reactor coolant pump(RCP) seal water injection or leakoff), that is captured and conducted to collectionsystems or a sump or collecting tankPage 196 of 276 ATTACHMENT 1EAL Baseso Leakage into the Containment atmosphere from sources that are both specificallylocated and known either not to interfere with the operation of leakage detectionsystems or not to be pressure boundary leakage.o Reactor Coolant System leakage through a steam generator to the Secondary System(primary to secondary leakage);" Unidentified Leakage: All leakage (except RCP seal water injection or leakoff) that is notidentified leakage." Pressure Boundary Leakage: Leakage (except primary to secondary leakage) through anon-isolable fault in an RCS component body, pipe wall, or vessel wall.Escalation of this EAL to the Alert level is via Category F, Fission Product Barrier Degradation,EAL FA1.1.NEI 99-01 Basis:This IC addresses RCS leakage which may be a precursor to a more significant event. In thiscase, RCS leakage has been detected and operators, following applicable procedures, havebeen unable to promptly isolate the leak. This condition is considered to be a potentialdegradation of the level of safety of the plant..EAL ..andE.AL.2The first and second EAL conditions are focused on a loss of mass fromthe RCS due to "unidentified leakage", "pressure boundary leakage" or "identified leakage" (asthese leakage types are defined in the plant Technical Specifications). EAL-#3The thirdcondition addresses an RCS mass loss caused by an UNISOLABLE leak through aninterfacing system. These EAL=s-conditions thus apply to leakage into the containment, asecondary-side system (e.g., steam generator tube leakage iO a PWR) or a location outside ofcontainment.The leak rate values for each E-AL-condition were selected because they are usuallyobservable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). EAL4#!-The firstcondition uses a lower value that reflects the greater significance of unidentified or pressureboundary leakage.The release of mass from the RCS due to the as-designed/expected operation of a relief valvedoes not warrant an emergency classification. FGF PWRS-,aAn emergency classification wouldbe required if a mass loss is caused by a relief valve that is not functioning asdesigned/expected (e.g., a relief valve sticks open and the line flow cannot be isolated). FGFBWRs,-a- stuck open Safety Relief Valve (SRV) oe SRV leakage is not on'sidered eitheridentified or unidentified leakage by Technical Specifications and, therefore, is not applicablete th~sFA--AThe 15-minute threshold duration allows sufficient time for prompt operator actions to isolatethe leakage, if possible.Escalation of the emergency classification level would be via ICs of Recognition Category A-Ror F.Page 197 of 276 ATTACHMENT 1EAL BasesCPNPP Basis Reference(s):I.2.3.4.5.6.7.8.9.OPT-303 Reactor Coolant System Water InventoryTechnical Specifications 3.4.13ABN-103 Excessive Reactor Coolant LeakageTechnical Specifications 1.1ABN-108 Shutdown Loss of CoolantFSAR 5.2.5.2FSAR 5.2.5.8ECA-1.2 LOCA Outside ContainmentNEI 99-01 SU4Page 198 of 276 ATTACHMENT 1EAL BasesCategory: S -System MalfunctionSubcategory: 6 -RPS FailureInitiating Condition: Automatic or manual trip fails to shut down the reactorEAL:SU6.1 Unusual EventAn automatic trip did not shut down the reactor as indicated by reactor power greater than5% after any RPS setpoint is exceededANDA subsequent automatic trip or manual trip action taken at the reactor control consoles(MCB reactor trip switches or deenergizing uB3 and uB4) is successful in shutting downthe reactor as indicated by reactor power less than or equal to 5% (Note 8)Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidlyinserted into the core, and does not include manually driving in control rods or implementation of boroninjection strategies.Mode Applicability:1 -Power OperationDefinition(s):NoneCPNPP Basis:The first condition of this EAL identifies the need to cease critical reactor operations byactuation of the automatic Reactor Protection System (RPS) trip function. A reactor trip isautomatically initiated by the RPS when certain continuously monitored parameters exceedpredetermined setpoints (ref. 1).Following a successful reactor trip, rapid insertion of the control rods occurs. Nuclear powerpromptly drops to a fraction of the original power level and then decays to a level severaldecades less with a negative startup rate. The reactor power drop continues until reactorpower reaches the point at which the influence of source neutrons on reactor power starts tobe observable. A predictable post-trip response from an automatic reactor trip signal shouldtherefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentationand a lowering of power into the source range. A successful trip has therefore occurred whenthere is sufficient rod insertion from the trip of RPS to bring the reactor power below theimmediate shutdown decay heat level of 5% (ref. 1, 2).For the purposes of emergency classification, successful manual trip actions are thosewhich can be quickly performed from the reactor control console; MCB reactor tripswitches or deenergizing uB3 and uB4. Reactor shutdown achieved by use of other tripactions specified in FR-S.1 Response to Nuclear Power Generation/ATWS (such as manuallyinsert control rods, opening the reactor trip and bypass breakers in the reactor switchgear,tripping the Rod Drive MG sets in the normal switchgear or emergency boration) do notconstitute a successful manual trip (ref. 2).Page 199 of 276 ATTACHMENT 1EAL BasesFollowing any automatic RPS trip signal, E-0.0 (ref. 1) and /FR-S.1 (ref. 2) prescribe insertionof redundant manual trip signals to back up the automatic RPS trip function and ensure reactorshutdown is achieved. Even if the first subsequent manual trip signal inserts all control rods tothe full-in position immediately after the initial failure of the automatic trip, the lowest level ofclassification that must be declared is an Unusual Event (ref. 2).In the event that the operator identifies a reactor trip is imminent and initiates a successfulmanual reactor trip before the automatic RPS trip setpoint is reached, no declaration isrequired. The successful manual trip of the reactor before it reaches its automatic trip setpointor reactor trip signals caused by instrumentation channel failures do not lead to a potentialfission product barrier loss. However, if subsequent manual reactor trip actions fail to reducereactor power to or below 5%, the event escalates to the Alert under EAL SA6.1.If by procedure, operator actions include the initiation of an immediate manual trip followingreceipt of an automatic trip signal and there are no clear indications that the automatic tripfailed (such as a time delay following indications that a trip setpoint was exceeded), it may bedifficult to determine if the reactor was shut down because of automatic trip or manual actions.If a subsequent review of the trip actuation indications reveals that the automatic trip did notcause the reactor to be shut down, then consideration should be given to evaluating the fuelfor potential damage, and the reporting requirements of 50.72 should be considered for thetransient event.NEI 99-01 Basis:This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor(trip [PrI, / Scram [81/14R]) that results in a reactor shutdown, and either a subsequentoperator manual action taken at the reactor control consoles or an automatic (trip fLrWR]--.scram r[{4R4) is successful in shutting down the reactor. This event is a precursor to a moresignificant condition and thus represents a potential degradation of the level of safety of theplant.I Following the failure on an automatic reactor (trip [PWR] I scram [BI',1.), operators willpromptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g.,I initiate a manual reactor (trip [P,44R] / sc,,ram [,WR.)). If these manual actions are successfulin shutting down the reactor, core heat generation will quickly fall to a level within thecapabilities of the plant's decay heat removal systems.If an initial manual reactor (trip [P-KR], / Gscam [,WRI) is unsuccessful, operators will promptlytake manual action at another location(s) on the reactor control consoles to shutdown thereactor (e.g., initiate a manual reactor (trip [PrW1RI / Scram [B"l'R])) using a different switch).Depending upon several factors, the initial or subsequent effort to manually (trip / sGrim[BW-R}) the reactor, or a concurrent plant condition, may lead to the generation of an automaticreactor (trip [PWR] / scram signal. If a subsequent manual or automatic (trip .PW.. ,sra is successful in shutting down the reactor, core heat generation will quickly fallto a level within the capabilities of the plant's decay heat removal systems.A manual action at the reactor control consoles is any operator action, or set of actions, whichcauses the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor(trip[PWR] / scram r],,. This action does not include manually driving in control rods orimplementation of boron injection strategies. Actions taken at back-panels or other locationsPage 200 of 276 ATTACHMENT 1EAL Baseswithin the Control Room, or any location outside the Control Room, are not considered to be"at the reactor control consoles".Taking the RoactoFr. Mode Sw~itcrh to61 SHT-DOWN is considered to be a mianual scram tion-.[BWR]The plant response to the failure of an automatic or manual reactor (trip [P,.R] / cram. [BVrR])will vary based upon several factors including the reactor power level prior to the event,availability of the condenser, performance of mitigation equipment and actions, otherconcurrent plant conditions, etc. If subsequent operator manual actions taken at the reactorcontrol consoles are also unsuccessful in shutting down the reactor, then the emergencyclassification level will escalate to an Alert via IC SA5SA6. Depending upon the plantresponse, escalation is also possible via IC FA1. Absent the plant conditions needed to meeteither IC SA5-SA6 or FA1, an Unusual Event declaration is appropriate for this event.A reactor shutdown is determined in accordance with applicable Emergency OperatingProcedure criteria.Should a reactor (trip [PV'RJ / [.. "R]) signal be generated as a result of plant work(e.g., RPS setpoint testing), the following classification guidance should be applied.* If the signal causes a plant transient that should have included an automatic reactor (trip[PWVR] / scram [BWR]) and the RPS fails to automatically shutdown the reactor, thenthis IC and the EALs are applicable, and should be evaluated.* If the signal does not cause a plant transient and the (trip [PWRI / scrGam [B,-R]) failureis determined through other means (e.g., assessment of test results), then this IC andthe EALs are not applicable and no classification is warranted.CPNPP Basis Reference(s):1. EOP-O.OA/B Reactor Trip or Safety Injection2. FR-S.1 Response to Nuclear Power Generation/ATWS3 NEI 99-01 SU5Page 201 of 276 ATTACHMENT IEAL BasesCategory: S -System MalfunctionSubcategory: 6 -RPS FailureInitiating Condition: Automatic or manual trip fails to shut down the reactorEAL:SU6.2 Unusual EventA manual trip did not shut down the reactor as indicated by reactor power greater than 5%after any manual trip action was initiatedANDA subsequent automatic trip or manual trip action taken at the reactor control console(MCB reactor trip switches or deenergizing uB3 and uB4) is successful in shutting downthe reactor as indicated by reactor power less than or equal to 5% (Note 8)Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidlyinserted into the core, and does not include manually driving in control rods or implementation of boroninjection strategies.Mode Applicability:1 -Power OperationDefinition(s):NoneCPNPP Basis:This EAL addresses a failure of a manually initiated trip in the absence of having exceeded anautomatic RTS trip setpoint and a subsequent automatic or manual trip is successful inshutting down the reactor (reactor power less than or equal to 5%). (ref. 1).Following a successful reactor trip, rapid insertion of the control rods occurs. Nuclear powerpromptly drops to a fraction of the original power level and then decays to a level severaldecades less with a negative startup rate. The reactor power drop continues until reactorpower reaches the point at which the influence of source neutrons on reactor power starts tobe observable. A predictable post-trip response from an automatic reactor trip signal shouldtherefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentationand a lowering of power into the source range. A successful trip has therefore occurred whenthere is sufficient rod insertion from the trip of RPS to bring the reactor power below theimmediate shutdown decay heat level of 5% (ref. 1, 2).For the purposes of emergency classification, successful manual trip actions are thosewhich can be quickly performed from the reactor control console; MCB reactor tripswitches or deenergizing uB3 and uB4. Reactor shutdown achieved by use of other tripactions specified in FR-S.1 Response to Nuclear Power Generation/ATWS (such as manuallyinsert control rods, opening the reactor trip and bypass breakers in the reactor switchgear,tripping the Rod Drive MG sets in the normal switchgear or emergency boration) do notconstitute a successful manual trip (ref. 2).Page 202 of 276 ATTACHMENT 1EAL BasesFollowing the failure of any manual trip signal, E-0.0 (ref. 1) and FR-S.1 (ref. 2) prescribeinsertion of redundant manual trip signals to back up the RPS trip function and ensure reactorshutdown is achieved. Even if a subsequent automatic trip signal or the first subsequentmanual trip signal inserts all control rods to the full-in position immediately after the initialfailure of the manual trip, the lowest level of classification that must be declared is an UnusualEvent (ref. 2).If both subsequent automatic and subsequent manual reactor trip actions in the Control Roomfail to reduce reactor power below the power associated with the safety system design (lessthan or equal to 5%) following a failure of an initial manual trip, the event escalates to an Alertunder EAL SA6.1.NEI 99-01 Basis:This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor(trip [PWR] / scram [BAR]) that results in a reactor shutdown, and either a subsequentoperator manual action taken at the reactor control consoles or an automatic (trip [PWk]-...am- is successful in shutting down the reactor. This event is a precursor to a moresignificant condition and thus represents a potential degradation of the level of safety of theplant.I Following the failure on an automatic reactor (trip [P,,rIq i scram operators willpromptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g.,I initiate a manual reactor (trip [PR] i scram [,WR])). If these manual actions are successfulin shutting down the reactor, core heat generation will quickly fall to a level within thecapabilities of the plant's decay heat removal systems.If an initial manual reactor (trip [PrK, R, scram [B",,q,) is unsuccessful, operators will promptlytake manual action at another location(s) on the reactor control consoles to shutdown thereactor (e.g., initiate a manual reactor (trip [Pr], / scram [Bl-r)) using a different switch).Depending upon several factors, the initial or subsequent effort to manually [P,/] / scram[BW-R) the reactor, or a concurrent plant condition, may lead to the generation of an automaticreactor (trip [PKR] i scram- [BIA'J) signal. If a subsequent manual or automatic (trip fP...scam is successful in shutting down the reactor, core heat generation will quickly fallto a level within the capabilities of the plant's decay heat removal systems.A manual action at the reactor control consoles is any operator action, or set of actions, whichcauses the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor#trip[PW'R]K ScGram ). This action does not include manually driving in control rods orimplementation of boron injection strategies. Actions taken at back-panels or other locationswithin the Control Room, or any location outside the Control Room, are not considered to be"at the reactor control consoles".Taking the Roactor Mode Switch to SHUTDOWN is considered to be a manual scram action-.fB9WRIThe plant response to the failure of an automatic or manual reactor (trip [PWR' / scram [BWR])will vary based upon several factors including the reactor power level prior to the event,availability of the condenser, performance of mitigation equipment and actions, otherconcurrent plant conditions, etc. If subsequent operator manual actions taken at the reactorcontrol consoles are also unsuccessful in shutting down the reactor, then the emergencyclassification level will escalate to an Alert via IC SA5SA6. Depending upon the plantPage 203 of 276 1 ATTACHMENT 1EAL Basesresponse, escalation is also possible via [C FAI. Absent the plant conditions needed to meeteither IC SA5-SA6 or FA1, an Unusual Event declaration is appropriate for this event.A reactor shutdown is determined in accordance with applicable Emergency OperatingProcedure criteria.I Should a reactor (trip [P"WR] ! S...m [B.. R]) signal be generated as a result of plant work(e.g., RPS setpoint testing), the following classification guidance should be applied.* If the signal causes a plant transient that should have included an automatic reactor (trip[PWR] [B,,-R]) and the RPS fails to automatically shutdown the reactor, thenthis IC and the EALs are applicable, and should be evaluated.e If the signal does not cause a plant transient and the (trip [PWR] / [BWR]) failureis determined through other means (e.g., assessment of test results), then this IC andthe EALs are not applicable and no classification is warranted.CPNPP Basis Reference(s):1.2.3.EOP-O.OA/B Reactor Trip or Safety InjectionFR-S.1 Response to Nuclear Power Generation/ATWSNEI 99-01 SU5Page 204 of 276 ATTACHMENT IEAL BasesCategory:Subcategory:Initiating Condition:S -System Malfunction2 -RPS FailureAutomatic or manual trip fails to shut down the reactor and subsequentmanual actions taken at the reactor control consoles are not successfulin shutting down the reactorEAL:SA6.1 AlertAn automatic or manual trip fails to shut down the reactor as indicated by reactor powergreater than 5%ANDManual trip actions taken at the reactor control console (MCB reactor trip switches ordeenergizing uB3 and uB4) are not successful in shutting down the reactor as indicated byreactor power greater than 5% (Note 8)Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidlyinserted into the core, and does not include manually driving in control rods or implementation of boroninjection strategies.Mode Applicability:1 -Power OperationDefinition(s):NoneCPNPP Basis:This EAL addresses any automatic or manual reactor trip signal that fails to shut down thereactor (reactor power less than or equal to 5%) followed by a subsequent manual trip that failsto shut down the reactor to an extent the reactor is producing energy in excess of the heat loadfor which the safety systems were designed (ref. 1).For the purposes of emergency classification, successful manual trip actions are thosewhich can be quickly performed from the reactor control console; MCB reactor tripswitches or deenergizing uB3 and uB4. Reactor shutdown achieved by use of other tripactions specified in FR-S.1 Response to Nuclear Power Generation/ATWS (such as manuallyinsert control rods, opening the reactor trip and bypass breakers in the reactor switchgear,tripping the Rod Drive MG sets in the normal switchgear or emergency boration) do notconstitute a successful manual trip (ref. 2).5% rated power is a minimum reading on the power range scale that indicates continuedpower production. It also approximates the decay heat which the shutdown systems weredesigned to remove and is indicative of a condition requiring immediate response to preventsubsequent core damage. Below 5%, plant response will be similar to that observed during anormal shutdown. Nuclear instrumentation can be used to determine if reactor power is greaterthan 5 % power (ref. 1, 2).Page 205 of 276 ATTACHMENT IEAL BasesEscalation of this event to a Site Area Emergency would be under EAL SS6.1 or EmergencyCoordinator judgment.NEI 99-01 Basis:This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor(trip [PWR] / Scram. [,LR]) that results in a reactor shutdown, and subsequent operatormanual actions taken at the reactor control consoles to shutdown the reactor are alsounsuccessful. This condition represents an actual or potential substantial degradation of thelevel of safety of the plant. An emergency declaration is required even if the reactor issubsequently shutdown by an action taken away from the reactor control consoles since thisevent entails a significant failure of the RPS.A manual action at the reactor control console is any operator action, or set of actions, whichcauses the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor(trip[PWR] / scram [9B4WR)). This action does not include manually driving in control rods orimplementation of boron injection strategies. If this action(s) is unsuccessful, operators wouldimmediately pursue additional manual actions at locations away from the reactor controlconsoles (e.g., locally opening breakers). Actions taken at back-panels or other locationswithin the Control Room, or any location outside the Control Room, are not considered to be"at the reactor control consoles".Taking the Mode SWitch to SHUTDOWN i,considered to be a manual scram action. [BWR]The plant response to the failure of an automatic or manual reactor (trip [PVVRD / scram [BrI/)will vary based upon several factors including the reactor power level prior to the event,availability of the condenser, performance of mitigation equipment and actions, otherconcurrent plant conditions, etc. If the failure to shut-down the reactor is prolonged enough tocause a challenge to the core cooling [PWR] / PV ..ate. level [BWR] or RCS heat removalsafety functions, the emergency classification level will escalate to a Site Area Emergency viaIC SS65. Depending upon plant responses and symptoms, escalation is also possible via ICFSI. Absent the plant conditions needed to meet either IC SS65 or FS1, an Alert declarationis appropriate for this event.It is recognized that plant responses or symptoms may also require an Alert declaration inaccordance with the Recognition Category F ICs; however, this IC and EAL are included toensure a timely emergency declaration.A reactor shutdown is determined in accordance with applicable Emergency OperatingProcedure criteria.CPNPP Basis Reference(s):1. EOP-O.OA/B Reactor Trip or Safety Injection2. FR-S.1 Response to Nuclear Power Generation/ATWS3. NEI 99-01 SA5Page 206 of 276 ATTACHMENT 1EAL BasesCategory:Subcategory:Initiating Condition:EAL:S -System Malfunction2 -RPS FailureInability to shut down the reactor causing a challenge to core cooling orRCS heat removalSS6.1 Site Area EmergencyAn automatic or manual trip fails to shut down the reactor as indicated by reactor powergreater than 5%ANDAll actions to shut down the reactor are not successful as indicated by reactor powergreater than 5%AND EITHER:* CSFST Core Cooling RED Path conditions met" CSFST Heat Sink RED Path conditions metMode Applicability:1 -Power OperationDefinition(s):NoneCPNPP Basis:This EAL addresses the following:* Any automatic reactor trip signal followed by a manual trip that fails to shut down thereactor to an extent the reactor is producing energy in excess of the heat load for whichthe safety systems were designed (EAL SA6.1), and" Indications that either core cooling is extremely challenged or heat removal is extremelychallenged.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 direct threatto the Fuel Clad and RCS barriers.Reactor shutdown achieved by use of FR-S.1 Response to Nuclear Power Generation/ATWS(such as manually insert control rods, opening the reactor trip and bypass breakers in thereactor switchgear, tripping the Rod Drive MG sets in the normal switchgear or emergencyboration) are also credited as a successful manual trip provided reactor power can be reducedbelow 5% before indications of an extreme challenge to either core cooling or heat removalexist (ref. 1, 2).5% rated power is a minimum reading on the power range scale that indicates continuedpower production. It also approximates the decay heat which the shutdown systems weredesigned to remove and is indicative of a condition requiring immediate response to preventPage 207 of 276 ATTACHMENT 1EAL Basessubsequent core damage. Below 5%, plant response will be similar to that observed during anormal shutdown. Nuclear instrumentation can be used to determine if reactor power is greaterthan 5% power (ref. 1, 2).Indication of continuing core cooling degradation is manifested by CSFST Core Cooling REDPath conditions being met. Specifically, Core Cooling RED Path conditions exist if either coreexit T/Cs are reading greater than or equal to 12001F (ref. 3).Indication of inability to adequately remove heat from the RCS is manifested by CSFST HeatSink RED Path conditions being met. Specifically, Heat Sink RED Path conditions exist ifnarrow range level in at least one steam generator is not greater than or equal to (43[50]%ACC) on Unit 1 or (10 [18]% ACC) on Unit 2 and total feedwater flow to the steam generatorsis less than or equal to 460 gpm (ref. 4).NEI 99-01 Basis:This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor(trip [P] ,'/ .cram [B,,R) that results in a reactor shutdown, all subsequent operator actionsto manually shutdown the reactor are unsuccessful, and continued power generation ischallenging the capability to adequately remove heat from the core and/or the RCS. Thiscondition will lead to fuel damage if additional mitigation actions are unsuccessful and thuswarrants the declaration of a Site Area Emergency.In some instances, the emergency classification resulting from this IC/EAL may be higher thanthat resulting from an assessment of the plant responses and symptoms against theRecognition Category F ICs/EALs. This is appropriate in that the Recognition Category FICs/EALs do not address the additional threat posed by a failure to shut-down the reactor. Theinclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency inresponse to prolonged failure to shutdown the reactor.A reactor shutdown is determined in accordance with applicable Emergency OperatingProcedure criteria.Escalation of the emergency classification level would be via IC AG--RG1 or FG1.CPNPP Basis Reference(s):1. EOP-0.OA/B Reactor Trip or Safety Injection2. FR-S.1 Response to Nuclear Power Generation/ATWS3. FR-C.1 Response to Inadequate Core Cooling4. FR-H.1 Response to Loss of Heat Sink5. NEI 99-01 SS5.I Page 208 of 276 ATTACHMENT 1EAL BasesCategory: S -System MalfunctionSubcategory: 7 -Loss of CommunicationsInitiating Condition: Loss of all onsite or offsite communications capabilitiesEAL:SU7.1 Unusual EventLoss of all Table S-4 onsite communication methodsORLoss of all Table S-4 offsite communication methodsORLoss of all Table S-4 NRC communication methodsTable S-4 Communication MethodsSystem Onsite Offsite NRCGai-tronics Page/Party (PA) XPlant Radios XPABX X X XPublic Telephone X X XFederal Telephone System (FTS) X XMode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):NoneCPNPP Basis:Onsite/offsite communications include one or more of the systems listed in Table S-4 (ref. 1,2).This EAL is the hot condition equivalent of the cold condition EAL CU5.1.NEI 99-01 Basis:This IC addresses a significant loss of on-site or offsite communications capabilities. While nota direct challenge to plant or personnel safety, this event warrants prompt notifications toOROs and the NRC.Page 209 of 276 ATTACHMENT 1EAL BasesThis IC should be assessed only when extraordinary means are being utilized to makecommunications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent tooffsite locations, etc.).EAL#4-The first EAL condition addresses a total loss of the communications methods used insupport of routine plant operations.EAL-#2The second EAL condition addresses a total loss of the communications methods usedto notify all OROs of an emergency declaration. The offsite (OROs. referred to here are-(seeDeveloper Notes) the State Department of Public Safety, Somervell and Hood County EOCs-EAL4#3The third EAL addresses a total loss of the communications methods used to notify theNRC of an emergency declaration.CPNPP Basis Reference(s):1. FSAR 9.5.22. DBD-EE-048 Communication System3. NEI 99-01 SU6I Page 210 of 276 ATTACHMENT 1EAL BasesCategory: S -System MalfunctionSubcategory: 8 -Containment FailureInitiating Condition: Failure to isolate containment or loss of containment pressure control.EAL:SU8.1 Unusual EventAny penetration is not isolated within 15 min. of a VALID containment isolation signalORContainment pressure greater than 18 psig with neither Containment Spray systemoperating per design for greater than or equal to 15 min.(Note 1)Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit hasbeen exceeded, or will likely be exceeded.Mode Applicability:I -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):VALID -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) by directobservation by plant personnel, such that doubt related to the indicator's operability, thecondition's existence, or the report's accuracy is removed. Implicit in this definition is the needfor timely assessment.CPNPP Basis:The Containment Spray System (CSS) is designed to remove heat from the Containmentenvironment following a LOCA, a main steam line break accident, or a feedwater line breakaccident. Each unit of the CPNPP is equipped with two redundant Containment spray trains,each designed to provide emergency Containment heat removal in the event of a LOCA. Thissystem, in conjunction with the ECCS, removes post-accident thermal energy from theContainment environment, thereby reducing the Containment pressure and temperature. Eachtrain includes two containment spray pumps, spray headers, nozzles, valves, and piping. Eachtrain is powered from a separate safeguard bus. (ref. 1)The Containment pressure setpoint (18 psig, ref. 2) is the pressure at which the ContainmentSpray System should actuate and begin performing its function. The design basis accident.analyses and evaluations assume the loss of one Containment Spray System train (ref. 1).NEI 99-01 Basis:This IC,-EAL addresses a failure of one or more containment penetrations to automaticallyisolate (close) when required by an actuation signal. It also addresses an event that results inhigh containment pressure with a concurrent failure of containment pressure control systems.Absent challenges to another fission product barrier, either condition represents potentialdegradation of the level of safety of the plant.Page 211 of 276 ATTACHMENT 1EAL BasesI For EAL--#-the first condition, the containment isolation signal must be generated as the resulton an off-normal/accident condition (e.g., a safety injection or high containment pressure); afailure resulting from testing or maintenance does not warrant classification. Thedetermination of containment and penetration status -isolated or not isolated -should bemade in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The15-minute criterion is included to allow operators time to manually isolate the requiredpenetrations, if possible.EAL #2The second condition addresses a condition where containment pressure is greaterthan the setpoint at which containment energy (heat) removal systems are designed toautomatically actuate, and less than one full train of equipment is capable of operating perdesign. The 15-minute criterion is included to allow operators time to manually start equipmentthat may not have automatically started, if possible. The inability to start the requiredequipment indicates that containment heat removal/depressurization systems (e.g.,containment sprays or ice condenser fans) are either lost or performing in a degraded manner.This event would escalate to a Site Area Emergency in accordance with IC FS1 if there were aconcurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers.CPNPP Basis Reference(s):1. FSAR Section 6.2.22. FRC-Z.1A/B Response to High Containment Pressure3. NEI 99-01 SU7IPage 212 of 276 ATTACHMENT IEAL BasesCategory: S -System MalfunctionSubcategory: 9 -Hazardous Event Affecting Safety SystemsInitiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the currentoperating modeEAL:SA9.1 AlertThe occurrence of any Table S-5 hazardous eventAND EITHER:* Event damage has caused indications of degraded performance in at least one trainof a SAFETY SYSTEM needed for the current operating mode" The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component orstructure needed for the current operating modeTable S-5 Hazardous Events" Seismic event (earthquake)" Internal or external FLOODING event* High winds or tornado strike" FIRE" EXPLOSION" Other events with similar hazard characteristicsas determined by the Emergency CoordinatorMode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):EXPLOSION -A rapid, violent and catastrophic failure of a piece of equipment due tocombustion, chemical reaction or overpressurization. A release of steam (from high energylines or components) or an electrical component failure (caused by short circuits, grounding,arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drivebelts or overheated electrical equipment do not constitute fires. Observation of flame ispreferred but is NOT required if large quantities of smoke and heat are observed.FLOODING -A condition where water is entering a room or area faster than installedequipment is capable of removal, resulting in a rise of water level within the room or area.SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/orPage 213 of 276 ATTACHMENT 1EAL Basesplacing it in the cold shutdown condition, including the ECCS. These are typically systemsclassified as safety-related (as defined in 10CFR50.2):Those structures, systems and components that are relied upon to remain functional duringand 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.VISIBLE DAMAGE -Damage to a component or structure that is readily observable withoutmeasurements, testing, or analysis. The visual impact of the damage is sufficient to causeconcern regarding the operability or reliability of the affected component or structure.CPNPP Basis:* The significance of seismic events are discussed under EAL HU2.1 (ref. 1)." Internal FLOODING may be caused by events such as component failures, equipmentmisalignment, or outage activity mishaps (ref. 2)." External flooding may be due to high lake level (ref. 3, 4).* Seismic Category I structures are analyzed to withstand a sustained, design wind velocityof at least 80 mph. (ref. 5)." Areas containing functions and systems required for safe shutdown of the plant areidentified by fire area (ref. 6, 7)." An explosion that degrades the performance of a SAFETY SYSTEM train or visiblydamages a SAFETY SYSTEM component or structure would be classified under this EAL.NEI 99-01 Basis:This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or astructure containing SAFETY SYSTEM components, needed for the current operating mode.This condition significantly reduces the margin to a loss or potential loss of a fission productbarrier, and therefore represents an actual or potential substantial degradation of the level ofsafety of the plant.EAL 1.b.!The first condition addresses damage to a SAFETY SYSTEM train that is inservice/operation since indications for it will be readily available. The indications of degradedperformance should be significant enough to cause concern regarding the operability orreliability of the SAFETY SYSTEM train.EAL .b.2The second condition addresses damage to a SAFETY SYSTEM component that isnot in service/operation or readily apparent through indications alone, or to a structurePage 214 of 276 ATTACHMENT 1EAL Basescontaining SAFETY SYSTEM components. Operators will make this determination based onthe totality of available event and damage report information. This is intended to be a briefassessment not requiring lengthy analysis or quantification of the damage.Escalation of the emergency classification level would be via IC FS1 or A&4-RS1.CPNPP Basis Reference(s):1. ABN-907 Acts of Nature2. CPNPP PRA Accident Sequence Analysis "Internal Flooding Sequences"3. FSAR Section 2.4.3.7 Flood Evaluations for Safe Shutdown Impoundment4. DBD-CS-071 Maximum Probable Flood5. FSAR Section 3.3.1.1 Wind Loadings6. CPNPP Fire Protection Report, Section 5.0 "Fire Safe Shutdown Equipment List"7. FSAR Section 7.4 Systems Required for Safe Shutdown8. NEI 99-01 SA9I Page 215 of 276 ATTACHMENT 1EAL BasesCategory F -Fission Product Barrier DegradationEAL Group: Hot Conditions (RCS temperature greater than2000F); 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 that precludesthe release of highly radioactive fission products to the environment. This concept relies onmultiple physical barriers any one of which, if maintained intact, precludes the release ofsignificant amounts of radioactive fission products to the environment. The primary fissionproduct barriers are:A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains thefuel pellets.B. Reactor Coolant System (RCS): The RCS Barrier includes the RCS primary side and itsconnections up to and including the pressurizer safety and relief valves, and otherconnections up to and including the primary isolation valves.C. Containment (CNTMT): The Containment Barrier includes the containment building andconnections up to and including the outermost containment isolation valves. This barrieralso includes the main steam, feedwater, and blowdown line extensions outside thecontainment building up to and including the outermost secondary side isolation valve.Containment Barrier thresholds are used as criteria for escalation of the ECL from Alertto a Site Area Emergency or a General Emergency.The EALs in this category require evaluation of the loss and potential loss thresholds listed inthe fission product barrier matrix of Table F-1 (Attachment 2). "Loss" and "Potential Loss"signify the relative damage and threat of damage to the barrier. "Loss" means the barrier nolonger assures containment of radioactive materials. "Potential Loss" means integrity of thebarrier is threatened and could be lost if conditions continue to degrade. The number ofbarriers that are lost or potentially lost and the following criteria determine the appropriateemergency classification level: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 barrierThe logic used for emergency classification based on fission product barrier monitoring shouldreflect the following considerations:* The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than theContainment Barrier.* Unusual Event ICs associated with RCS and Fuel Clad Barriers are addressed underSystem Malfunction ICs.Page 216 of 276 ATTACHMENT IEAL Bases* For accident conditions involving a radiological release, evaluation of the fission productbarrier thresholds will need to be performed in conjunction with dose assessments toensure correct and timely escalation of the emergency classification. For example, anevaluation of the fission product barrier thresholds may result in a Site Area Emergencyclassification while a dose assessment may indicate that an EAL for GeneralEmergency IC RG1 has been exceeded.* The fission product barrier thresholds specified within a scheme reflect plant-specificCPNPP design and operating characteristics." As used in this category, the term RCS leakage encompasses not just those typesdefined in Technical Specifications but also includes the loss of RCS mass to anylocation- inside the primary containment, an interfacing system, or outside of theprimary containment. The release of liquid or steam mass from the RCS due to the as-designed/expected operation of a relief valve is not considered to be RCS leakage." At the Site Area Emergency level, EAL users should maintain cognizance of how farpresent conditions are from meeting a threshold that would require a GeneralEmergency declaration. For example, if the Fuel Clad and RCS fission product barrierswere both lost, then there should be frequent assessments of containment radioactiveinventory and integrity. Alternatively, if both the Fuel Clad and RCS fission productbarriers were potentially lost, the Emergency Coordinator would have more assurancethat there was no immediate need to escalate to a General Emergency.I Page 217 of 276 ATTACHMENT 1EAL BasesCategory:Fission Product Barrier DegradationSubcategory: N/AInitiating Condition: Any loss or any potential loss of either Fuel Clad or RCSEAL:FAI.1 AlertAny loss or any potential loss of either Fuel Clad or RCS (Table F-I)Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):NoneCPNPP Basis:Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 (Attachment2) lists the fission product barrier thresholds, bases and references.At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavily than theContainment barrier. Unlike the Containment barrier, loss or potential loss of either the FuelClad or RCS barrier may result in the relocation of radioactive materials or degradation of corecooling capability. Note that the loss or potential loss of Containment barrier in combinationwith loss or potential loss of either Fuel Clad or RCS barrier results in declaration of a SiteArea Emergency under EAL FSI.1NEI 99-01 Basis:NoneCPNPP Basis Reference(s):1. NEI 99-01 FA1I Page 218 of 276 ATTACHMENT 1EAL 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 barriers (Table F-I)Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):NoneCPNPP Basis:Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 (Attachment2) lists the fission product barrier thresholds, bases and references.At the Site Area Emergency classification level, each barrier is weighted equally. A Site AreaEmergency 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 the proximityof present conditions with respect to the threshold for a General Emergency is important. Forexample, the existence of Fuel Clad and RCS Barrier loss thresholds in addition to offsite doseassessments would require continual assessments of radioactive inventory and Containmentintegrity in anticipation of reaching a General Emergency classification. Alternatively, if bothFuel Clad and RCS potential loss thresholds existed, the Emergency Coordinator would havegreater assurance that escalation to a General Emergency is less imminent.NEI 99-01 Basis:NoneCPNPP Basis Reference(s):1. NEI 99-01 FS1Page 219 of 276 ATTACHMENT 1EAL BasesCategory: Fission Product Barrier DegradationSubcategory: N/AInitiating Condition: Loss of any two barriers and loss or potential loss of third barrierEAL:FGI.1 General EmergencyLoss of any two barriersANDLoss or potential loss of third barrier (Table F-I)Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):NoneCPNPP Basis:Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 (Attachment2) 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 barrierNEI 99-01 Basis:NoneCPNPP Basis Reference(s):1. NEI 99-01 FG1Page 220 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesIntroductionTable 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 table isstructured so that each of the three barriers occupies adjacent columns. Each fission productbarrier column is further divided into two columns; one for Loss thresholds and one forPotential 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. RCS or SG Tube LeakageB. Inadequate Heat removalC. CNTMT Radiation / RCS ActivityD. CNTMT Integrity or BypassE. Emergency Coordinator JudgmentEach category occupies a row in Table F-1 thus forming a matrix defined by the categories.The intersection of each row with each Loss/Potential Loss column forms a cell in which one ormore fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for abarrier 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 category titleand number. For example, the first Fuel Clad barrier Loss in Category A would be assigned"FC Loss A.I," the third Containment barrier Potential Loss in Category C would be assigned"CNTMT P-Loss C.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 necessary toexceed 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 the applicablefission product barrier Loss and Potential Loss thresholds. This structure promotes asystematic approach to assessing the classification status of the fission product barriers.When equipped with knowledge of plant conditions related to the fission product barriers, theEAL-user first scans down the category column of Table F-i, locates the likely category andthen reads across the fission product barrier Loss and Potential Loss thresholds in thatcategory to determine if a threshold has been exceeded. If a threshold has not been exceeded,the EAL-user proceeds to the next likely category and continues review of the thresholds in thenew categoryIf the EAL-user determines that any threshold has been exceeded, by definition, the barrier islost or potentially lost -even if multiple thresholds in the same barrier column are exceeded,only that one barrier is lost or potentially lost. The EAL-user must examine each of the threefission product barriers to determine if other barrier thresholds in the category are lost orpotentially lost. For example, if containment radiation is sufficiently high, a Loss of the FuelClad and RCS barriers and a Potential Loss of the Containment barrier can occur. BarrierPage 221 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesLosses and Potential Losses are then applied to the algorithms given in EALs FG1.1, FSI.1,and FA1.1 to determine the appropriate emergency classification.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 222 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesTable F-1 Fission Product Barrier Threshold MatrixFuel Clad (FC) Barrier Reactor Coolant System (RCS) Barrier Containment (CNTMT) BarrierCategory Loss Potential Loss Loss Potential Loss Loss Potential Loss1. An automatic or manual ECCS 1. Operation of a standby chargingA (SI) actuation required by pump is required by EITHER:EITHER:
- UNISOLABLE RCS leakage 1. A leaking or RUPTURED SG is NoneRCS or None None UNISOLABLE RCS
- SG tube leakage FAULTED outside of containmentSG Tube leakage 2. CSFST Integrity-RED PathLeakage SG tube RUPTURE conditions met1. CSFST Core Cooling-ORANGEB Path conditions met 1. CSFST Heat Sink-RED Path 1. CSFST Core Cooling-RED PathInadequate 1. CSFST Core Cooling-RED 2. CSFST Heat Sink-RED Path conditions met conditions metPath conditions met conditions met None AND None ANDHeat AND Heat sink is required Restoration procedures notRemoval Heat sink is required effective within 15 min. (Note 1)RemovalHeat sink is requiredI. Containment radiation greater than85 RJhrCTEuo16 Containment HRRM(L-RE-6290A), or 1. Containment radiation greater than 1. Containment radiation greater thanC CTWu_17 Containment HRRM 5 R/hr 2,110 R/hrCNTMT W-RE-6290B) CTEu16 Containment HRRM CTEu16 Containment HRRMR NT 2. Dose equivalent 1-131 coolant None (L-RE-6290A), or None None (L-RE-6290A), orRadiation activity greater than 300 CTWu17 Containment HRRM CTWu17 Containment HRRMI RCS jCilco (L-RE-6290B) Lu-RE-6290B)Activity 3. Gross Failed Fuel Monitor,FFLu60 (Q-RE-0406),radiation greater than 1.0E04pCi/cc1. Containment isolation isrequiredAND EITHER: 1. CSFST Containment-RED Pathconditions metContainment integrity hasD been lost based on 2. Containment hydrogen concentratiorEmergency Coordinator greater than 4%CNTMT None None None None judgment 3. Containment pressure greater thanIntegrity UNISOLABLE pathwiay from 18 psig With neither Containmentor Bypass Containment to the environment Spray system train operatingexists greater than or equal to 15 rain.2. Indications of RCS leakage (Note 1)outside of ContainmentE 1. Any condition in the opinion of 1. Any condition in the opinion of 1. Any condition in the opinion of 1. Any condition in the opinion of the 1. Any condition in the opinion of 1. Any condition in the opinion of thethe Emergency Coordinator that the Emergency Coordinator that the Emergency Coordinator that Emergency Coordinator that the Emergency Coordinator that Emergency Coordinator thatEC indicates loss of the fuel clad indicates potential loss of the fuel indicates loss of the RCS barrier indicates potential loss of the RCS indicates loss of the Containment indicates potential loss of theJudgment barrier clad barrier barrier barrier Containment barrier[Document No.] Rev. 6 Page 223 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Fuel CladCategory: A. RCS or SG Tube LeakageDegradation Threat: LossThreshold:None[Document No.] Rev. 6 Page 224 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:Category:Degradation Threat:Threshold:Fuel CladA. RCS or SG Tube LeakagePotential LossLNone[[Document No.] Rev. 6 1 Page 225 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesFuel CladBarrier:Category:B. Inadequate Heat RemovalDegradation Threat: LossThreshold:1. CSFST Core Cooling-RED Path conditions metDefinition(s):NoneBasis:Plant-SpecificCritical Safety Function Status Tree (CSFST) Core Cooling-RED Path indicates significantcore exit superheating and core uncovery. The CSFSTs are normally monitored using theSPDS display on the Plant Computer (ref. 1).GenericThis reading indicates temperatures within the core are sufficient to cause significantsuperheating of reactor coolant.CPNPP Basis Reference(s):1. FRC-0.1A/B Response to Inadequate Core Cooling2. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.AS[Document No.] I Rev. 6 1 Page 226 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesFuel CladBarrier:Category:B. Inadequate Heat RemovalDegradation Threat: Potential LossThreshold:1. CSFST Core Cooling-ORANGE Path conditions metDefinition(s):NoneBasis:Plant-SpecificCritical Safety Function Status Tree (CSFST) Core Cooling-ORANGE path indicates indicatessubcooling has been lost and that some fuel clad damage may potentially occur. The CSFSTsare normally monitored using the SPDS display on the Plant Computer (ref. 1, 2).GenericThis reading indicates a reduction in reactor vessel water level sufficient to allow the onset ofheat-induced cladding damage.CPNPP Basis Reference(s):1. FRC-0.1A/B Response to Inadequate Core Cooling2. FRC-0.2A/B Response to Degraded Core Cooling3. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.AI[Document No.] I Rev. 6 Page 227 of 276 1 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Fuel CladCategory: B. Inadequate Heat RemovalDegradation Threat: Potential LossThreshold:2. CSFST Heat Sink-RED Path conditions metANDHeat sink is requiredDefinition(s):NoneBasis:Plant-SpecificIn combination with RCS Potential Loss B.1, meeting this threshold results in a Site AreaEmergency.Critical Safety Function Status Tree (CSFST) Heat Sink-RED Path indicates the ultimate heatsink function is under extreme challenge and that some fuel clad damage may potentiallyoccur (ref. 1).The CSFSTs are normally monitored using the SPDS display on the Plant Computer (ref. 1).The phrase "and heat sink required" precludes the need for classification for conditions inwhich RCS pressure is less than SG pressure or Heat Sink-RED path entry was createdthrough operator action directed by an ERG. For example, FRH-0.1 is entered from CSFSTHeat Sink-Red. Step 1 tells the operator to determine if heat sink is required by checking thatRCS pressure is greater than any non-faulted SG pressure and RCS temperature is greaterthan 3500F. If these conditions exist, Heat Sink is required. Otherwise, the operator is to eitherreturn to the procedure and step in effect and place RHR in service for heat removal. For largeLOCA events inside the Containment, the SGs are moot because heat removal through thecontainment heat removal systems takes place. Therefore, Heat Sink Red should not berequired and, should not be assessed for EAL classification because a LOCA event aloneshould not require higher than an Alert classification. (ref. 1).GenericThis condition indicates an extreme challenge to the ability to remove RCS-heat using thesteam generators (i.e., loss of an effective secondary-side heat sink). This conditionrepresents a potential loss of the Fuel Clad Barrier. In accordance with EOPs, there may beunusual accident conditions during which operators intentionally reduce the heat removalcapability of the steam generators; during these conditions, classification using threshold is notwarranted.I[Document No.] I Rev. 6 1 Page 228 of276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesCPNPP Basis Reference(s):1. FRH-O.1A/B Response to Loss of Secondary Heat Sink2. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.B[Document No.] Rev. 6 Page 229 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Fuel CladCategory: C. CNTMT Radiation / RCS ActivityDegradation Threat: LossThreshold:1. Containment radiation greater than 85 R/hrCTEu16 Containment HRRM (u-RE-6290A), orCTWu17 Containment HRRM (u-RE-6290B)Definition(s):NoneBasis:Plant-SpecificContainment radiation monitor readings greater than 85 R/hr indicate the release of reactorcoolant, with elevated activity indicative of fuel damage, into the Containment. The reading isderived assuming the instantaneous release and dispersal of the reactor coolant noble gasand iodine inventory associated with a 2% clad failures into the Containment atmosphere.Reactor coolant concentrations of this magnitude are several times larger than the maximumconcentrations (including iodine spiking) allowed within technical specifications and aretherefore indicative of fuel damage. This value is higher than that specified for RCS Loss C.1(ref. 2, 3).Per NUS-1 74, the design basis CPNPP RCS specific activity for 1% fuel defects is 340tICi/gm; therefore, a threshold corresponding to 2% fuel clad damage correlates to a coolantactivity of 680 pCi/gm. VL-03-000032 Figure 2A/2B (CRM2) corresponds to approximately50% clad damage released to the containment atmosphere. Figure 2A/2B provides severalpotential limits depending on the pressure of the RCS and the presence of containment spray.The high RCS pressure with containment spray is the most limiting threshold; however, perNEI 99-01, the fuel clad barrier loss threshold should represent a loss of both the fuel clad andRCS barriers. Therefore, the value of curve representing low RCS pressure with spray wasused. The change in dose rates based on amount of fuel defects is a linear function; therefore,the threshold at 2% fuel defects is: 2120R/hr *(2% / 50%) = 85 R/hr (ref. 2).The Containment High Range Radiation Monitors (HRRMs) provide indication of radiationlevels in Containment during and after postulated accidents. The monitors are two ion chamberdetectors located on the 905' level of Containment approximately 900 apart. The range of eachmonitor is 1 to 108 R/hr. The output of each detector is fed to an RM-80 located outsideContainment. The RM-80 provides monitoring, alarming, and recording functions for themonitor channel. The RM-80 works in conjunction with the PC-1 1, RM-21, and RM-23assemblies. (ref. 1)I [Document No.] Rev. 6 Page 230 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesGenericThe radiation monitor reading corresponds to an instantaneous release of all reactor coolantmass into the containment, assuming that reactor coolant activity equals 300 pCi/gm doseequivalent 1-131. Reactor coolant activity above this level is greater than that expected foriodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Sincethis condition indicates that a significant amount of fuel clad damage has occurred, itrepresents a loss of the Fuel Clad Barrier.The radiation monitor reading in this threshold is higher than that specified for RCS BarrierLoss threshold ,-.AC.1 since it indicates a loss of both the Fuel Clad Barrier and the RCSBarrier. Note that a combination of the two monitor readings appropriately escalates themrn..gc..Y classification I"ve'ECL to a Site Area Emergency.CPNPP Basis Reference(s):1. DBD-EE-023 Radiation Monitoring System2. Evaluation performed by Design Engineering & Analysis (Andrea Lemons) (AI-CR-2014-012646-15)3. EPP-312 Core Damage Assessment4. NEI 99-01 CNTMT Radiation / RCS Activity Fuel Clad Loss 3.Ai[Document No.] I Rev. 6 1 Page 231 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesFuel CladBarrier:Category:C. CNTMT Radiation / RCS ActivityDegradation Threat: LossThreshold:2. Dose equivalent 1-131 coolant activity greater than 300 pCi/ccDefinition(s):NoneBasis:Plant-SpecificNoneGenericThis threshold indicates that RCS radioactivity concentration is greater than 300 pCi/gm doseequivalent 1-131. Reactor coolant activity above this level is greater than that expected foriodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Sincethis condition indicates that a significant amount of fuel clad damage has occurred, itrepresents a loss of the Fuel Clad Barrier.There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.CPNPP Basis Reference(s):1. NEI 99-01 CNTMT Radiation / RCS Activity Fuel Clad Loss 3.B[Document No.] Rev. 6 Page 232 of276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Fuel CladCategory: C. CNTMT Radiation / RCS ActivityDegradation Threat: LossThreshold:3. Gross Failed Fuel Monitor, FFLu60 Lu-RE-0406), radiation greater than 1.0E04 pCi/mlIDefinition(s):NoneBasis:Plant-SpecificThe normal Chemical and Volume Control System (CVCS) charging and letdown flow pathallows purification of the reactor coolant and control of the RCS volume while maintaining acontinuous feed and bleed flow between the RCS and the CVCS. Reactor coolant is first"letdown" from the RCS through a regenerative heat exchanger, which minimizes heat lossesfrom the RCS. Additional cooling takes place in a letdown heat exchanger that acts as the heatsink for the system. Downstream of the letdown heat exchanger pressure control valve andupstream of the mixed bed demineralizers, the letdown stream passes by a Geiger-Muellerradiation detector, FFLu60 (u-RE-0406), mounted on the reactor coolant letdown line tomonitor coolant activity and warn of fission products in the letdown coolant if a fuel elementfailure occurs. Detection of increased coolant activity may be indicative of failed fuel. Themonitor initiates Alert and High alarms in the Control Room (PC-1 1 and Plant Computer). (ref.1).Core Damage Assessment Guidelines (VL-03-000032) which was incorporated into EPP-312"Core Damage Assessment" provides the basis for loss of the Fuel Cladding as monitored bythe Gross Failed Fuel Monitor. The setpoint recommended by Westinghouse is 1 E+04 pCi/ml(ref. 2, 3).FFLu60 (u-RE-0406) has a range of 1 E-2 -1 E+7 pCi/ml.GenericNoneCPNPP Basis Reference(s):1. DBD-EE-023 Radiation Monitoring System2. Evaluation performed by Design Engineering & Analysis (Andrea Lemons) (AI-CR-2014-012646-15)3. EPP-312 Core Damage Assessment4. NEI 99-01 Other Indications Fuel Clad Loss 5.A[Document No.] Rev. 6 Page 233 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Fuel CladCategory: C. CNTMT Radiation / RCS ActivityDegradation Threat: Potential LossThreshold:None[Document No.] Rev. 6 Page 234 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:Category:Degradation Threat:Threshold:Fuel CladD. CNTMT Integrity or BypassLossNoneI [Document No.] Rev. 6 1 Page 235 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:Category:Degradation Threat:Threshold:Fuel CladD. CNTMT Integrity or BypassPotential LossNoneI [Document No.] Rev. 6 Page 236 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesFuel CladBarrier:Category:E. Emergency Coordinator JudgmentDegradation Threat: LossThreshold:1. Any condition in the opinion of the Emergency Coordinator that indicates loss of theFuel Clad barrierDefinitNoneBasis:ion(s):Plant-SpecificThe Emergency Coordinator 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 relativelyshort period of time based on a projection of current safety system performance. Theterm "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 reliable indicators.This assessment should include instrumentation operability concerns, readings fromportable instrumentation and consideration of offsite monitoring results.* Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergency classificationdeclarations.GenericThis threshold addresses any other factors that are to be used by the Emergency DireotO-Coordinator in determining whether the Fuel Clad barrier is lostCPNPP Basis Reference(s):1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.AI [Document No.] Rev. 6 Page 237 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Fuel CladCategory: E. Emergency Coordinator JudgmentDegradation Threat: Potential LossThreshold:1. Any condition in the opinion of the Emergency Coordinator that indicates potential lossof the Fuel Clad barrierBasis:Plant-SpecificThe Emergency Coordinator 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 accident sequences.-Imminent barrier degradation exists if the degradation will likely occur within relativelyshort period of time based on a projection of current safety system performance. Theterm "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 reliable indicators.This assessment should include instrumentation operability concerns, readings fromportable instrumentation and consideration of offsite monitoring results.-Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergency classificationdeclarations.GenericThis threshold addresses any other factors that are to be used by the EmergencyCoordinatorDireGtei in determining whether the Fuel Clad barrier is potentially lost. TheEmergency D!reetE)FCoordinator should also consider whether or not to declare the barrierpotentially lost in the event that barrier status cannot be monitored.CPNPP Basis Reference(s):1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A[Document No.] Rev. 6 Page 238 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Reactor Coolant SystemCategory: A. RCS or SG Tube LeakageDegradation Threat: LossThreshold:1. An automatic or manual ECCS (SI) actuation required by EITHER:* UNISOLABLE RCS leakage* SG tube RUPTUREDefinition(s):UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally.RUPTURE -The condition of a steam generator in which primary-to-secondary leakage is ofsufficient magnitude to require a safety injection.Basis:Plant-SpecificECCS (SI) actuation is caused by (ref. 1):* Pressurizer low pressure less than 1820 psig* Steamline low pressure less than 610 psig* Containment high pressure greater than 3.0 psigGenericThis threshold is based on an UNISOLABLE RCS leak of sufficient size to require anautomatic or manual actuation of the Emergency Core Cooling System (ECCS). This conditionclearly represents a loss of the RCSBarrier.This threshold is applicable to unidentified and pressure boundary leakage, as well asidentified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacingsystem. The mass loss may be into any location -inside containment, to the secondary-side(i.e., steam generator tube leakage) or outside of containment.A steam generator with primary-to-secondary leakage of sufficient magnitude to require asafety injection is considered to be RUPTURED. If a RUPTURED steam generator is alsoFAULTED outside of containment, the declaration escalates to a Site Area Emergency sincethe Containment Barrier Loss threshold I.A will also be met.CPNPP Basis Reference(s):1. EOP-0.0A/B Reactor Trip or Safety Injection2. EOP-3.OA/B Steam Generator Tube Rupture3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Loss I.A[Document No.] I Rev. 6 Page 239 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Reactor Coolant SystemCategory: A. RCS or SG Tube LeakageDegradation Threat: Potential LossThreshold:1. Operation of a standby charging pump is required by EITHER:* UNISOLABLE RCS leakage* SG tube leakageDefinition(s):UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally.Basis:Plant-SpecificThe Chemical and Volume Control System (CVCS) includes three charging pumps (onepositive displacement pump and two centrifugal charging pumps) that take suction from thevolume control tank and return the cooled, purified reactor coolant to the RCS. The centrifugalcharging pumps in the CVCS also serve as the high-head safety injection pumps in theEmergency Core Cooling System. Positive displacement pump capacity is 98 gpm. Thecapacity of each centrifugal pump is 150 gpm. A second charging pump being required(positive displacement or centrifugal) is indicative of a substantial RCS leak. (ref. 1, 2, 3)GenericThis threshold is based on an UNISOLABLE RCS leak that results in the inability to maintainpressurizer level within specified limits by operation of a normally used charging (makeup)pump, but an ECCS (SI) actuation has not occurred. The threshold is met when an operatingprocedure, or operating crew supervision, directs that a standby charging (makeup) pump beplaced in service to restore and maintain pressurizer level.This threshold is applicable to unidentified and pressure boundary leakage, as well asidentified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacingsystem. The mass loss may be into any location -inside containment, to the secondary-side(i.e., steam generator tube leakage) or outside of containment.If a leaking steam generator is also FAULTED outside of containment, the declarationescalates to a Site Area Emergency since the Containment Barrier Loss threshold I.A will alsobe met.CPNPP Basis Reference(s):1. FSAR 9.3.42. FSAR Table 9.3-73. SOP-103A/B Chemical and Volume Control System4. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss I.A[Document No.] Rev. 6 Page 240 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesReactor Coolant SystemBarrier:Category:A. RCS or SG Tube LeakageDegradation Threat: Potential LossThreshold:2. CSFST Integrity-RED Path conditions metDefinition(s):NoneBasis:Plant-SpecificCritical Safety Function Status Tree (CSFST) RCS Integrity-RED path indicates the RCSbarrier is under significant challenge (ref. 1).GenericThis condition indicates an extreme challenge to the integrity of the RCS pressure boundaryI due to pressurized thermal shock -a transient that causes rapid RCS cooldown while the RCSis in Mode 3 or higher (i.e., hot and pressurized).CPNPP Basis Reference(s):1. FRP-0.1A/B Response to Imminent Pressurized Thermal Shock Condition2. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss I.BI [Document No.] Rev. 6 1 Page 241 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:Category:Degradation Threat:Threshold:Reactor Coolant SystemB. Inadequate Heat RemovalLossNone[Document No.] Rev. 6 Page 242 of2 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Reactor Coolant SystemCategory: B. Inadequate Heat RemovalDegradation Threat: Potential LossThreshold:1. CSFST Heat Sink-RED path conditions metANDHeat sink is requiredDefinition(s):NoneBasis:Plant-SpecificIn combination with FC Potential Loss B.2, meeting this threshold results in a Site AreaEmergency.Critical Safety Function Status Tree (CSFST) Heat Sink-RED Path indicates the ultimate heatsink function is under extreme challenge and that some fuel clad damage may potentiallyoccur (ref. 1).The CSFSTs are normally monitored using the SPDS display on the Plant Computer (ref. 1).The phrase "and heat sink required" precludes the need for classification for conditions inwhich RCS pressure is less than SG pressure or Heat Sink-RED path entry was createdthrough operator action directed by an ERG. For example, FRH-0.1 is entered from CSFSTHeat Sink-Red. Step 1 tells the operator to determine if heat sink is required by checking thatRCS pressure is greater than any non-faulted SG pressure and RCS temperature is greaterthan 3500F. If these conditions exist, Heat Sink is required. Otherwise, the operator is to eitherreturn to the procedure and step in effect and place RHR in service for heat removal. For largeLOCA events inside the Containment, the SGs are moot because heat removal through thecontainment heat removal systems takes place. Therefore, Heat Sink Red should not berequired and, should not be assessed for EAL classification because a LOCA event aloneshould not require higher than an Alert classification. (ref. 1).GenericThis condition indicates an extreme challenge to the ability to remove RCS heat using thesteam generators (i.e., loss of an effective secondary-side heat sink). This conditionrepresents a potential loss of the RCS Barrier. In accordance with EOPs, there may beunusual accident conditions during which operators intentionally reduce the heat removalcapability of the steam generators; during these conditions, classification using threshold is notwarranted.Meeting this threshold results in a Site Area Emergency because this threshold is identical toFuel Clad Barrier Potential Loss threshold 2B.2; both will be met. This condition warrants a[Document No.] I Rev. 6 Page 243 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesSite Area Emergency declaration because inadequate RCS heat removal may result in fuelheat-up sufficient to damage the cladding and increase RCS pressure to the point where masswill be lost from the system.CPNPP Basis Reference(s):1. FRH-O.1A/B Response to Loss of Secondary Heat Sink2. NEI 99-01 Inadequate Heat Removal RCS Loss 2.B[Document No.] Rev. 6 Page 244 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Reactor Coolant SystemCategory: C. CNTMT Radiation/ RCS ActivityDegradation Threat: LossThreshold:1. Containment radiation greater than 5 R/hrCTEu16 Containment HRRM (u-RE-6290A), orCTWu17 Containment HRRM (u-RE-6290B)Definition(s):N/ABasis:Plant-SpecificAs part of the elimination of the Post-Accident Sampling system, Westinghouse performedanalysis for Comanche Peak on Core Damage Assessment Guidelines (VL-03-000032) whichwas incorporated into EPP-312 "Core Damage Assessment". For setpoint CRM1,Westinghouse assumptions match the requirements of NEI 99-01 for RCS barrier loss with theexception that the level of radioactivity in the RCS is assumed to be at 10% of TechnicalSpecifications levels rather than 100% as recommended by NEI 99-01. However, this addsconservatism to this threshold. The limiting maximum value found in Figure 1A is 4.75 R/hr.This value is time dependent and corresponds to an hour after shutdown. This value has beenrounded to 5 R/hr for instrument readability (ref. 1, 3, 4, 5).The Containment High Range Radiation Monitors (HRRMs) provide indication of radiationlevels in Containment during and after postulated accidents. The monitors are two ion chamberdetectors located on the 905' level of Containment approximately 90' apart. The range of eachmonitor is 1 to 108 R/hr. The output of each detector is fed to an RM-80 located outsideContainment.The RM-80 provides monitoring, alarming, and recording functions for the monitor channel.The RM-80 works in conjunction with the PC-1 1, RM-21, and RM-23 assemblies. (ref. 2)GenericThe radiation monitor reading corresponds to an instantaneous release of all reactor coolantmass into the containment, assuming that reactor coolant activity equals TechnicalSpecification allowable limits. This value is lower than that specified for Fuel Clad Barrier Lossthreshold ,AC.1 since it indicates a loss of the RCS Barrier only.There is no Potential Loss threshold associated with RCS Activity I Containment Radiation.I [Document No.] Rev. 6 Page 245 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesCPNPP Basis Reference(s):1. Technical Specifications Table 3.3.3-12. DBD-EE-023 Radiation Monitoring System3. Evaluation performed by Design Engineering & Analysis (Andrea Lemons) (AI-CR-2014-012646-15)4. Technical Specifications B3.3.35. EPP-312 Core Damage Assessment6. NEI 99-01 CNTMT Radiation / RCS Activity RCS Loss 3.AI [Document No.] Rev. 6 Page 246 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Reactor Coolant SystemCategory: B. CNTMT Radiation/ RCS ActivityDegradation Threat: Potential LossThreshold:None[Document No.] Rev. 6 1 Page 247 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Reactor Coolant SystemCategory: D. CNTMT Integrity or BypassDegradation Threat: LossThreshold:NoneI [Document No.] I Rev. 6 Page 248 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Reactor Coolant SystemCategory: D. CNTMT Integrity or BypassDegradation Threat: Potential LossThreshold:NoneL[Document No.] I Rev. 6 Page 249 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesReactor Coolant SystemBarrier:Category:Degradation Threat:E. Emergency Coordinator JudgmentLossThreshold:Definition(s):NoneBasis:Plant-SpecificThe Emergency Coordinator judgment threshold addresses any other factors relevant todetermining if the RCS barrier is lost. Such a determination should include imminent barrierdegradation, barrier monitoring capability and dominant accident sequences." Imminent barrier degradation exists if the degradation will likely occur within relativelyshort period of time based on a projection of current safety system performance. Theterm "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 reliable indicators.This assessment should include instrumentation operability concerns, readings fromportable instrumentation and consideration of offsite monitoring results.* Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergency classificationdeclarations.GenericThis threshold addresses any other factors that may be used by the Emergency QieGteCoordinator in determining whether the RCS Barrier is lost.CPNPP Basis Reference(s):1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A[Document No.] Rev. 6 Page 250 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: Reactor Coolant SystemCategory: E. Emergency Coordinator JudgmentDegradation Threat: Potential LossThreshold:1. Any condition in the opinion of the Emergency Coordinator that indicates potential lossof the RCS barrierDefinition(s):NoneBasis:Plant-SpecificThe Emergency Coordinator judgment threshold addresses any other factors relevant todetermining if the RCS barrier is potentially 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 relativelyshort period of time based on a projection of current safety system performance. Theterm "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 reliable indicators.This assessment should include instrumentation operability concerns, readings fromportable instrumentation and consideration of offsite monitoring results.* Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergency classificationdeclarations.GenericThis threshold addresses any other factors that may be used by the Emergency Direet4Coordinator in determining whether the RCS Barrier is potentially lost. The EmergencyDirector should also consider whether or not to declare the barrier potentially lost in the eventthat barrier status cannot be monitored.CPNPP Basis Reference(s):1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6.AI [Document No.] Rev. 6 1 Page 251 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: ContainmentCategory: A. RCS or SG Tube LeakageDegradation Threat: LossThreshold:1. A leaking or RUPTURED SG is FAULTED outside of containmentDefinition(s):FAULTED -The term applied to a steam generator that has a steam leak on the secondaryside of sufficient size to cause an uncontrolled drop in steam generator pressure or the steamgenerator to become completely depressurized.RUPTURED -The condition of a steam generator in which primary-to-secondary leakage is ofsufficient magnitude to require a safety injection.Basis:Plant-SpecificNone.GenericThis threshold addresses a leaking or RUPTURED Steam Generator (SG) that is alsoFAULTED outside of containment. The condition of the SG, whether leaking or RUPTURED,is determined in accordance with the thresholds for RCS Barrier Potential Loss 4-.A.1 and Loss4-A.1, respectively. This condition represents a bypass of the containment barrier.FAULTED is a defined term within the NEI 99-01 methodology; this determination is notnecessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if thepressure in a steam generator is decreasing uncontrollably .(part of the FAULTED definition).and the FAULTED steam generator isolation procedure is not entered because EOP user rulesare dictating implementation of another procedure to address a higher priority condition, thesteam generator is still considered FAULTED for emergency classification purposes.The FAULTED criterion establishes an appropriate lower bound on the size of a steam releasethat may require an emergency classification. Steam releases of this size are readilyobservable with normal Control Room indications. The lower bound for this aspect of the[Document No.] Rev. 6 Page 252 of276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and Basescontainment barrier is analogous to the lower bound criteria specified in IC SU4 for the fuelclad barrier (i.e., RCS activity values) and IC SU5 for the RCS barrier (i.e., RCS leak ratevalues).This threshold also applies to prolonged steam releases necessitated by operationalconsiderations such as the forced steaming of a leaking or RUPTURED steam generatordirectly to atmosphere to cooldown the plant, or to drive an auxiliary (emergency) feed waterpump. These types of conditions will result in a significant and sustained release of radioactivesteam to the environment (and are thus similar to a FAULTED condition). The inability toisolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss ofcontainment.Steam releases associated with the expected operation of a SG power operated relief valve orsafety relief valve do not meet the intent of this threshold. Such releases may occurintermittently for a short period of time following a reactor trip as operators process throughemergency operating procedures to bring the plant to a stable condition and prepare to initiatea plant cooldown. Steam releases associated with the unexpected operation of a valve (e.g., astuck-open safety valve) do meet this threshold.Following an SG tube leak or rupture, there may be minor radiological releases through asecondary-side system component (e.g., air ejectors, glad seal exhausters, valve packing,etc.). These types of releases do not constitute a loss or potential loss of containment butshould be evaluated using the Recognition Category A-R ICs.The emergenYcy clasification levelECLs resulting from primary-to-secondary leakage, with orwithout a steam release from the FAULTED SG, are summarized below.I [Document No.] I Rev. 6 Page 253 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesAffected SG is FAULTEDOutside of Containment?P-to-S Leak RateYesNoLess than or equal to 25 gpmGreater than 25 gpmRequires operation of a standbycharging (makeup) pump (RCSBarrier Potential Loss)Requires an automatic or manualECCS (SI) actuation (RCSBarrierLoss)No classificationUnusual Event perSU4SU5.1Site Area Emergency perFS1.1Site Area Emergency perFS1.1No classificationUnusual Event perSJ4SU5.1Alert per FAI.1Alert per FAI.1There is no Potential Loss threshold associated with RCS or SG Tube Leakage.CPNPP Basis Reference(s):1. EOP-3.0 Steam Generator Tube Rupture2. EOP-2.OA/B Faulted Steam Generator Isolation3. NEI 99-01 RCS or SG Tube Leakage Containment Loss I.A[Document No.] Rev. 6 Page 254 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: ContainmentCategory: A. RCS or SG Tube LeakageDegradation Threat: Potential LossThreshold:NoneI [Document No.] Rev. 6 Page 255 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:Category:Degradation Threat:Threshold:NoneContainmentB. Inadequate heat RemovalLossI [Document No.] Rev. 6 Page 256 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesContainmentBarrier:Category:B. Inadequate heat RemovalDegradation Threat: Potential LossThreshold:1. CSFST Core Cooling-RED Path conditions metANDRestoration procedures not effective within 15 min. (Note 1)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has beenexceeded, or will likely be exceeded.Definition(s):NoneBasis:Plant-SpecificCritical Safety Function Status Tree (CSFST) Core Cooling-RED path indicates significant coreexit superheating and core uncovery. The CSFSTs are normally monitored using the SPDSdisplay on the Plant Computer (ref. 1).The function restoration procedures are those emergency operating procedures that addressthe recovery of the core cooling critical safety functions. The procedure is considered effectiveif the temperature is decreasing or if the vessel water level is increasing (ref. 1).A direct correlation to status trees can be made if the effectiveness of the restorationprocedures is also evaluated. If core exit thermocouple (TC) readings are greater than 1,2000F(ref. 1), Fuel Clad barrier is also lost.GenericThis threshold addresses any other factors that may be used by the Emergency ir-eeGtCoordinator in determining whether the RCS Barrier is potentially lost. The EmergencyDirector should also consider whether or not to declare the barrier potentially lost in the eventthat barrier status cannot be monitored.CPNPP Basis Reference(s):1. FRC-O.1A/B Response to Inadequate Core Cooling2. NEI 99-01 Inadequate Heat Removal Containment Potential Loss 2.AI[Document No.] I Rev. 6 -Page 257 of 276 ý ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: ContainmentCategory: C. CNTMT Radiation/RCS ActivityDegradation Threat: LossThreshold:None[Document No.] Rev. 6 Page 258 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: ContainmentCategory: C. CNTMT Radiation/RCS ActivityDegradation Threat: Potential LossThreshold:1. Containment radiation greater than 1,110 R/hrCTEul 6 Containment HRRM (u-RE-6290A), orCTWu17 Containment HRRM (u-RE-6290B)Definition(s):NoneBasis:Plant-SpecificContainment radiation monitor readings greater than 1,110 R/hr indicate significant fueldamage well in excess of that required for loss of the RCS barrier and the Fuel Clad barrier.Regardless of whether the Containment barrier itself is challenged, this amount of activity incontainment could have severe consequences if released. It is, therefore, prudent to treat thisas a Potential Loss of the Containment barrier. (ref. 2, 3)The readings are higher than that specified for Fuel Clad Loss C.3 and RCS Loss C.1.Containment radiation readings at or above the Containment barrier Potential Loss threshold,therefore, signify a loss of two fission product barriers and Potential Loss of a third, indicatingthe need to upgrade the emergency classification to a General Emergency.The analysis performed in VL-03-00032 was used to determine the potential containment lossthreshold. Since the containment potential loss threshold also assumes a loss of the RCSbarrier, the Figure 3A (CRM3) curve which represents the dose response at low RCS pressurewith sprays present was used. The setpoint was developed on the assumption of 100% fuelrod rupture with 100% of the noble gas and 50% of the iodine and cesium in the RCS releasedto containment. With containment spray operating, the containment inventory of all fissionproducts except the noble gases are reduced by a factor of 100. Per Figure 3A, the value onehour after shutdown for 100% rod rupture is 5560 R/hr; therefore, the EAL threshold at 20%fuel defects is: 5,560 R/hr *20% = 1,112 R/hr (rounded to 1,110 R/hr for instrument readability)(ref. 2, 3).The Containment High Range Radiation Monitors (HRRMs) provide indication of radiationlevels in Containment during and after postulated accidents. The monitors are two ion chamberdetectors located on the 905' level of Containment approximately 90' apart. The range of eachmonitor is 1 to 108 R/hr. The output of each detector is fed to an RM-80 located outsideContainment. The RM-80 provides monitoring, alarming, and recording functions for themonitor channel. The RM-80 works in conjunction with the PC-1 1, RM-21, and RM-23assemblies. (ref. 1).I [Document No.] I Rev. 6 Page 259 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesGenericThe radiation monitor reading corresponds to an instantaneous release of all reactor coolantmass into the containment, assuming that 20% of the fuel cladding has failed. This level offuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss andRCS Barrier Loss thresholds.NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power PlantAccidents, indicates the fuel clad failure must be greater than approximately 20% in order forthere to be a major release of radioactivity requiring offsite protective actions. For thiscondition to exist, there must already have been a loss of the RCS Barrier and the Fuel CladBarrier. It is therefore prudent to treat this condition as a potential loss of containment whichwould then escalate the emergencY claSificatilo; levelECL to a General Emergency.CPNPP Basis Reference(s):1. DBD-EE-023 Radiation Monitoring System2. Evaluation performed by Design Engineering & Analysis (Andrea Lemons) (AI-CR-2014-012646-15)3. EPP-312 Core Damage Assessment4. NEI 99-01 CNTMT Radiation / RCS Activity Containment Potential Loss 3.AI [Document No.] Rev. 6 1 Page 260 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: ContainmentCategory: D. CNTMT Integrity or BypassDegradation Threat: LossThreshold:1. Containment isolation is requiredAND EITHER:* Containment integrity has been lost based on Emergency Coordinator judgment" UNISOLABLE pathway from containment to the environment existsDefinition(s):UNISOLABLE- An open or breached system line that cannot be isolated, remotely or locally.Basis:Plant-SpecificNoneGenericThese thresholds address a situation where containment isolation is required and one of twoconditions exists as discussed below. Users are reminded that there may be accident andrelease conditions that simultaneously meet both bulleted thresholds 4.A.1 and 4.A.2.4-A41First Threshold -Containment integrity has been lost, i.e., the actual containmentatmospheric leak rate likely exceeds that associated with allowable leakage (or sometimesreferred to as design leakage). Following the release of RCS mass into containment,containment pressure will fluctuate based on a variety of factors; a loss of containment integritycondition may (or may not) be accompanied by a noticeable drop in containment pressure.Recognizing the inherent difficulties in determining a containment leak rate during accidentconditions, it is expected that the Emergency DirectEr Coordinator will assess this thresholdusing judgment, and with due consideration given to current plant conditions, and availableoperational and radiological data (e.g., containment pressure, readings on radiation monitorsoutside containment, operating status of containment pressure control equipment, etc.).Refer to the middle piping run of Figure 9-F--41. Two simplified examples are provided. One isleakage from a penetration and the other is leakage from an in-service system valve.I [Document No.] Rev. 6 1 Page 261 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesDepending upon radiation monitor locations and sensitivities, the leakage could be detected byany of the four monitors depicted in the figure.Another example would be a loss or potential loss of the RCS barrier, and the simultaneousoccurrence of two FAULTED locations on a steam generator where one fault is located insidecontainment (e.g., on a steam or feedwater line) and the other outside of containment. In thiscase, the associated steam line provides a pathway for the containment atmosphere to escapeto an area outside the containment.Following the leakage of RCS mass into containment and a rise in containment pressure, theremay be minor radiological releases associated with allowable (design) containment leakagethrough various penetrations or system components. These releases do not constitute a lossor potential loss of containment but should be evaluated using the Recognition Category A--RICs.4-A.2Second Threshold -Conditions are such that there is an UNISOLABLE pathway for themigration of radioactive material from the containment atmosphere to the environment. Asused here, the term "environment" includes the atmosphere of a room or area, outside thecontainment, that may, in turn, communicate with the outside-the-plant atmosphere (e.g.,through discharge of a ventilation system or atmospheric leakage). Depending upon a varietyof factors, this condition may or may not be accompanied by a noticeable drop in containmentpressure.Refer to the top piping run of Figure 8-F--41. In this simplified example, the inboard andoutboard isolation valves remained open after a containment isolation was required (i.e.,containment isolation was not successful). There is now an UNISOLABLE pathway from thecontainment to the environment.The existence of a filter is not considered in the threshold assessment. Filters do not removefission product noble gases. In addition, a filter could become ineffective due to iodine and/orparticulate loading beyond design limits (i.e., retention ability has been exceeded) or watersaturation from steam/high humidity in the release stream.Leakage between two interfacing liquid systems, by itself, does not meet this threshold.I [Document No.] Rev. 6 1 Page 262 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesRefer to the bottom piping run of Figure 8-F--4-1. In this simplified example, leakage in an RCPseal cooler is allowing radioactive material to enter the Auxiliary Building. The radioactivitywould be detected by the Process Monitor. If there is no leakage from the closed watercooling system to the Auxiliary Building, then no threshold has been met. If the pumpdeveloped a leak that allowed steam/water to enter the Auxiliary Building, then secondthreshold-44B would be met. Depending upon radiation monitor locations and sensitivities, thisleakage could be detected by any of the four monitors depicted in the figure and cause the firstthreshold 4-A-_--to be met as well.Following the leakage of RCS mass into containment and a rise in containment pressure, theremay be minor radiological releases associated with allowable containment leakage throughvarious penetrations or system components. Minor releases may also occur if a containmentisolation valve(s) fails to close but the containment atmosphere escapes to an enclosedsystem. These releases do not constitute a loss or potential loss of containment but should beevaluated using the Recognition Category A-RICs.The status of the containment barrier during an event involving steam generator tube leakageis assessed using Loss Threshold 4-.A.1.CPNPP Basis Reference(s):1. NEI 99-01 CNTMT Integrity or Bypass Containment Loss 4.A[Document No.] Rev. 6 Page 263 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: ContainmentCategory: D. CNTMT Integrity or BypassDegradation Threat: LossThreshold:2. Indications of RCS leakage outside of containmentDefinition(s):NoneBasis:Plant-SpecificECA-1.2A/B LOCA Outside Containment (ref. 1) provides instructions to identify and isolate aLOCA outside of the containment. Potential RCS leak pathways outside containment include(ref. 1):" Residual Heat Removal* Safety Injection" Chemical & Volume Control" RCP seals" PZR/RCS Loop sample linesGenericContainment sump, temperature, pressure and/or radiation levels will increase if reactorcoolant mass is- leaking into the containment. If these parameters have not increased, then thereactor coolant mass may be leaking outside of containment (i.e., a containment bypasssequence). Increases in sump, temperature, pressure, flow and/or radiation level readingsoutside of the containment may indicate that the RCS mass is being lost outside ofcontainment.Unexpected elevated readings and alarms on radiation monitors with detectors outsidecontainment should be corroborated with other available indications to confirm that the sourceis a loss of RCS mass outside of containment. If the fuel clad barrier has not been lost,radiation monitor readings outside of containment may not increase significantly; however,other unexpected changes in sump levels, area temperatures or pressures, flow rates, etc.should be sufficient to determine if RCS mass is being lost outside of the containment.Refer to the middle piping run of Figure 9-F--41. In this simplified example, a leak has occurredat a reducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending uponradiation monitor locations and sensitivities, the leakage could be detected by any of the fourmonitors depicted in the figure and cause threshold 4AD.1 to be met as well.To ensure proper escalation of the emergency classification, the RCS leakage outside ofcontainment must be related to the mass loss that is causing the RCS Loss and/or PotentialLoss threshold 4-.A.1 to be met.[Document No.] Rev. 6 Page 264 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesCPNPP Basis Reference(s):1. ECA-1.2A/B LOCA Outside Containment2. NEI 99-01 CNTMT Integrity or Bypass Containment Loss[Document No.] Rev. 6 Page 265 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesFigure 1: Containment Integrity or Bypass Examples-easeEffluent oing Monitor 9O wa-----------I [Document No.] Rev. 6 Page 266 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: ContainmentCategory: D. CNTMT Integrity or BypassDegradation Threat: Potential LossThreshold:1. CSFST Containment-RED Path conditions metDefinition(s):NoneBasis:Plant-SpecificCritical Safety Function Status Tree (CSFST) Containment-RED path is entered if containmentpressure is greater than or equal to 50 psig and represents an extreme challenge to safetyfunction. The CSFSTs are normally monitored using the SPDS display on the Plant Computer(ref. 1).50 psig is the containment design pressure and is the pressure used to define CSFSTContainment Red Path conditions (ref. 2).GenericIf containment pressure exceeds the design pressure, there exists a potential to lose theContainment Barrier. To reach this level, there must be an inadequate core cooling conditionfor an extended period of time; therefore, the RCS and Fuel Clad barriers would already belost. Thus, this threshold is a discriminator between a Site Area Emergency and GeneralEmergency since there is now a potential to lose the third barrier.CPNPP Basis Reference(s):1. FRC-Z.1A/B Response to High Containment Pressure2. FSAR Table 6.2.1-13. NEI 99-01 CNTMT Integrity or Bypass Containment Potential Loss 4.AI [Document No.] Rev. 6 1 Page 267 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: ContainmentCategory: D. CNTMT Integrity or BypassDegradation Threat: Potential LossThreshold:2. Containment hydrogen concentration greater than 4%Definition(s):NoneBasis:Plant-SpecificIn the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a coreuncovery could result in an explosive mixture of dissolved gasses in Containment. However,Containment monitoring and/or sampling should be performed to verify this assumption and aGeneral Emergency declared if it is determined that an explosive mixture exists. A combustiblemixture can be formed when hydrogen gas concentration in the Containment atmosphere isgreater than 4% by volume. All hydrogen measurements are referenced to concentrations indry air even though the actual Containment environment may contain significant steamconcentrations. The plant has two hydrogen monitoring systems. Each monitoring systemconsists of four sensor modules and one microprocessor analyzer. Two sensors from eachContainment are coupled to one of the two hydrogen microprocessors located in the ControlRoom. Thus each microprocessor analyzer is shared by Units 1 and 2. The analyzer systemhas a range of 0-10% hydrogen by volume. The detector modules are located on the 905',873', and 860' elevations in Containment. A fourth detector is located on 832' level across fromthe loop room entrance for loops 1 and 4. Hydrogen concentration is displayed in the ControlRoom on u-AI-5506A/B and u-AI-5506C/D.Hydrogen concentration can also be displayed on the Plant Computer. Alarms at -3% areprovided for high hydrogen concentration, u-ALB-3A, window 3.7. If a hydrogen concentrationvalue can not be obtained from the hydrogen monitoring system, a grab sample from thecontainment PIG radiation monitor may be used to determine the hydrogen concentration (ref.1,2,3,4).To generate such levels of combustible gas, loss of the Fuel Clad and RCS barriers must haveoccurred. With the Potential Loss of the Containment barrier, the threshold hydrogenconcentration, therefore, will likely warrant declaration of a General Emergency.GenericThe existence of an explosive mixture means, at a minimum, that the containment atmospherichydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagrationlimit). A hydrogen burn will raise containment pressure and could result in collateral equipmentdamage leading to a loss of containment integrity. It therefore represents a potential loss ofthe Containment Barrier.[Document No.] Rev. 6 Page 268 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesCPNPP Basis Reference(s):1. FRC-O.1A/B Response to Inadequate Core Cooling, Attachment 52. FSAR Section 6.2.53. FSAR Table 7.5-7A4. CHM-1 11, Primary Chemistry Accident Assessment Sampling Program7. NEI 99-01 CNTMT Integrity or Bypass Containment Potential Loss 4.BI[Document No.] Rev. 6 1 Page 269 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: ContainmentCategory: D. CNTMT Integrity or BypassDegradation Threat: Potential LossThreshold:3. Containment pressure greater than 18 psig with neither Containment Spray systemtrain operating per design for greater than or equal to 15 min. (Note 1)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has beenexceeded, or will likely be exceeded.Definition(s):NoneBasis:Plant-SpecificThis threshold represents a Potential Loss of the Containment barrier because theContainment heat removal and depressurization equipment (but not including Containmentventing strategies) is either lost or degraded. The Containment Spray System (CSS) isdesigned to remove heat from the Containment environment following a LOCA, a main steamline break accident, or a feedwater line break accident. Each unit of the CPNPP is equippedwith two redundant Containment spray trains, each designed to provide emergencyContainment heat removal in the event of a LOCA. This system, in conjunction with the ECCS,removes postaccident thermal energy from the Containment environment, thereby reducingthe Containment pressure and temperature. Each train includes two containment spraypumps, spray headers, nozzles, valves, and piping. Each train is powered from a separatesafeguard bus. (ref. 1)The Containment pressure setpoint (18 psig, ref. 2) is the pressure at which the ContainmentSpray System should actuate and begin performing its function. The design basis accidentanalyses and evaluations assume the loss of one Containment Spray System train (ref. 1).GenericThis threshold describes a condition where containment pressure is greater than the setpointat which containment energy (heat) removal systems are designed to automatically actuate,and less than one full train of equipment is capable of operating per design. The 15-minutecriterion is included to allow operators time to manually start equipment that may not haveautomatically started, if possible. This threshold represents a potential loss of containment inthat containment heat removal/depressurization systems (e.g., containment sprays, icecondenser fans, etc., but not including containment venting strategies) are either lost orperforming in a degraded manner.I [Document No.] I Rev. 6 Page 270 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesCPNPP Basis Reference(s):1. FSAR Section 6.2.22. FRC-Z.1A/B Response to High Containment Pressure3. NEI 99-01 CNTMT Integrity or Bypass Containment Potential Loss 4.C[Document No.] Rev. 6 Page 271 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: ContainmentCategory: E. Emergency Coordinator JudgmentDegradation Threat: LossThreshold:1. Any condition in the opinion of the Emergency Coordinator that indicates loss of theContainment barrierDefinition(s):NoneBasis:Plant-SpecificThe Emergency Coordinator judgment threshold addresses any other factors relevant todetermining if the Primary Containment barrier is lost. Such a determination should includeimminent barrier degradation, barrier monitoring capability and dominant accident sequences.* Imminent barrier degradation exists if the degradation will likely occur within relativelyshort period of time based on a projection of current safety system performance. Theterm "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 reliable indicators.This assessment should include instrumentation operability concerns, readings fromportable instrumentation and consideration of offsite monitoring results.* Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergency classificationdeclarations.GenericThis threshold addresses any other factors that may be used by the Emergency QikeeterCoordinator in determining whether the Containment Barrier is lost.CPNPP Basis Reference(s):1. NEI 99-01 Emergency Director Judgment PC Loss 6.A[Document No.] Rev. 6 Page 272 of 276 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier: ContainmentCategory: E. Emergency Coordinator JudgmentDegradation Threat: Potential LossThreshold:1. Any condition in the opinion of the Emergency Coordinator that indicates potential lossof the Containment barrierDefinition(s):NoneBasis:Plant-SpecificThe Emergency Coordinator judgment threshold addresses any other factors relevant todetermining if the Primary Containment barrier is potentially lost. Such a determination shouldinclude imminent barrier degradation, barrier monitoring capability and dominant accidentsequences." Imminent barrier degradation exists if the degradation will likely occur within relativelyshort period of time based on a projection of current safety system performance. Theterm "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 reliable indicators.This assessment should include instrumentation operability concerns, readings fromportable instrumentation and consideration of offsite monitoring results." Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergency classificationdeclarations.GenericThis threshold addresses any other factors that may be used by the Emergency DiQeetoCoordinator in determining whether the Containment Barrier is lost.CPNPP Basis Reference(s):1. NEI 99-01 Emergency Director Judgment PC Potential Loss 6.AI [Document No.] I Rev. 6 1 Page 273 of 276 ATTACHMENT 3Safe Operation & Shutdown Areas Tables R-3 & H-2 BasesBackgroundNEI 99-01 Revision 6 ICs AA3 and HA5 prescribe declaration of an Alert based on impededaccess to rooms or areas (due to either area radiation levels or hazardous gas concentrations)where equipment necessary for normal plant operations, cooldown or shutdown is located.These areas are intended to be plant operating mode dependent. Specifically the DevelopersNotes for AA3 and HA5 states:The "site-specific list of plant rooms or areas with entry-related mode applicability identified"should specify those rooms or areas that contain equipment which require a manual/localaction as specified in operating procedures used for normal plant operation, cooldown andshutdown. Do not include rooms or areas in which actions of a contingent or emergencynature would be performed (e.g., an action to address an off-normal or emergency conditionsuch as emergency repairs, corrective measures or emergency operations). In addition, thelist should specify the plant mode(s) during which entry would be required for each room orarea.The list should not include rooms or areas for which entry is required solely to performactions of an administrative or record keeping nature (e.g., normal rounds or routineinspections).Further, as specified in IC HA5:The list need not include the Control Room if adequate engineered safety/design featuresare in place to preclude a Control Room evacuation due to the release of a hazardous gas.Such features may include, but are not limited to, capability to draw air from multiple airintakes at different and separate locations, inner and outer atmospheric boundaries, or thecapability to acquire and maintain positive pressure within the Control Room envelope.I [Document No.] Rev. 6 Page 274 of 276 ATTACHMENT 3Safe Operation & Shutdown Areas Tables R-3 & H-2 BasesCPNPP Table R-3 and H-2 BasesA review of station operating procedures identified the following mode dependent in-plantactions and associated areas that are required for normal plant operation, cooldown orshutdown:Location- Modes- Modes-Safe Shutdown Area 1, 2 3, 4, 5, or 6Charging Pump Rooms SDC Equipment. Shut Down Cooling (SDC)-No entry required -No entry requiredInventory Control Equipment Inventory Control Equipment-Entry required during pump -Entry required during pumpstarts and stops starts and stopsReactivity Control.-Entry required during pumpstarts and stops-Containment Spray Pumps A Post-accident Containment Post-accident Containmentand B Pressure Control Pressure Control (modes 3 and-No entry required 4)-No entry required-SI Pumps A and B Post-accident ECCS Post-accident ECCS-No entry required -No entry required-Residual Heat Removal Pumps Post-accident ECCS Decay Heat Removal (Modes 4, 5,A and B -No entry required and 6-Entry required for pumpsstarts and stops-CVCS Valve Rooms, Auxiliary Inventory Control Equipment Inventory Control EquipmentBuilding 810' and 822' -Entry required during -Entry required during pumppump starts and stops starts and stopsReactivity Control. Reactivity Control.-Entry required during -Entry required during pumppump starts and stops starts and stops-Service Water Intake Structure Ultimate Heat Sink Equipment for Ultimate Heat Sink Equipment forHabitability Control, Habitability Control, ContainmentContainment Temperature, Temperature, and Shutdownand Shutdown Cooling Cooling-No entry required No entry required1E Switchgear Rooms 810', Electrical Power. Electrical Power.832', and 852' -Entry required for manual -Entry required for manualbreaker manipulations on breaker manipulations oncomponent operations, component operationsreactor startup andshutdownControl Building 807'Cable Electrical Power. Electrical Power.Spreading Room -No entry required -No entry requiredControl Building 792' UPS and Electrical Power. Electrical Power.Battery Rooms -No entry requiredEmergency Diesel Generators A Electrical Power. Electrical Power.& B -No entry required -No entry requiredEmergency Diesel Generators Electrical Power. Electrical Power.Day Tank Rooms -No entry required -No entry requiredControl Building 830' Control Continuously occupied, Continuously occupied, capableRoom capable of Ventilation Isolation of Ventilation Isolation mode,mode, covered under H-6 covered under H-6Control Building 840' Technical -No entry required No entry requiredSupport Center, [Document No.] Rev. 6 Page 275 of 276 ATTACHMENT 3Safe Operation & Shutdown Areas Tables R-3 & H-2 BasesLocation- Modes- Modes-Safe Shutdown Area 1, 2 3, 4, 5, or 6Control Building 778' Safety -No entry required -No entry requiredChiller RoomsAuxiliary Building 790' -No entry required -No entry requiredAuxiliary Building 810' other -No entry required -No entry requiredthan CVCS Valave Rooms andCharging Pump RoomsAuxiliary Building 830' -No entry required -No entry requiredAuxiliary Building 852' -No entry required -No entry requiredAuxiliary Building 873' -No entry required -No entry requiredAuxiliary Building 886' -No entry required -No entry requiredSafeguards 790' -No entry required -No entry requiredSafeguards 810' -No entry required -No entry requiredSafeguards 831' -No entry required -No entry requiredSafeguards 852' -No entry required -No entry requiredSafeguards 873' -No entry required -No entry requiredTurbine Building Elevations -No entry required -No entry requiredAux. Feedwater Pump Rooms Steam Generator Heat Removal Steam Generator Heat RemovalA, B, and Turbine Driven -No entry required -No entry requiredTable R-3 & H-2 ResultsTable R-3/H-2 Safe Operation & Shutdown Rooms/AreasRoom/Area Mode ApplicabilityCharging Pump Rooms 1, 2, 3, 4, 5, 6CVCS Valve Rooms 1,2, 3, 4, 5, 61 E Switchgear Rooms AllRHR Pump Rooms 4, 5, 6Plant Operating Procedures Reviewed1.2.3.4.5.IPO-003A/BIPO-005A/BIPO-001A/BIPO-002A/BSOP-1 036. SOP-1047. SOP-102I [Document No.] Rev. 6 Page 276 of 276 ATTACHMENT 5 TO TXX-15101CPNPP RADIOLOGICAL EFFLUENTEAL VALUES(7 PAGES)
EAL Section RRevision 6Table R-1 Effluent MonitorClassification Thresholds ReviewSubmitted By:Signature: /JACmrygWiecheringEieerg'ercyb~'nningDate:Review~ed By:Signature: JA .f4 .Hrf Nt 4,Jeffery Hull, Emergency IFlgrhiing ManagerDate: 6 Signature: 16y ( " aer-Ky Fishencord, E-m g-enc-y Pla'n'ningSignaure:cr1 IDe6 O' onnor, Radiation Protection ManagerSignature:Date: __2 3Date:-- ........Andre'a Lemons, Senior EngineerI of 706/18/2015 EAL Section R Revision 6Table R-I Effluent Monitor Classification ThresholdsTable R- 1 Effluent Monitor Classification ThresholdsType Release Point Monitor GE SAE ALERT UEGaseous Plant Vent X-RE-5567 A+B --- --------6.52E-4*PVG-384 +PVG-385 uCi/ccPlant Vent X-RE-5570 A+B 4.0E07
- 4.0E06
- 4.0E05
- 4.0E04*(WRGM) uCi/sec uCi/sec uCi/sec uCi/secPVF-684 + PVF-685 uMain Steam 90 uCi/cc 9 uCi/cc .9 uCi/cc 2 X HighMSL-a78 M-RE-2325 Alarm SetpointMSL-i79 u-RE-2326MSL-*180 U-RE-2327MSL-u8l P-RE-2328Liquid Liquid Waste 2 X HighLWE-076 X-RE-__53 Alarm SetpointService Water u-RE.4269 ---- ---- 2 X HighSSW..,65 u-RE-4270 Alarm SetpointSSW_-A66 I I I* Total of the two monitors equal this value or greater.Reference Material for Table R-1 Effluent Monitor Classification ThresholdsType Release Point Monitor High Source for Range**Alarm AlarmGaseous Plant Vent X-RE-5567 A or B 3.26E-4 CLI-744-3 I E-06 to I E-02PVG-384 +PVG-385 uCi/cc uCilccPlant Vent X-RE-5570 A or B 2.0E04 CLI-744-3 IE-06 to IE+05(WRGM) uCi/sec uCifccPVF-684 + PVF-685 I E-04 to 1E+12uCi/Sec***Main Steam u-RE-2325 2.OE-1 PC-11 IE-01 to IE+03MSL--78 u-RE-2326 uCi/cc uCi/ccMSL-U79MSL-8U0 u--RE-2327MSL-1u81 u-RE-2328Liquid Liquid Waste X-RE-5253 2.48E-04 CLI-744-1 1 E-05 to 5E-02LWE-076 uCi/sec uCi/ccService Water u-RE-4269 4.OE-06 PC-I1 IE-05 to 5E-02SSw%-65 u-RE-4270 uCi/sec uCi/ccSSW-3j66 I I II** Range DBD-EE-023 Specified Instrument Range unless noted.*** Range is from the PC-11.2 of 706/18/2015 EAL Section R Revision 6Table R-1 Effluent Monitor Classification ThresholdsDose Projection AssumptionsThe dose projections used the following:Only one unit was used to make dose projections.One mile was used for Exclusion Area Boundary.Reactor has been Shutdown for one hour.The event start time is same as Reactor Shutdown.Accident Type for Vent release was a LOCA.Filtration: NoStability Class: DWind Speed: 10 miles per hourPlant Vent UE AssumptionsPVG-384 +PVG-385 X-RE-5567 A+B (This monitor is used for an UE only)High Alarm Setpoint: 3.26E-4 uCi/ccPlant Vent Total: 6.52E-04 uCi/ccFlow Rate: 140,000 scfmReactor has been Shut Down for one hour.Projected Dose using CPAMPEDE 8.0PROJECTED DOSE I Mile 2 Mile 5 Mile 10 MileTEDE (Rem) 0.002 0.0008 0.0002 0.0001CDE (Rem) 0.009 0.003 0.0009 0.0003Plume Shine (R/hr) 0.0002 0.0001 0.0000 0.00003 of 706/18/2015 EAL Section R Revision 6Table R-1 Effluent Monitor Classification ThresholdsPlant Vent (WRGM) UE AssumptionsPlant Vent (WRGM) PVF-684 + PVF-685 X-RE-5570 A+BHigh Alarm Setpoint: 2.0E04 uCi/secCPAMPEDEValue U1 Value U22E4 uCi/Sec 2E4 uCi/SecPlant Vent Total: 4E04 uCi/secProjected Dose using CPAMPEDE 8.0PROJECTED DOSE I Mile 2 Mile 5 Mile 10 MileTEDE (Rem) 0.001 0.0005 0.0001 0.0000CDE (Rem) 0.006 0.002 0.0005 0.0002Plume Shine (R/hr) 0.00 0.00 0.00 0.00Plant Vent (WRGM) ALERTPlant Vent (WRGM) PVF-684 + PVF-68510 mRem TEDE or 50 mRem thyroid CDEHigh Alarm Setpoint: 2.0E04 uCi/secAssumptionsX-RE-5570 A+BCPAMPEDEValue UI I Value U22E5 uCi/Sec 2E5 uCi/SecPlant Vent Total: 4.0E05 uCi/secProjected Dose using CPAMPEDE 8.0PROJECTED DOSE I Mile 2 Mile 5 Mile 10 MileTEDE (Rem) 0.013 0.005 0.001 0.0005CDE (Rem) 0.061 0.022 0.006 0.002Plume Shine (R/hr) 0.001 0.00 0.00 0.004 of 706/18/2015 EAL Section R Revision 6Table R-1 Effluent Monitor Classification ThresholdsPlant Vent (WRGM) SAE AssumptionsPlant Vent (WRGM) PVF-684 + PVF-685 X-RE-5570 A+B100 mRem TEDE or 500 mRem thyroid CDEHigh Alarm Setpoint: 2.0E04 uCi/secICPAMPEDEValue U I Value U22E6 uCi/Sec 2E6 uCi/SecPlant Vent Total: 4E06 uCi/secProjected Dose using CPAMPEDE 8.0PROJECTED DOSE I Mile 2 Mile 5 Mile 10 MileTEDE (Rem) 0.132 0.047 0.012 0.005CDE (Rem) 0.577 0.208 0.054 0.020Plume Shine (R/hr) 0.010 0.003 0.001 0.00Plant Vent (WRGM) GE AssumptionsPlant Vent (WRGM) PVF-684 + PVF-685 X-RE-5570 A+B1000 mRem TEDE or 5000 mRem thyroid CDEHigh Alarm Setpoint: 2.0E04 uCi/secCPAMPEDEValue U 1 Value U22E7 uCi/Sec 2E7 uCi/SecPlant Vent Value: 4.0E07 uCi/secProjected Dose using CPAMPEDE 8.0PROJECTED DOSE 1 Mile 2 Mile 5 Mile 10 MileTEDE (Rem) 1.320 0.474 0.124 0.045CDE (Rem) 5.780 2.080 0.542 0.1 99Plume Shine (R/hr) 0.102 10.037 0.010 0.0045 of 706/18/2015 EAL Section R Revision 6Table R-1 Effluent Monitor Classification ThresholdsMain Steam UE AssumptionsThe setpoints differ on the main steam lines so an absolute number cannot be derived.Following are the current setpoints from the PC-1 1:Main Steam Line MonitorHigh (uCi/cc) [ Alert(uCi/ec)2.OE-01 1.OE-01Dose Projection AssumptionsThe dose projections used the following:Only one unit was used to make dose projections.One mile was used for Exclusion Area Boundary.Reactor has been Shutdown for one hour.The event start time is same as Reactor Shutdown.Accident Type for Main Steam Line monitors release was a SGTR.Flow Rate 120,000 lb/hrFiltration: NoStability Class: DWind Speed: 10 miles per hourMain Steam UE AssumptionsHigh Alarm Setpoint: 2.OE-1 uCi/ccMSL Total: 0.4 uCi/ccProjected Dose using CPAMPEDE 8.0PROJECTED DOSE 1 Mile 2 Mile 5 Mile 10 MileTEDE (Rem) 0.002 0.0006 0.0002 0.0001CDE (Rem) 0.020 0.007 0.002 0.0007Plume Shine (R/hr) 0.00 0.00 0.00 0.006 of 706/18/2015 EAL Section R Revision 6Table R-1 Effluent Monitor Classification ThresholdsMain Steam ALERT Assumptions10 mRem TEDE or 50 mRem thyroid CDEPlant Vent Value: 0.9 uCi/ccProjected Dose using CPAMPEDE 8.0PROJECTED DOSE I Mile 2 Mile 5 Mile 10 MileTEDE (Rem) 0.004 0.001 0.0004 0.0001CDE (Rem) 0.046 0.017 0.004 0.002Plume Shine (R/hr) 0.003 0.001 0.00 0.00Main Steam SAE Assumptions100 mRem TEDE or 500 mRem thyroid CDEPlant Vent Value: 9.0 uCi/ccProjected Dose using CPAMPEDE 8.0PROJECTED DOSE I Mile 2 Mile 5 Mile 10 MileTEDE (Rem) 0.041 0.015 0.004 0.001CDE (Rem) 0.461 0.166 0.043 0.016Plume Shine (R/hr) 0.002 0.001 10.00 0.00Main Steam GE Assumptions1000 mRem TEDE or 5000 mRem thyroid CDEPlant Vent Value: 90 uCi/ecProjected Dose using CPAMPEDE 8.0PROJECTED DOSE I Mile 2 Mile 5 Mile 10 MileTEDE (Rem) 0.405 0.146 0.038 0.014CDE (Rem) 4.610 1.660 0.433 0.158Plume Shine (R/hr) 0.020 0.007 10.002 0.0017 of 706/18/2015 ATTACHMENT 6 TO TXX-15101EMERGENCY ACTION LEVELWALLCHARTSFOR CPNPP(3 PAGES)}}