ML18094A155

Enclosures 2 - 6: Proposed EAL Technical Basis Document, NEI 99-01. Rev. 6., Proposed EAL Matrix Chart/Review Table and Supporting Referenced Document Pages

ML18094A155
Person / Time
Site:	Arkansas Nuclear
Issue date:	03/29/2018
From:	Entergy Operations
To:	Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
Shared Package
ML18094A154	List: 0CAN031801, Unit 2 - License Amendment Request - Adoption of Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6 ML18094A155
References
OCAN031801
Download: ML18094A155 (703)

Text

Enclosure 2 to 0CAN031801 Proposed EAL Technical Basis Document (Markup) to 0CAN031801 Page 1 of 255 Table of Contents Section Page

1.0 INTRODUCTION

.................................................................................................................2 2.0 DISCUSSION ......................................................................................................................2 2.1 Background ................................................................................................................2 2.2 Fission Product Barriers .............................................................................................3 2.3 Fission Product Barrier Classification Criteria ............................................................3 2.4 EAL Organization .......................................................................................................4 2.5 Technical Basis Information .......................................................................................5 2.6 Operations Mode Applicability ....................................................................................7 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS ..........................................8 3.1 General Considerations .............................................................................................8 3.2 Classification Methodology ......................................................................................10

4.0 REFERENCES

..................................................................................................................13 4.1 Developmental .........................................................................................................13 4.2 Implementing............................................................................................................13 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS ...........................................................13 5.1 Definitions (ref. 4.1.1 except as noted) ....................................................................13 5.2 Abbreviations/Acronyms ..........................................................................................18 6.0 ANO-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE ..................................................22

7.0 ATTACHMENTS ...............................................................................................................24 7.1 Attachment 1, Emergency Action Level Technical Bases ........................................24 Category A - Abnormal Rad Levels / Rad Effluents ................................................25 Category C - Cold Shutdown / Refueling System Malfunction ................................70 Category E - Independent Spent Fuel Storage Installation (ISFSI) .......................109 Category F - Fission Product Barrier Degradation ................................................112 Table 1[2]F-1, Fission Product Barrier Threshold Matrix & Bases ...119 Category H - Hazards and Other Conditions Affecting Plant Safety .....................164 Category S - System Malfunction ..........................................................................210 7.2 Attachment 2, Safe Operation & Shutdown Areas Tables 1[2]A-3 & 1[2]H-2 Bases .....................................................253 to 0CAN031801 Page 2 of 255

1.0 INTRODUCTION

This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for Arkansas Nuclear One (ANO). It should be used to facilitate review of the ANO EALs and provide historical documentation for future reference. Decision-makers responsible for implementation of 1903.010, Emergency Action Level Classification, may use this document as a technical reference in support of EAL interpretation.

This information may assist the Emergency Director in making classifications, particularly those involving judgment or multiple events. The basis information may also be useful in training and for explaining event classifications to off-site officials.

The expectation is that emergency classifications are to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes or less in all cases when conditions are present and have been recognized. Use of this document for assistance is not intended to delay the emergency classification.

Because the information in a basis document can affect emergency classification decision-making (e.g., the Emergency Director refers to it during an event), the NRC staff expects that changes to the basis document will be evaluated in accordance with the provisions of 10 CFR 50.54(q). 2.0 DISCUSSION 2.1 Background EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the ANO Emergency Plan.

In 1992, the NRC endorsed NUMARC/NESP-007 "Methodology for Development of Emergency Action Levels" as an alternative to NUREG-0654 EAL guidance. NEI 99-01 (NUMARC/NESP-007) Revisions 4 and 5 were subsequently issued for industry implementation. Enhancements over earlier revisions included:

Consolidating the system malfunction initiating conditions and example emergency action levels which address conditions that may be postulated to occur during plant shutdown conditions. Initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent 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 to numerous implementation issues including the NRC EAL Frequently Asked Questions (FAQs). Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors," November 2012 (ref. 4.1.1), ANO conducted an EAL implementation upgrade project that produced the EALs discussed herein. to 0CAN031801 Page 3 of 255 2.2 Fission Product Barriers Fission product barrier thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment. This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment.

Many of the EALs derived from the NEI methodology are fission product barrier threshold based. That is, the conditions that define the EALs are based upon thresholds that represent the 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 a greater probability of barrier loss and reduced certainty of maintaining the barrier. The primary fission product barriers are:

A. Fuel Clad Barrier (FCB): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets. B. Reactor Coolant System Barrier (RCB): The Reactor Coolant System Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves. C. Containment Barrier (CNB): The Containment Barrier includes the Reactor Building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the Reactor Building up to and including the outermost secondary side isolation valve.

Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (ECL) from an Alert to a Site Area Emergency or a General Emergency. 2.3 Fission Product Barrier Classification Criteria

The following criteria are the bases for event classification related to fission product barrier loss or potential loss: Alert: Any loss or any potential loss of either Fuel Clad or RCS Barrier Site Area Emergency: Loss or potential loss of any two barriers General Emergency: Loss of any two barriers and loss or potential loss of the third barrier to 0CAN031801 Page 4 of 255 2.4 EAL Organization The ANO EAL scheme includes the following features:

Division of the EAL set into three broad groups: o EALs applicable under any plant operating modes - This group would be reviewed by the EAL-user any time emergency classification is considered. o EALs applicable only under hot operating modes - This group would only be reviewed 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 be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling or Defueled mode. The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition. This approach significantly minimizes the total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency. Within each group, assignment of EALs to categories and subcategories: Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. The ANO EAL categories are aligned to and represent the NEI 99-01, "Recognition Categories." Subcategories are used in the ANO scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The ANO EAL categories and subcategories are listed below.

The primary tool for determining the emergency classification level is the EAL Classification Matrix. The user of the EAL Classification Matrix may (but is not required to) consult the EAL technical bases in order to obtain additional information concerning the EALs under classification consideration. The user should consult Section 3.0 and Attachment 1 of this document for such information.

to 0CAN031801 Page 5 of 255 EAL Groups, Categories and Subcategories EAL Group/Category EAL Subcategory Any Operating Mode: A - Abnormal Rad Levels / Rad Effluent 1 - Radiological Effluent 2 - Irradiated Fuel Event 3 - Area Radiation Levels H - Hazards and Other Conditions Affecting Plant Safety 1 - Security 2 - Seismic Event 3 - Natural or Technological Hazard 4 - Fire 5 - Hazardous Gas 6 - Control Room Evacuation 7 - Emergency Director Judgment E - Independent Spent Fuel Storage Installation (ISFSI) 1 - Confinement Boundary Hot Conditions: S - System Malfunction 1 - Loss of Essential AC Power 2 - Loss of Vital DC Power 3 - Loss of Control Room Indications 4 - RCS Activity 5 - RCS Leakage 6 - RPS Failure 7 - Loss of Communications 8 - Containment Failure 9 - Hazardous Event Affecting Safety Systems F - Fission Product Barrier Degradation None Cold Conditions: C - Cold Shutdown / Refueling System Malfunction 1 - RCS Level 2 - Loss of Essential AC Power 3 - RCS Temperature 4 - Loss of Vital DC Power 5 - Loss of Communications 6 - Hazardous Event Affecting Safety Systems 2.5 Technical Bases Information EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (Any, Hot, Cold), EAL category (A, C, E, F, H and S) and EAL subcategory. A summary explanation of each category and subcategory is given at the beginning of the technical bases discussions of the EALs included in the category. For each EAL, the following information is provided:

to 0CAN031801 Page 6 of 255 Category Letter & Title Subcategory Number & Title Initiating 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 the emergency classification to onsite and offsite personnel. Four characters define each EAL identifier: 1. First character (letter): Corresponds to the EAL category as described above (A, C, E, F, H or S) 2. Second character (letter): The emergency classification (G, S, A or U) G = General Emergency S = Site Area Emergency A = Alert U = Unusual Event 3. Third character (number): Subcategory number within the given category. Subcategories are sequentially numbered beginning with the number one (1). If a category 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 EAL subcategory. 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 Matrix. If an ANO Unit 2 EAL threshold value differs from Unit 1, the Unit 2 threshold is enclosed in brackets. For example, in the EAL threshold "RVLMS Levels 1 through 8 indicate DRY [RVLMS Levels 1 through 5 indicate DRY]", "RVLMS Levels 1 through 5 indicate DRY" apply only to Unit 2. Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling, DEF - Defueled, or All. (See Section 2.6 for operating mode definitions).

to 0CAN031801 Page 7 of 255 Definitions: If the EAL wording contains a defined term, the definition of the term is included in this section. These definitions can also be found in Section 5.1. Basis: An EAL basis section that provides ANO-relevant information concerning the EAL as well as a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6. Reference(s): Source documentation from which the EAL is derived. 2.6 Operating Mode Applicability Unit 1 (ref. 4.1.6):

1 Power Operation Keff 0.99, reactor power > 5% 2 Startup Keff 0.99, reactor power 5% 3 Hot Standby Keff < 0.99, reactor coolant temperature 280°F 4 Hot Shutdown Keff < 0.99, reactor coolant temperature 280°F > Tavg > 200°F and all reactor vessel head closure bolts fully tensioned 5 Cold Shutdown Keff < 0.99, reactor coolant temperature 200°F and all reactor vessel head closure bolts fully tensioned 6 Refueling One or more reactor vessel head closure bolts less than fully tensioned DEF Defueled All fuel assemblies have been removed from Containment and placed in the spent fuel pool.

to 0CAN031801 Page 8 of 255 Unit 2 (ref. 4.1.6): 1 Power Operation Keff 0.99, reactor power > 5%, average coolant temperature 300°F 2 Startup Keff 0.99, reactor power 5%, average coolant temperature 300°F 3 Hot Standby Keff < 0.99, average coolant temperature 300°F 4 Hot Shutdown Keff < 0.99, average coolant temperature 300°F > Tavg > 200°F 5 Cold Shutdown Keff < 0.99, average coolant temperature 200°F 6 Refueling Keff 0.95, average coolant temperature 140°F, reactor vessel head unbolted or removed, and fuel in the vessel. DEF Defueled All fuel assemblies have been removed from Containment and placed in the spent fuel pool.

The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action being initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred.

3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 3.1 General Considerations When making an emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes, and the informing basis information. In the Recognition Category F matrices, EALs are based on loss or potential loss of Fission Product Barrier Thresholds.

EAL matrices should be read from left to right, from General Emergency to Unusual Event, and top to bottom. Declaration decisions should be independently verified before declaration is made except when gaining this verification would exceed the 15 minute declaration requirement. Place keeping should be used on all EAL matrices. to 0CAN031801 Page 9 of 255 3.1.1 Classification Timeliness NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (ref. 4.1.8). 3.1.2 Valid Indications

All emergency classification assessments shall be based upon valid indications, reports or conditions. A valid indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy. For example, verification could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel.

An indication, report, or condition is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

3.1.3 Imminent Conditions For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.

3.1.4 Planned vs. Unplanned Events A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that: 1) the activity proceeds as planned, and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 CFR 50.72 (ref. 4.1.4).

3.1.5 Classification Based on Analysis The assessment of some EALs is based on the results of analyses that are necessary to ascertain 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 the to 0CAN031801 Page 10 of 255 associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift).

3.1.6 Emergency Director Judgment

While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-01 EAL scheme provides the Emergency Director with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to Category H). The Emergency Director will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL 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 Methodology To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, the associated IC is likewise met, the emergency classification process "clock" starts, and the ECL must be declared in accordance 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. 4.1.8).

3.2.1 Classification of Multiple Events and Conditions When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared. For example: If an Alert EAL and a Site Area Emergency EAL are met, whether at one unit or at two 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 units, an Alert should be declared. Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events (ref. 4.1.2). to 0CAN031801 Page 11 of 255 3.2.2 Consideration of Mode Changes During Classification The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or 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 emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition. For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher. 3.2.3 Classification of Imminent Conditions Although EALs provide specific thresholds, the Emergency Director must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT). If, in the judgment of the Emergency Director, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met. While applicable to all emergency classification levels, this approach is particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures.

3.2.4 Emergency Classification Level Upgrading and Termination An ECL may be terminated when the event or condition that meets the classified IC and EAL no longer exists, and other site-specific termination requirements are met. As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02 (ref. 4.1.2). 3.2.5 Classification of Short-Lived Events Event-based ICs and EALs define a variety of specific occurrences that have potential or actual 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 occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration. Examples of such events include an earthquake or a failure of the reactor protection system to automatically trip the reactor followed by a successful manual trip.

3.2.6 Classification of Transient Conditions Many of the ICs and/or EALs employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions. to 0CAN031801 Page 12 of 255 EAL momentarily met during expected plant response - In instances in which an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures.

EAL momentarily met but the condition is corrected prior to an emergency declaration - If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required. For illustrative purposes, consider the following example:

An ATWS occurs and the high pressure ECCS systems fail to automatically start. The plant enters an inadequate core cooling condition (a potential loss of both the Fuel Clad and RCS Barriers). If an operator manually starts a high pressure ECCS system in accordance with an EOP step and clears the inadequate core cooling condition prior to an emergency declaration, then the classification should be 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 allow the 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 an operator is able to take a successful corrective action prior to the Emergency Director completing the review and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.

3.2.7 After-the-Fact Discovery of an Emergency Event or Condition

In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process. In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 (ref. 4.1.3) is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50.72 (ref. 4.1.4) within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements. 3.2.8 Retraction of an Emergency Declaration

Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022 (ref. 4.1.3). to 0CAN031801 Page 13 of 255

4.0 REFERENCES

4.1 Developmental 4.1.1 NEI 99-01 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML12326A805 4.1.2 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007. 4.1.3 NUREG-1022 Event Reporting Guidelines: 10CFR50.72 and 50.73 4.1.4 10 § CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors 4.1.5 10 § CFR 50.73 License Event Report System 4.1.6 Unit 1[2] Technical Specifications Table 1.1-1[1.1], Modes[Operational Modes]

4.1.7 Arkansas Nuclear One Offsite Dose Calculation Manual (ODCM) 4.1.8 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.9 Arkansas Nuclear One Emergency Plan 4.1.10 1015.008 Unit 2 SDC Control 4.2 Implementing 4.2.1 1903.010 Emergency Action Level Classification 4.2.2 NEI 99-01 Rev. 6 to ANO EAL Comparison Matrix 4.2.3 ANO EAL Matrix 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS 5.1 Definitions (ref. 4.1.1 except as noted)

Selected terms used in Initiating Condition, Emergency Action Level statements and EAL bases are set in all capital letters (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document. The definitions of these terms are provided below. Alert Events are in progress, or have occurred, which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION.

Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

to 0CAN031801 Page 14 of 255 Confinement Boundary The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the ANO ISFSI, the Confinement Boundary is comprised of either the Multi-assembly Sealed Basket (MSB) (SNC System) or Multi-Purpose Canister (MPC) (Holtec System). Containment Closure The procedurally defined actions taken to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown existing plant conditions (ref. 4.1.10). As applied to ANO, Containment Closure must be capable of being set within 30 minutes. Containment Closure is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent. Emergency Action Level (EAL) A pre-determined, site-specific, observable threshold for an INITIATING CONDITION that, when met or exceeded, places the plant in a given emergency classification level. Emergency Classification Level (ECL) One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are:

Notification of Unusual Event (NOUE) Alert Site Area Emergency (SAE) General Emergency (GE) Explosion A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines 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.

Faulted The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized. to 0CAN031801 Page 15 of 255 Fire Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. Fission Product Barrier Threshold A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier. Flooding A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. General Emergency Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

Hostage A person(s) held as leverage against the station to ensure that demands will be met by the station.

Hostile Action An act toward a NPP ANO or its personnel that includes the use of violent force 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 other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.

Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPPANO. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA).

Hostile Force One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction. Imminent The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. to 0CAN031801 Page 16 of 255 Impede(d) Personnel access to a room or area is hindered to an extent that extraordinary measures 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). Independent Spent Fuel Storage Installation (ISFSI) A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. Initiating Condition (IC) An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences. Normal Levels As applied to radiological IC/EALs, the highest reading in the past twenty-four hours excluding the current peak value. Owner Controlled Area (OCA) For the purposes of classification this is the Security Owner Controlled Area (SOCA). The SOCA is demarcated as a Vehicle Barrier System (VBS) and a detection fence on the outside and a delay fence on the inside of the passive and active barriers. The SOCA is the area inside the SOCA VBS up to the PROTECTED AREA fence line. Access to this area is controlled by the SOCA Personnel Access Control Point (ref. 4.1.9). Projectile An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

Protected Area An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled (ref. 4.1.9). RCS Intact The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams). Refueling Pathway All the cavities, tubes, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel. to 0CAN031801 Page 17 of 255 Ruptured The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection (automatic or manual). Safety System A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the 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 result in potential offsite exposures. Security Condition Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A Security Condition does not involve a HOSTILE ACTION.

Site Area Emergency Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; (1) toward site personnel or equipment that could lead to the likely failure of or; (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the SITE BOUNDARY.

Site Boundary That boundary defined by a 1046 meter (0.65 mile) radius around the plant (ref. 4.1.7). 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 intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. to 0CAN031801 Page 18 of 255 Unusual Event Events are in progress or have occurred which indicate a potential degradation in the level of 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 expected unless further degradation of SAFETY SYSTEMS occurs.

Valid An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Visible Damage Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train.Damage 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 the operability or reliability of the affected component or structure. 5.2 Abbreviations/Acronyms °F .................................................................................................................... Degrees Fahrenheit ° ......................................................................................................................................... Degrees AC ..................................................................................................................... Alternating Current ANO ............................................................................................................ Arkansas Nuclear One AOP .............................................................................................. Abnormal Operating Procedure ATWS .................................................................................... Anticipated Transient Without Scram BMS ................................................................................................... Boron Management System BWST ................................................................................................ Borated Water Storage Tank CDE ................................................................................................... Committed Dose Equivalent CET .......................................................................................................... Core Exit Thermocouple CFR ................................................................................................... Code of Federal Regulations CIAS .................................................................................. Containment Isolation Actuation Signal CMT, CNTMT, CTMT .................................................................................................. Containment CNB ................................................................................................................ Containment Barrier DBA ............................................................................................................. Design Basis Accident DBE ......................................................................................................... Design Basis Earthquake to 0CAN031801 Page 19 of 255 DC ............................................................................................................................. Direct Current DEF ................................................................................................................................... Defueled D/G ....................................................................................................................... Diesel Generator DHR .............................................................................................................. Decay Heat Removal DROPS ............................................................ Diverse Reactor Overpressure Protection System DSC ............................................................................................................. Dry Shielded Canister DSS ............................................................................................................ Diverse Scram System EAL .......................................................................................................... Emergency Action Level ECCS ......................................................................................... Emergency Core Cooling System ECL ............................................................................................... Emergency Classification Level DEF ................................................................................................................................... Defueled ENS ............................................................................................... Emergency Notification System EOF ................................................................................................ Emergency Operations Facility EOP ........................................................................................... Emergency Operating Procedure EPA ............................................................................................ Environmental Protection Agency ERG ............................................................................................ Emergency Response Guideline EPIP ............................................................................. Emergency Plan Implementing Procedure ESAS ........................................................................... Engineered Safeguards Actuation System ESF ...................................................................................................... Engineered Safety Feature ESFAS .................................................................. Engineered Safety Features Actuation System FAA ............................................................................................... Federal Aviation Administration FBI ................................................................................................ Federal Bureau of Investigation FCB ...................................................................................................................... Fuel Clad Barrier FEMA ............................................................................ Federal Emergency Management Agency GE ................................................................................................................... General Emergency HPI ............................................................................................................. High Pressure Injection IC ...................................................................................................................... Initiating Condition IPEEE ............................. Individual Plant Examination of External Events (Generic Letter 88-20)

ISFSI ......................................................................... Independent Spent Fuel Storage Installation Keff ...................................................................................... Effective Neutron Multiplication Factor LCO ................................................................................................ Limiting Condition of Operation LER ............................................................................................................. Licensee Event Report LOCA ...................................................................................................... Loss of Coolant Accident LRW .................................................................................................................... Liquid Rad Waste to 0CAN031801 Page 20 of 255 LTOP ............................................................................................ Low Temperature Overpressure LWR ................................................................................................................ Light Water Reactor MCC .............................................................................................................. Motor Control Center MPC ................................................ Maximum Permissible Concentration/Multi-Purpose Canister mR, mRem, mrem, mREM ............................................................ milli-Roentgen Equivalent Man MSL ...................................................................................................................... Main Steam Line MTS ................................................................................................................ Margin to Saturation MW .................................................................................................................................. Megawatt NDTT ...................................................................................... Nil Ductility Transition Temperature NEI ............................................................................................................ Nuclear Energy Institute NEIC ................................................................................ National Earthquake Information Center NESP ................................................................................ National Environmental Studies Project NORAD ................................................................ North American Aerospace Defense Command NOT .............................................................................................. Normal Operating Temperature (NO)UE ............................................................................................. Notification of Unusual Event NPP ................................................................................................................ Nuclear Power Plant NRC ............................................................................................ Nuclear Regulatory Commission NSSS ............................................................................................. Nuclear Steam Supply System OBE ................................................................................................... Operating Basis Earthquake OCA ........................................................................................................... Owner Controlled Area ODCM ........................................................................................ Off-site Dose Calculation Manual ORO ............................................................................................... Offsite Response Organization PA ........................................................................................................................... Protected Area PAG ..................................................................................................... Protective Action Guideline PRA/PSA .................................. Probabilistic Risk Assessment / Probabilistic Safety Assessment P-T .............................................................................................................. Pressure-Temperature PTS ..................................................................................................... Pressurized Thermal Shock PWR ..................................................................................................... Pressurized Water Reactor PSIG ............................................................................................ Pounds per Square Inch Gauge R ..................................................................................................................................... Roentgen RB ......................................................................................................................... Reactor Building RCC ......................................................................................................... Reactor Control Console RCB ............................................................................................. Reactor Coolant System Barrier RCP ............................................................................................................ Reactor Coolant Pump to 0CAN031801 Page 21 of 255 RCS ......................................................................................................... Reactor Coolant System Rem, rem, REM ..................................................................................... Roentgen Equivalent Man Rep CET ....................................................................... Representative Core Exit Thermocouples RETS ...................................................................... Radiological Effluent Technical Specifications RPS ...................................................................................................... Reactor Protection System RV ........................................................................................................................... Reactor Vessel RVLMS ........................................................................... Reactor Vessel Level Monitoring System RWT ............................................................................................................. Refueling Water Tank SAR ............................................................................................................. Safety Analysis Report SBO ...................................................................................................................... Station Blackout SCBA ................................................................................... Self-Contained Breathing Apparatus SDC ................................................................................................................... Shutdown Cooling SOCA ........................................................................................... Security Owner Controlled Area SG ....................................................................................................................... Steam Generator SI ............................................................................................................................ Safety Injection SPDS ........................................................................................ Safety Parameter Display System SPING ..................................................................................... Super Particulate Iodine Noble Gas SRO ......................................................................................................... Senior Reactor Operator TEDE ............................................................................................ Total Effective Dose Equivalent TOAF ................................................................................................................. Top of Active Fuel TSC ........................................................................................................ Technical Support Center USGS .......................................................................................... United States Geological Survey VBS ............................................................................................................ Vehicle Barrier System

to 0CAN031801 Page 22 of 255 6.0 ANO-TO-NEI 99-01 REV. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of an ANO EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the ANO EALs based on the NEI guidance can be found in the EAL Comparison Matrix.

ANO NEI 99-01 Rev. 6 EAL IC Example EAL AU1.1 AU1 1, 2 AU1.2 AU1 3 AU2.1 AU2 1 AA1.1 AA1 1 AA1.2 AA1 2 AA1.3 AA1 3 AA1.4 AA1 4 AA2.1 AA2 1 AA2.2 AA2 2 AA2.3 AA2 3 AA3.1 AA3 1 AA3.2 AA3 2 AS1.1 AS1 1 AS1.2 AS1 2 AS1.3 AS1 3 AS2.1 AS2 1 AG1.1 AG1 1 AG1.2 AG1 2 AG1.3 AG1 3 AG2.1 AG2 1 CU1.1 CU1 1 CU1.2 CU1 2 CU2.1 CU2 1 CU3.1 CU3 1 CU3.2 CU3 2 CU4.1 CU4 1 to 0CAN031801 Page 23 of 255 ANO NEI 99-01 Rev. 6 EAL IC Example EAL CU5.1 CU5 1, 2, 3 CA1.1 CA1 1 CA1.2 CA1 2 CA2.1 CA2 1 CA3.1 CA3 1, 2 CA6.1 CA6 1 CS1.1 CS1 1 CS1.2 CS1 2 CS1.3 CS1 3 CG1.1 CG1 1 CG1.2 CG1 2 EU1.1 EU1 1 FA1.1 FA1 1 FS1.1 FS1 1 FG1.1 FG1 1 HU1.1 HU1 1, 2, 3 HU2.1 HU2 1 HU3.1 HU3 1 HU3.2 HU3 2 HU3.3 HU3 3 HU3.4 HU3 4 HU4.1 HU4 1 HU4.2 HU4 2 HU4.3 HU4 3 HU4.4 HU4 4 HU7.1 HU7 1 HA1.1 HA1 1, 2 HA5.1 HA5 1 HA6.1 HA6 1 HA7.1 HA7 1 HS1.1 HS1 1 to 0CAN031801 Page 24 of 255 ANO NEI 99-01 Rev. 6 EAL IC Example EAL HS6.1 HS6 1 HS7.1 HS7 1 HG7.1 HG7 1 SU1.1 SU1 1 SU3.1 SU2 1 SU4.1 SU3 1 SU4.2 SU3 2 SU5.1 SU4 1, 2, 3 SU6.1 SU5 1 SU6.2 SU5 2 SU7.1 SU6 1, 2, 3 SU8.1 SU7 1, 2 SA1.1 SA1 1 SA3.1 SA2 1 SA6.1 SA5 1 SA9.1 SA9 1 SS1.1 SS1 1 SS2.1 SS8 1 SS6.1 SS5 1 SG1.1 SG1 1 SG1.2 SG8 1 7.0 ATTACHMENTS 7.1 Attachment 1, Emergency Action Level Technical Bases 7.2 Attachment 2, Safe Operation & Shutdown Areas Tables 1[2]A-3 & 1[2]H-2 Bases

to 0CAN031801 Page 25 of 255 Attachment 1 - Emergency Action Level Technical Bases Category A - Abnormal Rad Levels / Rad Effluent EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via non-radiological symptoms. Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emergency classification.

At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety.

Events of this category pertain to the following subcategories: 1. Radiological Effluent Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits. 2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification. 3. Area Radiation Levels Sustained general area radiation levels which may preclude access to areas requiring continuous occupancy also warrant emergency classification.

to 0CAN031801 Page 26 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer EAL: AU1.1 Unusual Event Reading on any Table 1[2]A-1 effluent radiation monitor > column "UE" for 60 min. (Notes 1, 2, 3) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Table 1A-1 Unit 1 Effluent Monitor Classification Thresholds (2 min. avg reading) Release Point Monitor GE SAE Alert UE Gaseous Containment Purge RX-9820 (SPING 1) 4.15E+01 Ci/cc 4.15E+00 µCi/cc4.15E-01 µCi/cc 1.21E-03 µCi/cc Radwaste Area RX-9825 (SPING 2) 2.67E+01 Ci/cc 2.67E+00 µCi/cc2.67E-01 µCi/cc 4.94E-04 µCi/cc Fuel Handling Area RX-9830 (SPING 3) 6.20E+02 Ci/cc 6.20E+01 µCi/cc6.20E+00 µCi/cc 5.44E-04 µCi/cc Emergency Penetration Room RX-9835 (SPING 4) 6.55E+02 Ci/cc 6.55E+01 µCi/cc6.55E+00 µCi/cc 1.21E-02 µCi/cc Liquid Liquid Radwaste RE-4642 ---- ---- ---- 2.46E+05 cpm to 0CAN031801 Page 27 of 255 Attachment 1 - Emergency Action Level Technical Bases Table 2A-1 Unit 2 Effluent Monitor Classification Thresholds (2 min. avg reading) Release Point Monitor GE SAE Alert UE Gaseous Containment Purge 2RX-9820 (SPING 5) 1.88E+01 Ci/cc 1.88E+00 µCi/cc1.88E-01 µCi/cc 5.48E-04 µCi/cc Radwaste Area 2RX-9825 (SPING 6) 2.35E+01 Ci/cc 2.35E+00 µCi/cc2.35E-01 µCi/cc 4.35E-04 µCi/cc Fuel Handling Area 2RX-9830 (SPING 7) 6.86E+02 Ci/cc 6.86E+01 µCi/cc6.86E+00 µCi/cc 6.04E-04 µCi/cc Emergency Penetration Room 2RX-9835 (SPING 8) 5.88E+02 Ci/cc 5.88E+01 µCi/cc5.88E+00 µCi/cc 1.09E-02 µCi/cc Liquid BMS Liquid Discharge 2RE-2330 ---- ---- ---- 2.45E+04 cpm Regenerative Waste Discharge 2RE-4423 ---- ---- ---- 2.45E+05 cpm Mode Applicability: All Definition(s): VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This IC addresses a potential decrease reduction 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 is normally prepared.

Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls. to 0CAN031801 Page 28 of 255 Attachment 1 - Emergency Action Level Technical Bases Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. EAL #1 - This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways . EAL #2 - This EAL addressesas well as radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL will Such releases are typically be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas). EAL #3 - This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.). Escalation of the emergency classification level would be via IC AA1.

Reference(s):

1. OP-1604.051 Eberline Radiation Monitor System 2. Offsite Dose Calculation Manual

3. EP-CALC-ANO-1701 Radiological Effluent EAL Values

4. NEI 99-01 AU1 to 0CAN031801 Page 29 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer EAL: AU1.2 Unusual Event Sample analysis for a gaseous or liquid release indicates a concentration or release rate 2 x ODCM limits for 60 min. (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All

Definition(s): None Basis: This IC addresses a potential decrease reduction 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 is normally prepared. Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. to 0CAN031801 Page 30 of 255 Attachment 1 - Emergency Action Level Technical Bases Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. EAL #1 - This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways. EAL #2 - This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL will typically be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas). EAL #3 - This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.).

Escalation of the emergency classification level would be via IC AA1.

Reference(s): 1. Offsite Dose Calculation Manual

2. NEI 99-01 AU1 to 0CAN031801 Page 31 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL: AA1.1 Alert Reading on any Table 1[2]A-1 effluent radiation monitor > column "ALERT" for 15 min. (Notes 1, 2, 3, 4) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Note 4 The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Table 1A-1 Unit 1 Effluent Monitor Classification Thresholds (2 min. avg reading) Release Point Monitor GE SAE Alert UE Gaseous Containment Purge RX-9820 (SPING 1) 4.15E+01 Ci/cc 4.15E+00 µCi/cc4.15E-01 µCi/cc 1.21E-03 µCi/cc Radwaste Area RX-9825 (SPING 2) 2.67E+01 Ci/cc 2.67E+00 µCi/cc2.67E-01 µCi/cc 4.94E-04 µCi/cc Fuel Handling Area RX-9830 (SPING 3) 6.20E+02 Ci/cc 6.20E+01 µCi/cc6.20E+00 µCi/cc 5.44E-04 µCi/cc Emergency Penetration Room RX-9835 (SPING 4) 6.55E+02 Ci/cc 6.55E+01 µCi/cc6.55E+00 µCi/cc 1.21E-02 µCi/cc Liquid Liquid Radwaste RE-4642 ---- ---- ---- 2.46E+05 cpm to 0CAN031801 Page 32 of 255 Attachment 1 - Emergency Action Level Technical Bases Table 2A-1 Unit 2 Effluent Monitor Classification Thresholds (2 min. avg reading) Release Point Monitor GE SAE Alert UE Gaseous Containment Purge 2RX-9820 (SPING 5) 1.88E+01 Ci/cc 1.88E+00 µCi/cc1.88E-01 µCi/cc 5.48E-04 µCi/cc Radwaste Area 2RX-9825 (SPING 6) 2.35E+01 Ci/cc 2.35E+00 µCi/cc2.35E-01 µCi/cc 4.35E-04 µCi/cc Fuel Handling Area 2RX-9830 (SPING 7) 6.86E+02 Ci/cc 6.86E+01 µCi/cc6.86E+00 µCi/cc 6.04E-04 µCi/cc Emergency Penetration Room 2RX-9835 (SPING 8) 5.88E+02 Ci/cc 5.88E+01 µCi/cc5.88E+00 µCi/cc 1.09E-02 µCi/cc Liquid BMS Liquid Discharge 2RE-2330 ---- ---- ---- 2.45E+04 cpm Regenerative Waste Discharge 2RE-4423 ---- ---- ---- 2.45E+05 cpm Mode Applicability: All Definition(s): VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. to 0CAN031801 Page 33 of 255 Attachment 1 - Emergency Action Level Technical Bases The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was 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 the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Escalation of the emergency classification level would be via IC AS1. Reference(s): 1. OP-1604.051 Eberline Radiation Monitor System 2. EP-CALC-ANO-1701 Radiological Effluent EAL Values

3. NEI 99-01 AA1 to 0CAN031801 Page 34 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL: AA1.2 Alert Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY. (Note 4) Note 4: The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Mode Applicability: All

Definition(s): SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant. Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the 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 CDE was 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 the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. to 0CAN031801 Page 35 of 255 Attachment 1 - Emergency Action Level Technical Bases Escalation of the emergency classification level would be via IC AS1. Reference(s): 1. OP-1904.002 Offsite Dose Projections

2. NEI 99-01 AA1 to 0CAN031801 Page 36 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL: AA1.3 Alert Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for 60 min. of exposure. (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All

Definition(s): SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant. Basis:

This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the 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 CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

to 0CAN031801 Page 37 of 255 Attachment 1 - Emergency Action Level Technical Bases Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.This EAL is assessed per the ODCM (ref. 2). Escalation of the emergency classification level would be via IC AS1.

Reference(s): 1. OP-1904.002 Offsite Dose Projections

2. Offsite Dose Calculation Manual

3. NEI 99-01 AA1 to 0CAN031801 Page 38 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL: AA1.4 Alert Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: Closed window dose rates > 10 mR/hr expected to continue for 60 min. Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation. (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All Definition(s):

SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant. Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

to 0CAN031801 Page 39 of 255 Attachment 1 - Emergency Action Level Technical Bases The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was 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 the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. Escalation of the emergency classification level would be via IC AS1.

Reference(s): 1. OP-1905.002 Offsite Emergency Monitoring 2. NEI 99-01 AA1 to 0CAN031801 Page 40 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL: AS1.1 Site Area Emergency Reading on any Table A-1 effluent radiation monitor > column "SAE" for 15 min. (Notes 1, 2, 3, 4) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Note 4: The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Table 1A-1 Unit 1 Effluent Monitor Classification Thresholds (2 min. avg reading) Release Point Monitor GE SAE Alert UE Gaseous Containment Purge RX-9820 (SPING 1) 4.15E+01 Ci/cc 4.15E+00 µCi/cc4.15E-01 µCi/cc 1.21E-03 µCi/cc Radwaste Area RX-9825 (SPING 2) 2.67E+01 Ci/cc 2.67E+00 µCi/cc2.67E-01 µCi/cc 4.94E-04 µCi/cc Fuel Handling Area RX-9830 (SPING 3) 6.20E+02 Ci/cc 6.20E+01 µCi/cc6.20E+00 µCi/cc 5.44E-04 µCi/cc Emergency Penetration Room RX-9835 (SPING 4) 6.55E+02 Ci/cc 6.55E+01 µCi/cc6.55E+00 µCi/cc 1.21E-02 µCi/cc Liquid Liquid Radwaste RE-4642 ---- ---- ---- 2.46E+05 cpm to 0CAN031801 Page 41 of 255 Attachment 1 - Emergency Action Level Technical Bases Table 2A-1 Unit 2 Effluent Monitor Classification Thresholds (2 min. avg reading) Release Point Monitor GE SAE Alert UE Gaseous Containment Purge 2RX-9820 (SPING 5) 1.88E+01 Ci/cc 1.88E+00 µCi/cc1.88E-01 µCi/cc 5.48E-04 µCi/cc Radwaste Area 2RX-9825 (SPING 6) 2.35E+01 Ci/cc 2.35E+00 µCi/cc2.35E-01 µCi/cc 4.35E-04 µCi/cc Fuel Handling Area 2RX-9830 (SPING 7) 6.86E+02 Ci/cc 6.86E+01 µCi/cc6.86E+00 µCi/cc 6.04E-04 µCi/cc Emergency Penetration Room 2RX-9835 (SPING 8) 5.88E+02 Ci/cc 5.88E+01 µCi/cc5.88E+00 µCi/cc 1.09E-02 µCi/cc Liquid BMS Liquid Discharge 2RE-2330 ---- ---- ---- 2.45E+04 cpm Regenerative Waste Discharge 2RE-4423 ---- ---- ---- 2.45E+05 cpm Mode Applicability: All Definition(s): VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

to 0CAN031801 Page 42 of 255 Attachment 1 - Emergency Action Level Technical Bases The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was 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 the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Escalation of the emergency classification level would be via IC AG1.

Reference(s):

1. OP-1604.051 Eberline Radiation Monitor System 2. EP-CALC-ANO-1701 Radiological Effluent EAL Values

3. NEI 99-01 AS1 to 0CAN031801 Page 43 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL: AS1.2 Site Area Emergency Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY. (Note 4) Note 4 The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Mode Applicability: All

Definition(s): SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant). Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the 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 CDE was 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 the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Escalation of the emergency classification level would be via IC AG1. to 0CAN031801 Page 44 of 255 Attachment 1 - Emergency Action Level Technical Bases Reference(s): 1. OP-1904.002 Offsite Dose Projections 2. NEI 99-01 AS1 to 0CAN031801 Page 45 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL: AS1.3 Site Area Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: Closed window dose rates > 100 mR/hr expected to continue for 60 min. Analyses of field survey samples indicate thyroid CDE > 500 mrem for 60 min. of inhalation. (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All Definition(s):

SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the 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 CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. to 0CAN031801 Page 46 of 255 Attachment 1 - Emergency Action Level Technical Bases Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. Escalation of the emergency classification level would be via IC AG1.

Reference(s): 1. OP-1905.002 Offsite Emergency Monitoring

2. NEI 99-01 AS1 to 0CAN031801 Page 47 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL: AG1.1 General Emergency Reading on any Table A-1 effluent radiation monitor > column "GE" for 15 min. (Notes 1, 2, 3, 4) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Note 4: The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Table 1A-1 Unit 1 Effluent Monitor Classification Thresholds (2 min. avg reading) Release Point Monitor GE SAE Alert UE Gaseous Containment Purge RX-9820 (SPING 1) 4.15E+01 Ci/cc 4.15E+00 µCi/cc4.15E-01 µCi/cc 1.21E-03 µCi/cc Radwaste Area RX-9825 (SPING 2) 2.67E+01 Ci/cc 2.67E+00 µCi/cc2.67E-01 µCi/cc 4.94E-04 µCi/cc Fuel Handling Area RX-9830 (SPING 3) 6.20E+02 Ci/cc 6.20E+01 µCi/cc6.20E+00 µCi/cc 5.44E-04 µCi/cc Emergency Penetration Room RX-9835 (SPING 4) 6.55E+02 Ci/cc 6.55E+01 µCi/cc6.55E+00 µCi/cc 1.21E-02 µCi/cc Liquid Liquid Radwaste RE-4642 ---- ---- ---- 2.46E+05 cpm to 0CAN031801 Page 48 of 255 Attachment 1 - Emergency Action Level Technical Bases Table 2A-1 Unit 2 Effluent Monitor Classification Thresholds (2 min. avg reading) Release Point Monitor GE SAE Alert UE Gaseous Containment Purge 2RX-9820 (SPING 5) 1.88E+01 Ci/cc 1.88E+00 µCi/cc1.88E-01 µCi/cc 5.48E-04 µCi/cc Radwaste Area 2RX-9825 (SPING 6) 2.35E+01 Ci/cc 2.35E+00 µCi/cc2.35E-01 µCi/cc 4.35E-04 µCi/cc Fuel Handling Area 2RX-9830 (SPING 7) 6.86E+02 Ci/cc 6.86E+01 µCi/cc6.86E+00 µCi/cc 6.04E-04 µCi/cc Emergency Penetration Room 2RX-9835 (SPING 8) 5.88E+02 Ci/cc 5.88E+01 µCi/cc5.88E+00 µCi/cc 1.09E-02 µCi/cc Liquid BMS Liquid Discharge 2RE-2330 ---- ---- ---- 2.45E+04 cpm Regenerative Waste Discharge 2RE-4423 ---- ---- ---- 2.45E+05 cpm Mode Applicability: All Definition(s): VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

to 0CAN031801 Page 49 of 255 Attachment 1 - Emergency Action Level Technical Bases The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was 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 the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Reference(s): 1. OP-1604.051 Eberline Radiation Monitor System 2. EP-CALC-ANO-1701 Radiological Effluent EAL Values 3. NEI 99-01 AG1 to 0CAN031801 Page 50 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL: AG1.2 General Emergency Dose assessment using actual meteorology indicates doses > 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4) Note 4 The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Mode Applicability: All

Definition(s): SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant). Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the 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 was 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 the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. to 0CAN031801 Page 51 of 255 Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. OP-1904.002 Offsite Dose Projections 2. NEI 99-01 AG1 to 0CAN031801 Page 52 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL: AG1.3 General Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: Closed window dose rates > 1,000 mR/hr expected to continue for 60 min. Analyses of field survey samples indicate thyroid CDE > 5,000 mrem for 60 min. of inhalation (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All Definition(s):

SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the 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 was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. to 0CAN031801 Page 53 of 255 Attachment 1 - Emergency Action Level Technical Bases Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Reference(s):

1. OP-1905.002 Offsite Emergency Monitoring 2. NEI 99-01 AG1 to 0CAN031801 Page 54 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: UNPLANNED loss of water level above irradiated fuel EAL: AU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by low water level alarm, visual observation, or BWST[RWT] level drop due to makeup demands AND UNPLANNED rise in corresponding area radiation levels as indicated by any of the following radiation monitors: Unit 1 o RE-8009 Spent Fuel Area o RE-8017 Fuel Handling Area Unit 2 o 2RE-8914 Spent Fuel Area o 2RE-8915 Spent Fuel Area o 2RE-8916 Spent Fuel Area o 2RE-8912 Containment Incore Instrumentation Mode Applicability: All

Definition(s): UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. REFUELING PATHWAY - All the cavities, tubes, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel. to 0CAN031801 Page 55 of 255 Attachment 1 - Emergency Action Level Technical Bases Basis:

This IC addresses a decrease drop in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant.

A water level decrease drop will be primarily determined by indications from available level instrumentation. 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 in the water level may also cause an increase a rise in the radiation levels of adjacent areas that can be detected by monitors in those locations. The effects of planned evolutions should be considered. For example, a refueling bridge area radiation monitor reading may increase rise due to planned evolutions such as lifting of the reactor vessel head or movement of a fuel assembly. Note that this EAL is applicable only in cases where 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 in accordance Recognition Category C during the Cold Shutdown and Refueling modes.

Escalation of the emergency classification level would be via IC AA2.

Reference(s): 1. OP-1203.050 Unit 1 Spent Fuel Pool Emergencies 2. OP-2203.002 Spent Fuel Pool Emergencies

3. 1SAR 11.2.5 Area Radiation Monitoring Systems Table 11-15 Area Radiation Monitors

4. 2SAR 12.1.4 Area Radiation Monitoring System 5. NEI 99-01 AU2 to 0CAN031801 Page 56 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL: AA2.1 Alert IMMINENT uncovery of irradiated fuel in the REFUELING PATHWAY. Mode Applicability: All Definition(s): CONFINEMENT BOUNDARY - The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the ANO ISFSI, the Confinement Boundary is comprised of either the Multi-assembly Sealed Basket (MSB) (SNC System) or Multi-Purpose Canister (MPC) (Holtec System).

IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

REFUELING PATHWAY - All the cavities, tubes, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel. Basis: This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel poolREFUELING PATHWAY (see Developer Notes). These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant. This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E-HU1. Escalation of the emergency would be based on either Recognition Category A or C ICs. EAL #1 to 0CAN031801 Page 57 of 255 Attachment 1 - Emergency Action Level Technical Bases This EAL escalates from AU2 AU2.1 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. 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 and radiation 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 available indications, reports and observations. While an area radiation monitor could detect an increasea rise in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should 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 in accordance with Recognition Category C during the Cold Shutdown and Refueling modes. EAL #2 This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident). EAL #3 Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool. Escalation of the emergency classification level would be via ICs AS1 AS1or AS2 (see AS2 Developer Notes). Reference(s): 1. NEI 99-01 AA2

to 0CAN031801 Page 58 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL: AA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity AND High alarm on any Table 1[2]A-2 radiation monitor. Table 1A-2 Unit 1 Fuel Damage Radiation Monitors RE-8009 Spent Fuel Area RE-8017 Fuel Handling RE-8060 Containment High Range Radiation Monitors RE-8061 Containment High Range Radiation Monitors RX-9820 (SPING 1) Containment Purge RX-9825 (SPING 2) Radwaste Area RX-9830 (SPING 3) Fuel Handling Area Table 2A-2 Unit 2 Fuel Damage Radiation Monitors 2RE-8905 Containment Equipment Hatch Area 2RE-8909 Containment Personnel Access Area 2RE-8912 Containment Incore Inst. 2RE-8914 Spent Fuel Area 2RE-8915 Spent Fuel Area 2RE-8916 Spent Fuel Area 2RE-8925-1 Containment High Range Radiation Monitors 2RE-8925-2 Containment High Range Radiation Monitors 2RX-9820 (SPING 5) Containment Purge 2RX-9825 (SPING 6) Radwaste Area 2RX-9830 (SPING 7) Fuel Handling Area to 0CAN031801 Page 59 of 255 Attachment 1 - Emergency Action Level Technical Bases Mode Applicability:

All Definition(s): CONFINEMENT BOUNDARY - The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the ANO ISFSI, the Confinement Boundary is comprised of either the Multi-assembly Sealed Basket (MSB) (SNC System) or Multi-Purpose Canister (MPC) (Holtec System).

VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis: This IC EAL addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool (see Developer Notes). These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.

This IC EAL applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E-HU1. EAL #This EAL escalates from AU2 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. 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 and radiation 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 available indications, reports and observations. While an area radiation monitor could detect an increase in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should 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 in accordance Recognition Category C during the Cold Shutdown and Refueling modes.

This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be to 0CAN031801 Page 60 of 255 Attachment 1 - Emergency Action Level Technical Bases considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident). EAL #3Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool. Escalation of the emergency classification level would be via ICs AS1 or AS2 (see AS2 Developer Notes). Reference(s): 1. OP-1203.050 Unit 1 Spent Fuel Pool Emergencies 2. OP-1305.001 Radiation Monitoring System Check and Test

3. OP-2203.002 Spent Fuel Pool Emergencies

4. OP-1604.051 Eberline Radiation Monitoring System

5. OP-2304.133 Containment High Range Radiation Monitor Calibration 6. Offsite Dose Calculation Manual 7. NEI 99-01 AA2 to 0CAN031801 Page 61 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL: AA2.3 Alert Lowering of spent fuel pool level to 387.0 ft. [389.5 ft.] (Alarm 2) on LIT-2020-3(4) [2LIT-2020-1(2)] Mode Applicability: All Definition(s):

None Basis: This IC EAL addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool (see Developer Notes). These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E-HU1. Escalation of the emergency would be based on either Recognition Category A or C ICs. EAL #This EAL escalates from AU2 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. 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 and radiation 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 available indications, reports and observations. While an area radiation monitor could detect an increase in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should be considered in combination with other available indications of inventory loss. to 0CAN031801 Page 62 of 255 Attachment 1 - Emergency Action Level Technical Bases A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes. This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident). EAL #3Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool.

Escalation of the emergency classification level would be via ICs AS1 or AS2 (see AS2 Developer Notes). Post-Fukushima order EA 051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3) (ref. 1, 2). Reference(s): 1. MOHR-ANO-1, ANO-1 SFPI (Level) Configuration, Sheet 1, Revision 0 2. MOHR-ANO-2, ANO-2 SFPI (Level) Configuration, Sheet 1, Revision 0

3. NEI 99-01 AA2 to 0CAN031801 Page 63 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level at the top of the fuel racks EAL: AS2.1 Site Area Emergency Lowering of spent fuel pool level to 377.0 ft. [379.5 ft.] (Alarm 3) on LIT-2020-3(4) [2LIT-2020-1(2)] Mode Applicability: All Definition(s):

IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

Basis:

This IC EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity.

Escalation of the emergency classification level would be via IC AG1 or AG2A. Post-Fukushima order EA 051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3) (ref. 1, 2). Reference(s):

1. MOHR-ANO-1, ANO-1 SFPI (Level) Configuration, Sheet 1, Revision 0 2. MOHR-ANO-2, ANO-2 SFPI (Level) Configuration, Sheet 1, Revision 0

3. NEI 99-01 AS2 to 0CAN031801 Page 64 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer EAL: AG2.1 General Emergency Spent fuel pool level cannot be restored to at least 377.0 ft. [379.5 ft.] (Alarm 3) on LIT-2020-3(4) [2LIT-2020-1(2)] for 60 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: All

Definition(s): None Basis: This IC EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment.

It is recognized that this IC would likely not be met until well after another General Emergency IC was met; however, it is included to provide classification diversity.

Post-Fukushima order EA 051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3) (ref. 1, 2). Reference(s): 1. MOHR-ANO-1, ANO-1 SFPI (Level) Configuration, Sheet 1, Revision 0 2. MOHR-ANO-2, ANO-2 SFPI (Level) Configuration, Sheet 1, Revision 0 3. NEI 99-01 AG2 to 0CAN031801 Page 65 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 3 - Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL: AA3.1 Alert Dose rate > 15 mR/hr in EITHER of the following areas: Control Room Central Alarm Station (by survey) Mode Applicability:

All Definition(s): IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures 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). Basis: Areas that meet this threshold include the Control Room (CR) and the Central Alarm Station (CAS). The Control Room envelope (Unit 1 and Unit 2) is monitored for excessive radiation by five detectors. These radiation detectors are RE-8001, 2RE-8001A, 2RE-8001B, 2RE-8750-1A, and 2RE-8750-1B (ref. 1). The CAS is included in this EAL because of its importance to permitting access to areas required to assure safe plant operations. There are no permanently installed area radiation monitors in CAS that may be used to assess this EAL threshold. Therefore, this threshold is evaluated using local radiation survey for this area. This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or IMPEDE personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the increased rise in radiation levels and determine if another IC may be applicable. For EAL #2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the increased radiation levels. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the to 0CAN031801 Page 66 of 255 Attachment 1 - Emergency Action Level Technical Bases affected room/area (e.g., installing temporary shielding, requiring use of non-routine protective equipment, 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 affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation increase occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4. The increased radiation levels are a result of a planned activity that includes compensatory 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 record keeping nature (e.g., normal rounds or routine inspections). The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action. Escalation of the emergency classification level would be via Recognition Category A, C or F ICs. Reference(s): 1. STM 1-62 Radiation Monitoring 2. NEI 99-01 AA3 to 0CAN031801 Page 67 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 3 - Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL: AA3.2 Alert An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table 1[2]A-3 room or area (Note 5) Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. Table 1A-3 Unit 1 Safe Operation & Shutdown Rooms/Areas Room/Area Mode A-4 Switchgear Room 3, 4 Upper North Electrical Penetration Room 3, 4 Lower South Electrical Equipment Room 3, 4 Table 2A-3 Unit 2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Aux Building 317' Emergency Core Cooling Rooms 3, 4 Aux Building 317' Tendon Gallery Access 3, 4 Aux Building 335' Charging Pumps / MCC 2B-52 3, 4 Aux Building 354' MCC 2B-62 Area 3, 4 Emergency Diesel Generator Corridor 3, 4 Lower South Piping Penetration Room 3, 4 Aux Building 386' Containment Hatch 3, 4 to 0CAN031801 Page 68 of 255 Attachment 1 - Emergency Action Level Technical Bases Mode Applicability:

3 - Hot Standby, 4 - Hot Shutdown Definition(s): IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures 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 intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or IMPEDE personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the increased rise in radiation levels and determine if another IC may be applicable. For EAL #2 AA3.2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the increased higher radiation levels. Access should be considered as IMPEDED if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding, requiring use of non-routine protective equipment, 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 affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation increase rise occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 43. The increased higher radiation levels are a result of a planned activity that includes compensatory 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 record keeping nature (e.g., normal rounds or routine inspections). The access control measures are of a conservative or precautionary nature, and would not actually prevent or IMPEDE a required action. to 0CAN031801 Page 69 of 255 Attachment 1 - Emergency Action Level Technical Bases Escalation of the emergency classification level would be via Recognition Category A, C or F ICs.

If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact 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 those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1). EAL AA3.2 mode applicability has been limited to the mode limitations of Table 1[2]A-3 (Modes 3 and 4 only). Reference(s): 1. Attachment 2 Safe Operation & Shutdown Areas Tables 1[2]A-3 & 1[2]H-2 Bases 2. NEI 99-01 AA3 to 0CAN031801 Page 70 of 255 Attachment 1 - Emergency Action Level Technical Bases Category C - Cold Shutdown / Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature 200°F); EALs in this category are applicable only in one or more cold operating modes. Category C EALs are directly associated with cold shutdown or refueling system safety functions. Given the variability of plant configurations (e.g., systems out-of-service for maintenance, containment open, reduced AC power redundancy, time since shutdown) during these periods, the consequences of any given initiating event can vary greatly. For example, a loss of decay heat removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable. The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, CONTAINMENT CLOSURE, and fuel clad integrity for the applicable operating modes (5 - Cold Shutdown, 6 - Refueling, DEF - Defueled). The events of this category pertain to the following subcategories: 1. RCS Level RCS water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity. 2. Loss of Vital AC Power Loss of vital plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite power sources for 4.16 KV vital buses. 3. RCS Temperature Uncontrolled or inadvertent temperature or pressure rises are indicative of a potential loss of safety functions. 4. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the 125V DC vital buses. 5. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification. 6. Hazardous Event Affecting Safety Systems Certain hazardous natural and technological events may result in VISIBLE DAMAGE to or degraded performance of safety systems warranting classification. to 0CAN031801 Page 71 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: UNPLANNED loss of RCS inventory EAL: CU1.1 Unusual Event UNPLANNED loss of reactor coolant results in RCS water level less than a required lower limit for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 5 - Cold Shutdown, 6 - Refueling Definition(s): UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis: This IC EAL 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 (reactor vessel/RCS [PWR] or RPV [BWR]) level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.

Refueling evolutions that decreaselower RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level decreasinglowering below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered. This EAL #1 recognizes that the minimum required (reactor vessel/RCS [PWR] or RPV [BWR]) level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document.

to 0CAN031801 Page 72 of 255 Attachment 1 - Emergency Action Level Technical Bases The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level.

EAL #2 addresses a condition where all means to determine (reactor vessel/RCS [PWR] or RPV [BWR]) level have been lost. In this condition, operators may determine 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 water flow to ensure they are indicative of leakage from the (reactor vessel/RCS [PWR] or RPV [BWR]). Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3. Reference(s): 1. OP-1015.002 Decay Heat Removal and LTOP System 2. OP-1015.008 Unit 2 SDC Control

3. NEI 99-01 CU1 to 0CAN031801 Page 73 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: UNPLANNED loss of RCS inventory EAL: CU1.2 Unusual Event RCS level cannot be monitored AND EITHER UNPLANNED rise in any Table 1[2]C-1 sump/tank level due to loss of RCS inventory Visual observation of UNISOLABLE RCS leakage Table 1C-1 Unit 1 Sumps / Tanks Reactor Building Sump Reactor Drain Tank Aux. Building Equipment Drain Tank Aux. Building Sump Quench Tank Table 2C-1 Unit 2 Sumps / Tanks CNTMT Sump Reactor Drain Tank LRW Waste Tank (2T-20) Holdup Tank Aux. Building Sump Quench Tank Mode Applicability:

5 - Cold Shutdown, 6 - Refueling to 0CAN031801 Page 74 of 255 Attachment 1 - Emergency Action Level Technical Bases Definition(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 intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. 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 (reactor vessel/RCS [PWR] or RPV [BWR]) level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.

Refueling evolutions that decrease lower RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level decreasing lowering below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered.

EAL #1 recognizes that the minimum required (reactor vessel/RCS [PWR] or RPV [BWR]) level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document. The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level. This EAL #2 addresses a condition where all means to determine (reactor vessel/RCS [PWR] or RPV [BWR]) level have been lost. In this condition, operators may determine that an inventory 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 water flow to ensure they are indicative of leakage from the (reactor vessel/RCS [PWR] or RPV [BWR]). Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.

Reference(s): 1. OP-1203.039 Excess RCS Leakage 2. OP-2203.016 Excess RCS Leakage

3. NEI 99-01 CU1 to 0CAN031801 Page 75 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Significant Loss of RCS inventory EAL: CA1.1 Alert Loss of RCS inventory as indicated by EITHER: RVLMS Levels 1 through 8 [1 through 5] indicate DRY Reactor vessel level 368.5 ft. (LT-1195/LT-1196) [0 in. (L4791/L4792)] (bottom of hot leg) Mode Applicability: 5 - Cold Shutdown, 6 - Refueling Definition(s): None Basis: This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.

For this EAL #1, a lowering of RPV water level below (site-specific level) ft the specified level indicates that operator actions have not been successful in restoring and maintaining RCS (reactor vessel/RCS [PWR] or RPV [BWR]) water level. The heat-up rate of the coolant will increase rise as the available water inventory is reduced. A continuing decrease drop in water level will lead to core uncovery. Although related, this EAL #1 is concerned with the loss of RPV inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Residual Decay Heat Removal suction point). An increase rise in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3. For EAL #2, the inability to monitor (reactor vessel/RCS [PWR] or RPV [BWR]) level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine 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 water flow to ensure they are indicative of leakage from the (reactor vessel/RCS [PWR] or RPV [BWR]). to 0CAN031801 Page 76 of 255 Attachment 1 - Emergency Action Level Technical Bases The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1 If RCS the (reactor vessel/RCS [PWR] or RPV [BWR]) inventory water level continues to lower, then escalation to Site Area Emergency would be via IC CS1.

The bottom of the RCS hot leg penetration into the reactor vessel is approximately RLVMS Level 8 (Unit 1) or RVLMS Level 5 (Unit 2). However, RVLMS may not be available in the cold shutdown modes. Redundant means of level indication is provided in these modes and included in this EAL. The bottom of the RCS hot leg penetration into the reactor vessel is 368 ft., 0 in. (Unit 1) or 369 ft., 1.5 in. (Unit 2). Below this level, reactor vessel level indication may be lost and loss of suction to decay heat removal systems will occur (ref. 1, 2, 3). Where redundant means of level indication cannot read below this level, the lowest indication that can be read on scale is used for this EAL. The inability to restore and maintain level after reaching this setpoint would be indicative of a failure of the RCS barrier. Reference(s):

1. OP-1105.008 Inadequate Core Cooling Monitor and Display 2. OP-2105.003 Reactor Vessel Level Monitoring System Operations

3. Calculation No. 90-E-0116-01 ANO-2 EOP Setpoint Basis Document, Setpoints R.3 and R.9

4. NEI 99-01 CA1 to 0CAN031801 Page 77 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Significant Loss of RCS inventory EAL: CA1.2 Alert RCS level cannot be monitored for 15 min. (Note 1) AND EITHER UNPLANNED rise in any Table 1[2]C-1 Sump / Tank level due to a loss of RCS inventory Visual observation of UNISOLABLE RCS leakage Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table 1C-1 Unit 1 Sumps / Tanks Reactor Building Sump Reactor Drain Tank Aux. Building Equipment Drain Tank Aux. Building Sump Quench Tank Table 2C-1 Unit 2 Sumps / Tanks CNTMT Sump Reactor Drain Tank LRW Waste Tank (2T-20) Holdup Tank Aux. Building Sump Quench Tank to 0CAN031801 Page 78 of 255 Attachment 1 - Emergency Action Level Technical Bases Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Definition(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 intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis: This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.

For EAL #1, a lowering of water level below (site-specific level) indicates that operator actions have not been successful in restoring and maintaining (reactor vessel/RCS [PWR] or RPV [BWR]) water level. The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery. Although related, EAL #1 is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Residual Heat Removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3. For this EAL #2, the inability to monitor RCS (reactor vessel/RCS [PWR] or RPV [BWR]) level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine 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 water flow to ensure they are indicative of leakage from the (reactor vessel/RCS [PWR] or RPV [BWR]). The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1. If the (reactor vessel/RCS [PWR] or RPV [BWR]) inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1. Reference(s): 1. OP-1203.039 Excess RCS Leakage 2. OP-2203.016 Excess RCS Leakage 3. NEI 99-01 CA1 to 0CAN031801 Page 79 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability EAL: CS1.1 Site Area Emergency CONTAINMENT CLOSURE not established AND RVLMS Levels 1 through 9 [1 through 6] indicate DRY Mode Applicability: 5 - Cold Shutdown, 6 - Refueling Definition(s): CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: This IC addresses a significant and prolonged loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection 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 will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.

Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions.

The difference in the specified RCS/reactor vessel levels of EALs 1.b and 2.b reflect the fact to 0CAN031801 Page 80 of 255 Attachment 1 - Emergency Action Level Technical Bases that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment. In EAL 3.a, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor (reactor vessel/RCS [PWR] or RPV [BWR]) level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine 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 water flow to ensure they are indicative of leakage from the (reactor vessel/RCS [PWR] or RPV [BWR]). Thisese EALs addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; 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 United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

Escalation of the emergency classification level would be via IC CG1 or AG1. Reference(s): 1. OP-1105.008 Inadequate Core Cooling Monitor and Display 2. OP-2105.003 Reactor Vessel Level Monitoring System Operations

3. NEI 99-01 CS1 to 0CAN031801 Page 81 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability EAL: CS1.2 Site Area Emergency [RVLMS Levels 1 through 7 indicate DRY OR] RCS level cannot be monitored for 30 min. (Note 1) AND Core uncovery is indicated by any of the following: UNPLANNED rise in any Table 1[2]C-1 sump/tank level of sufficient magnitude to indicate core uncovery Containment high range radiation monitor RE-8060/8061 [2RE-8925-1/8925-2] reading > 10 R/hr Erratic Source Range Monitor indication Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table 1C-1 Unit 1 Sumps / Tanks Reactor Building Sump Reactor Drain Tank Aux. Building Equipment Drain Tank Aux. Building Sump Quench Tank to 0CAN031801 Page 82 of 255 Attachment 1 - Emergency Action Level Technical Bases Table 2C-1 Unit 2 Sumps / Tanks CNTMT Sump Reactor Drain Tank LRW Waste Tank (2T-20) Holdup Tank Aux. Building Sump Quench Tank Mode Applicability: 5 - Cold Shutdown, 6 - Refueling

Definition(s): CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: When in service, the Unit 2 RVLMS can measure RCS level below the top of active fuel. Level 7 DRY on this system is an indication of core uncovery. This IC addresses a significant and prolonged loss of (reactor vessel/RCS RCS [PWR] or RPV [BWR]) inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. to 0CAN031801 Page 83 of 255 Attachment 1 - Emergency Action Level Technical Bases Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.

Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions.

The difference in the specified RCS/reactor vessel levels of EALs 1.b and 2.b CS1.1 and CS1.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.

In EAL 3.a, Tthe 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

The inability to monitor (reactor vessel/RCS [PWR] or RPV [BWR]) level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine 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 water flow to ensure they are indicative of leakage from the (reactor vessel/RCS [PWR] or RPV [BWR]). These This EALs addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; 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 United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Containment High Range Radiation Monitors RE-8060/8061 [2RE-8925-1/8925-2] are the site-specific radiation monitors that would be indicative of possible core uncovery in the Refueling mode. The dose rate due to core shine when the top of the core becomes uncovered should result in dose rates > 10 R/hr. Escalation of the emergency classification level would be via IC CG1 or AG1. Reference(s): 1. OP-1203.039 Excess RCS Leakage 2. OP-2203.016 Excess RCS Leakage 3. OP-2105.003 Reactor Vessel Level Monitoring System Operations 4. 1SAR Table 7-11

5. 2SAR 12.1.4.2

6. NEI 99-01 CS1 to 0CAN031801 Page 84 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting fuel clad integrity with containment challenged EAL: CG1.1 General Emergency - UNIT 2 ONLY RVLMS Levels 1 through 7 indicate DRY AND Any Containment Challenge indication, Table 2C-2 Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required. Table 1[2]C-2 Containment Challenge Indications CONTAINMENT CLOSURE not established (Note 6) Containment hydrogen concentration > 3% UNPLANNED rise in containment pressure Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Definition(s): CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent. IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. to 0CAN031801 Page 85 of 255 Attachment 1 - Emergency Action Level Technical Bases Basis: When in service, the Unit 2 RVLMS can measure RCS level below the top of active fuel. Level 7 DRY on this system is an indication of core uncovery. This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable. With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity.

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged. In EAL 2.b, tThe 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor (reactor vessel/RCS [PWR] or RPV [BWR]) level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine 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 water flow to ensure they are indicative of leakage from the (reactor vessel/RCS [PWR] or RPV [BWR]). to 0CAN031801 Page 86 of 255 Attachment 1 - Emergency Action Level Technical Bases This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; 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 United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Reference(s): 1. OP-1203.039 Excess RCS Leakage 2. OP-2203.016 Excess RCS Leakage

3. OP-2105.003 Reactor Vessel Level Monitoring System Operations

4. 1SAR Table 7-11 5. 2SAR 12.1.4.2 6. Unit 1 SAMG Figure III-1B

7. Unit 2 SAMG Phase 1 Instructions, Containment Flowchart

8. NEI 99-01 CG1 to 0CAN031801 Page 87 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting fuel clad integrity with containment challenged EAL: CG1.2 General Emergency RCS level cannot be monitored for 30 min. (Note 1) AND Core uncovery is indicated by any of the following: UNPLANNED rise in any Table 1[2]C-1 sump/tank level of sufficient magnitude to indicate core uncovery Containment High Range Radiation Monitor RE-8060/8061 [2RE-8925-1/8925-2] reading > 10 R/hr Erratic Source Range Monitor indication AND Any Containment Challenge indication, Table 1[2]C-2 Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required. Table 1C-1 Unit 1 Sumps / Tanks Reactor Building Sump Reactor Drain Tank Aux. Building Equipment Drain Tank Aux. Building Sump Quench Tank to 0CAN031801 Page 88 of 255 Attachment 1 - Emergency Action Level Technical Bases Table 2C-1 Unit 2 Sumps / Tanks CNTMT Sump Reactor Drain Tank LRW Waste Tank (2T-20) Holdup Tank Aux. Building Sump Quench Tank Table 1[2]C-2 Containment Challenge Indications CONTAINMENT CLOSURE not established (Note 6) Containment hydrogen concentration > 3% UNPLANNED rise in containment pressure Mode Applicability: 5 - Cold Shutdown, 6 - Refueling Definition(s): CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

to 0CAN031801 Page 89 of 255 Attachment 1 - Emergency Action Level Technical Bases Basis:

When in service, the Unit 2 RVLMS can measure RCS level below the top of active fuel. Level 7 DRY on this system is an indication of core uncovery. This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable. With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity.

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged.

In EAL 2.b, tThe 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor (reactor vessel/RCS [PWR] or RPV [BWR]) level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine 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 water flow to ensure they are indicative of leakage from the (reactor vessel/RCS [PWR] or RPV [BWR]). to 0CAN031801 Page 90 of 255 Attachment 1 - Emergency Action Level Technical Bases Containment High Range Radiation Monitor RE-8060/8061 [2RE-8925-1/8925-2] are the site-specific radiation monitors that would be indicative of possible core uncovery in the Refueling mode. The dose rate due to core shine when the top of the core becomes uncovered should result in dose rates > 10 R/hr. This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; 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 United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Reference(s):

1. OP-1203.039 Excess RCS Leakage 2. OP-2203.016 Excess RCS Leakage

3. OP-2105.003 Reactor Vessel Level Monitoring System Operations

4. 1SAR Table 7-11

5. 2SAR 12.1.4.2

6. Unit 1 SAMG Figure III-1B 7. Unit 2 SAMG Phase 1 Instructions, Containment Flowchart 8. NEI 99-01 CG1

to 0CAN031801 Page 91 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Vital AC Power Initiating Condition: Loss of all but one AC power source to vital buses for 15 minutes or longer EAL: CU2.1 Unusual Event AC power capability, Table 1[2]C-3, to vital 4.16 KV buses A3 [2A3] and A4 [2A4] reduced to a single power source for 15 min. (Note 1) AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table 1C-3 Unit 1 AC Power Sources Offsite Startup Transformer No. 1 Startup Transformer No. 2 Unit Auxiliary Transformer (from 22 KV switchyard) Onsite DG1 DG2 AAC Gen to 0CAN031801 Page 92 of 255 Attachment 1 - Emergency Action Level Technical Bases Table 2C-3 Unit 2 AC Power Sources Offsite Startup Transformer No. 3 Startup Transformer No. 2 Unit Auxiliary Transformer (backfed from main transformer) Onsite 2DG1 2DG2 AAC Gen Mode Applicability:

5 - Cold Shutdown, 6 - Refueling, DEF - Defueled Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the 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 result in potential offsite exposures. Basis: This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.

When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the increased greater time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant. to 0CAN031801 Page 93 of 255 Attachment 1 - Emergency Action Level Technical Bases An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to a vital bus. Some examples of this condition are presented below. A loss of all offsite power with a concurrent failure of all but one emergency vital power source (e.g., an onsite diesel generator). A loss of all offsite power and loss of all emergency vital power sources (e.g., onsite diesel generators) with a single train of emergency vital buses being back-fed from the unit main generator. A loss of emergency vital power sources (e.g., onsite diesel generators) with a single train of emergency vital buses being back-fed from an offsite power source. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2.

This EAL is the cold condition equivalent of the hot condition EAL SA1.1. Reference(s): 1. 1SAR Figure 8-1 Station Single Line Diagram 2. OP-1202.007 Degraded Power

3. OP-1202.008 Blackout 4. OP-2104.037 Alternate AC Diesel Generator Operations 5. 2SAR Figure 8.3-1 Station Single Line Diagram

6. OP-2202.007 Loss of Off-Site Power

7. OP-2202.008 Station Blackout

8. OP-2107.006 Backfeed of Unit Auxiliary Transformer 9. NEI 99-01 CU2 to 0CAN031801 Page 94 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Vital AC Power Initiating Condition: Loss of all offsite and all onsite AC power to vital buses for 15 minutes or longer EAL: CA2.1 Alert Loss of all offsite and all onsite AC power to vital 4.16 KV buses A3 [2A3] and A4 [2A4] for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 5 - Cold Shutdown, 6 - Refueling, DEF - Defueled

Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the 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 result in potential offsite exposures. Basis:

Although the AAC may be considered available, it will not prevent declaration of this EAL unless it is powering a vital bus within the 15-minute time period of the EAL. This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. Mitigative strategies using non-safety related power sources (FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in to 0CAN031801 Page 95 of 255 Attachment 1 - Emergency Action Level Technical Bases accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased greater time available to restore an emergency vital bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.

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 AS1.

This EAL is the cold condition equivalent of the hot condition EAL SS1.1. Reference(s): 1. 1SAR Figure 8-1 Station Single Line Diagram 2. OP-1202.007 Degraded Power

3. OP-1202.008 Blackout

4. OP-2104.037 Alternate AC Diesel Generator Operations 5. 2SAR Figure 8.3-1 Station Single Line Diagram 6. OP-2202.007 Loss of Off-Site Power

7. OP-2202.008 Station Blackout

8. OP-2107.006 Backfeed of Unit Auxiliary Transformer

9. NEI 99-01 CU2 to 0CAN031801 Page 96 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: UNPLANNED rise in RCS temperature EAL: CU3.1 Unusual Event UNPLANNED rise in RCS temperature to > 200°F due to loss of decay heat removal capability Mode Applicability: 5 - Cold Shutdown, 6 - Refueling° Definition(s): CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis: This IC addresses an UNPLANNED increase rise in RCS temperature above the Technical Specification cold shutdown temperature limit or the inability to determine RCS temperature and level,and represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to IC EAL CA3.1. A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. EAL #1This EALThis EAL involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. to 0CAN031801 Page 97 of 255 Attachment 1 - Emergency Action Level Technical Bases During an outage, the level in the reactor vessel will normally be maintained at or above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled. A loss of forced decay heat removal at loweredreduced inventory may result in a rapid increase rise in reactor coolant temperature depending on the time after shutdown.

EAL #2 reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.

Reference(s): 1. Unit 1 and Unit 2 Technical Specifications Table 1.1-1 2. NEI 99-01 CU3 to 0CAN031801 Page 98 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: UNPLANNED rise in RCS temperature EAL: CU3.2 Unusual Event Loss of all RCS temperature and RCS level indication for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 5 - Cold Shutdown, 6 - Refueling Definition(s):

CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: This IC EALEAL addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to determine RCS temperature and level, andand represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to IC EAL CA3.1. A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. EAL #1 involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be to 0CAN031801 Page 99 of 255 Attachment 1 - Emergency Action Level Technical Bases maintained below the cold shutdown temperature limit specified in Technical Specifications. During this condition, there is no immediate threat of fuel damage because the core 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 above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled. A loss of forced decay heat removal at reduced inventory may result in a rapid increase in reactor coolant temperature depending on the time after shutdown. EAL #2This EAL reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.

Reference(s):

1. NEI 99-01 CU3 to 0CAN031801 Page 100 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: Inability to maintain plant in cold shutdown EAL: CA3.1 Alert UNPLANNED rise in RCS temperature to > 200°F for > Table 1[2]C-4 duration (Note 1) OR UNPLANNED RCS pressure rise > 10 psig due to a loss of RCS cooling (this EAL does not apply during water-solid plant conditions) Note 1: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table 1[2]C-4 RCS Heat-up Duration Thresholds RCS Status CONTAINMENT CLOSURE Status Heat-up Duration Intact (but not lowered inventory) N/A 60 min.* Not intact OR lowered inventory established 20 min.* not established 0 min.

• If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable. Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Definition(s): CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

to 0CAN031801 Page 101 of 255 Attachment 1 - Emergency Action Level Technical Bases As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on the RCS pressure rise criteria when the RCS is intact in Mode 5 or based on time to boil data when in Mode 6 or the RCS is not intact in Mode 5. This IC EAL addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant. A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. The RCS Heat-up Duration Thresholds table addresses an increasea rise in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS inventory is reduced (e.g., lowered inventory operationmid-loop operation in PWRs). The 20-minute criterion was included to allow time for operator action to address the temperature increaserise. The RCS Heat-up Duration Thresholds table also addresses an increasea rise in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release.

The 60-minute time frame should allow sufficient time to address the temperature increase rise without a substantial degradation in plant safety. Finally, in the case where there is an increase rise in RCS temperature, the RCS is not intact or is at loweredreduced inventory [PWR], and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel. EAL #2The RCS pressure rise threshold provides a pressure-based indication of RCS heat-up in the absence of RCS temperature monitoring capability. Escalation of the emergency classification level would be via IC CS1 or AS1. Reference(s): 1. Unit 1 and Unit 2 Technical Specifications Table 1.1-1 2. NEI 99-01 CA3 to 0CAN031801 Page 102 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 4 - Loss of Vital DC Power Initiating Condition: Loss of Vital DC power for 15 minutes or longer EAL: CU4.1 Unusual Event Indicated voltage is < 105 VDC on vital 125 VDC buses for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 5 - Cold Shutdown, 6 - Refueling Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the 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 result in potential offsite exposures. Basis: Unit 1 batteries D06 and D07 and Unit 2 batteries 2D11 and 2D12 contain 58 cells each with a minimum cell voltage of 1.81 V or 105 VDC. This IC addresses a loss of vital DC power which compromises the ability to monitor and control 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 system temperatures and pressures are lower; these conditions increase raise the time available to restore a vital DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant.

to 0CAN031801 Page 103 of 255 Attachment 1 - Emergency Action Level Technical Bases As used in this EAL, "required" means the vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency 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 IC CA1 or CA3, or an IC in Recognition Category A.

This EAL is the cold condition equivalent of the hot condition EAL SS2.1. Reference(s): 1. 1SAR 8.3.2.1.1 Batteries 2. 2SAR 8.3.2.1.1 Batteries 3. NEI 99-01 CU4 to 0CAN031801 Page 104 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 5 - Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL: CU5.1 Unusual Event Loss of all Table 1[2]C-5 onsite communication methods OR Loss of all Table 1[2]C-5 State and local agency communication methods OR Loss of all Table 1[2]C-5 NRC communication methods Table 1[2]C-5 Communication Methods System Onsite ORO NRC Station radio system X ANO plant phone system X Gaitronics X Telephone Systems: Commercial Microwave Satellite VOIP X X INFORM Notification System X Emergency Notification System (ENS) X Mode Applicability: 5 - Cold Shutdown, 6 - Refueling, DEF - Defueled Definition(s): None to 0CAN031801 Page 105 of 255 Attachment 1 - Emergency Action Level Technical Bases Basis: This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to State and local agenciesOROs and the NRC. This IC should be assessed only when extraordinary means are being utilized to make communications 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 to offsite locations, etc.).

EAL #1The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations.

EAL #2The second EAL condition addresses a total loss of the communications methods used to notify all State and local agenciesOROs of an emergency declaration. The State and local agenciesOROs referred to here are the Arkansas Department of Health, Arkansas Department of Emergency Management, Pope, Yell, Johnson, and Logan County offsite agencies (see Developer Notes). EAL #3The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. This EAL is the cold condition equivalent of the hot condition EAL SU7.1. Reference(s): 1. OP-1903.062 Communications System Operating Procedure 2. NEI 99-01 CU5 to 0CAN031801 Page 106 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 6 - Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode EAL: CA6.1 Alert The occurrence of any Table 1[2]C-6 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER: Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 10, 11) Note 10: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted. Note 11: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted. Table 1[2]C-6 Hazardous Events Seismic event (earthquake) Internal or external FLOODING event High winds or tornado strike FIRE EXPLOSION Other events with similar hazard characteristics as determined by the Shift Manager to 0CAN031801 Page 107 of 255 Attachment 1 - Emergency Action Level Technical Bases Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Definition(s): EXPLOSION - A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines 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 drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred 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 installed equipment 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/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the 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 result in potential offsite exposures.

VISIBLE DAMAGE - Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train. Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM to 0CAN031801 Page 108 of 255 Attachment 1 - Emergency Action Level Technical Bases train is from the same SAFETY SYSTEM that has indications of degraded performance; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events. Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure 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 product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant. EAL 1.b.1 addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. EAL 1.b.2 addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. Escalation of the emergency classification level would be via IC CS1 or AS1.

This EAL is the cold condition equivalent of the hot condition EAL SA9.1. Reference(s): 1. EP FAQ 2016-002 2. NEI 99-01 CA6 to 0CAN031801 Page 109 of 255 Attachment 1 - Emergency Action Level Technical Bases Category E - Independent Spent Fuel Storage Installation (ISFSI) EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

An independent spent fuel storage installation (ISFSI) is a complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. A significant amount of the radioactive material contained within a canister must escape its packaging and enter the biosphere for there to be a significant environmental effect resulting from an accident involving the dry storage of spent nuclear fuel.

An Unusual Event is declared on the basis of the occurrence of an event of sufficient magnitude that a loaded cask CONFINEMENT BOUNDARY is damaged or violated.

The ANO ISFSI is located wholly within the plant PROTECTED AREA. Therefore any security event related to the ISFSI is classified under Category H1 security event related EALs.

to 0CAN031801 Page 110 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: E - ISFSI Subcategory: Confinement Boundary Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY EAL: EU1.1 Unusual Event Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask (VSC-24 VCC or HI-STORM overpack) > any Table 1[2]-E-1 value Table 1[2]E-1 ISFSI Dose Rates VSC-24 VCC HI-STORM 200 mrem/hr on the sides 400 mrem/hr on the top 700 mrem/hr at the air inlet 200 mrem/hr at the air outlet 60 mrem/hr (gamma + neutron) on the top or outlet vent 600 mrem/hr (gamma + neutron on the side of the side of the overpack (excluding inlet and outlet ducts)

Mode Applicability: All Definition(s): CONFINEMENT BOUNDARY - The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the ANO ISFSI, the Confinement Boundary is comprised of either the Multi-assembly Sealed Basket (MSB) (SNC System) or Multi-Purpose Canister (MPC) (Holtec System).

INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) - A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.

Basis: This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the to 0CAN031801 Page 111 of 255 Attachment 1 - Emergency Action Level Technical Bases creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.

The existence of "damage" is determined by radiological survey. The specified EAL threshold values correspond to 2 times the cask technical specification values (ref. 1, 2). The technical specification (licensing bases document) multiple of "2 times", which is also used in Recognition Category A IC AU1, is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.

Security-related events for ISFSIs are covered under ICs HU1 and HA1.

Reference(s):

1. Certificate of Compliance Appendix A Technical Specifications for the HI-STORM 100 Cask System Section 5.7.4 2. VSC-24 Storage Cask Final Safety Analysis Report Section 1.2.4 Maximum External Surface Dose Rate 3. NEI 99-01 E-HU1 to 0CAN031801 Page 112 of 255 Attachment 1 - Emergency Action Level Technical Bases Category F - Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature > 200°F); EALs in this category are applicable only in one or more hot operating modes.

EALs in this category represent threats to the defense in depth design concept that precludes the release of highly radioactive fission products to the environment. This concept relies on multiple physical barriers any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. The primary fission product barriers are:

A. Fuel Clad Barrier (FCB): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets. B. Reactor Coolant System Barrier (RCB): The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves. C. Containment Barrier (CNB): The Containment Barrier includes the Reactor Building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the Reactor Building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (ECL) from an Alert to a Site Area Emergency or a General Emergency.

The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-1. "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. "Loss" means the barrier no longer assures containment of radioactive materials. "Potential Loss" means integrity of the barrier is threatened and could be lost if conditions continue to degrade. The number of barriers that are lost or potentially lost and the following criteria determine the appropriate emergency classification level: Alert: Any loss or any potential loss of either Fuel Clad or RCS Barrier Site Area Emergency: Loss or potential loss of any two barriers General Emergency: Loss of any two barriers and loss or potential loss of third barrier to 0CAN031801 Page 113 of 255 Attachment 1 - Emergency Action Level Technical Bases The logic used for emergency classification based on fission product barrier monitoring should reflect the following considerations: The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Containment Barrier. Unusual Event ICs associated with RCS and Fuel Clad Barriers are addressed under System Malfunction ICs. For accident conditions involving a radiological release, evaluation of the fission product barrier thresholds will need to be performed in conjunction with dose assessments to ensure correct and timely escalation of the emergency classification. For example, an evaluation of the fission product barrier thresholds may result in a Site Area Emergency classification while a dose assessment may indicate that an EAL for General Emergency IC AG1 has been exceeded. The fission product barrier thresholds specified within a scheme reflect plant-specific ANO design and operating characteristics. As used in this category, the term RCS leakage encompasses not just those types defined in Technical Specifications but also includes the loss of RCS mass to any location - inside the containment, an interfacing system, or outside of the 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 far present conditions are from meeting a threshold that would require a General Emergency declaration. For example, if the Fuel Clad and RCS fission product barriers were both lost, then there should be frequent assessments of containment radioactive inventory and integrity. Alternatively, if both the Fuel Clad and RCS fission product barriers were potentially lost, the Emergency Director would have more assurance that there was no immediate need to escalate to a General Emergency.

to 0CAN031801 Page 114 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: F - Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of either Fuel Clad or RCS EAL: FA1.1 Alert Any loss or any potential loss of either Fuel Clad or RCS barrier (Table 1[2]F-1) Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s): None Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table 1[2]F-1 lists the fission product barrier thresholds, bases and references. At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavily than the Containment barrier. Unlike the Containment barrier, loss or potential loss of either the Fuel Clad or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability. Note that the loss or potential loss of Containment barrier in combination with loss or potential loss of either Fuel Clad or RCS barrier results in declaration of a Site Area Emergency under EAL FS1.1. Reference(s): 1. NEI 99-01 FA1 to 0CAN031801 Page 115 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: F - Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss or potential loss of any two barriers EAL: FS1.1 Site Area Emergency Loss or potential loss of any two barriers (Table 1[2]F-1) Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s): IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table 1[2]F-1 lists the fission product barrier thresholds, bases and references. At the Site Area Emergency classification level, each barrier is weighted equally. A Site Area Emergency is therefore appropriate for any combination of the following conditions: One barrier loss and a second barrier loss (i.e., loss - loss) One barrier loss and a second barrier potential loss (i.e., loss - potential loss) One barrier potential loss and a second barrier potential loss (i.e., potential loss - potential loss) At the Site Area Emergency classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a General Emergency is important. For example, the existence of Fuel Clad and RCS Barrier loss thresholds in addition to offsite dose assessments would require continual assessments of radioactive inventory and Containment integrity in anticipation of reaching a General Emergency classification. Alternatively, if both Fuel Clad and RCS potential loss thresholds existed, the Emergency Director would have greater assurance that escalation to a General Emergency is less IMMINENT. Reference(s): 1. NEI 99-01 FS1 to 0CAN031801 Page 116 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: F - Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss of any two barriers and loss or potential loss of third barrier EAL: FG1.1 General Emergency Loss of any two barriers AND Loss or potential loss of the third barrier (Table 1[2]F-1) Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s): None Basis:

Fuel Clad, RCS and Containment comprise the fission product barriers. Table 1[2]F-1 lists the fission product barrier thresholds, bases and references. At the General Emergency classification level each barrier is weighted equally. A General Emergency 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 Barrier Reference(s): 1. NEI 99-01 FG1 to 0CAN031801 Page 117 of 255 Attachment 1 - Emergency Action Level Technical Bases Table 1[2]F-1 Fission Product Barrier Threshold Matrix & Bases Table 1[2]F-1 lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment). The table is structured so that each of the three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential Loss thresholds. The first column of the table (to the left of the Fuel Clad Loss column) lists the categories (types) of fission product barrier thresholds. The fission product barrier categories are:

A. RCS or S/G Tube Leakage B. Inadequate Heat removal C. Containment Radiation / RCS Activity D. Containment Integrity or Bypass E. Emergency Director Judgment Each category occupies a row in Table 1[2]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 or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word "None" is entered in the cell. Thresholds are assigned sequential numbers within each barrier column beginning with number one (ex., FCB1, FCB2-FCB9). If a cell in Table 1[2]F-1 contains more than one numbered threshold, each of the numbered thresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessary to exceed all of the thresholds in a category before declaring a barrier Loss/Potential Loss. Subdivision of Table 1[2]F-1 by category facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss thresholds. This structure promotes a systematic approach to assessing the classification status of the fission product barriers. When equipped with knowledge of plant conditions related to the fission product barriers, the EAL-user first scans down the category column of Table 1[2]F-1, locates the likely category and then reads across the fission product barrier Loss and Potential Loss thresholds in that category 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 the new category. If the EAL-user determines that any threshold has been exceeded, by definition, the barrier is lost 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 three fission product barriers to determine if other barrier thresholds in the category are lost or potentially lost. For example, if containment radiation is sufficiently high, a Loss of the Fuel to 0CAN031801 Page 118 of 255 Attachment 1 - Emergency Action Level Technical Bases Clad and RCS Barriers and a Potential Loss of the Containment Barrier can occur. Barrier Losses and Potential Losses are then applied to the algorithms given in EALs FG1.1, FS1.1, 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 each barrier, the bases are given according category Loss followed by category Potential Loss beginning with Category A, then B,-, E. to 0CAN031801 Page 119 of 255 Attachment 1 - Emergency Action Level Technical Bases Table 1[2]F-1 Fission Product Barrier Threshold Matrix Fuel Clad Barrier (FCB) Reactor Coolant System Barrier (RCB) Containment Barrier (CNB) Category Loss Potential Loss Loss Potential Loss Loss Potential Loss A RCS or S/G Tube Leakage None FCB1 RVLMS Levels 1 through 9 [1 through 7] indicate DRY RCB1 An automatic or manual ESAS [ESFAS] actuation required by EITHER: UNISOLABLE RCS leakage S/G tube RUPTURE RCB2 UNISOLABLE RCS leakage or S/G tube leakage > 50[44] gpm excluding normal reductions in RCS inventory (e.g., letdown, RCP seal leakoff) RCB3 Unit 1: PTS limits apply (RT14) AND RCS pressure and temperature are left of the NDTT/LTOP limit lines on EOP Figure 3 (Note 12) Unit 2: Uncontrolled RCS cooldown (50°F step change which is below 500°F from NOT) AND RCS pressure and temperature are to the left of line B (200 degrees MTS),

Standard Attachment 1, P-T Limits (Note 12) CNB1 A S/G that is leaking > 50[44] gpm (excluding normal reductions in RCS inventory) or that is RUPTURED is also FAULTED outside of containment None B Inadequate Heat Removal FCB2 CETs > 1200°F FCB3 CETs > 700°F FCB4 RCS heat removal cannot be established using steam generators AND HPI [Once Through] cooling initiated None RCB4 RCS heat removal cannot be established using steam generators AND HPI [Once Through] cooling initiated None CNB2 CETs > 1200°F AND Restoration procedures not effective within 15 min. (Note 1) C CTMT Radiation / RCS Activity FCB5 Containment High Range Radiation Monitor RE-8060/8061

[2RE-8925-1/ 8925-2]

> 750 [700] R/hr FCB6 Coolant activity

> 300 Ci/gm dose equivalent I-131 None RCB5 Containment High Range Radiation Monitor RE-8060/8061 [2RE-8925-1/8925-2] > 40 [50] R/hr None None CNB3 Containment High Range Radiation Monitor RE-8060/8061 [2RE-8925-1/8925-2] > 10,000 [12,000] R/hr to 0CAN031801 Page 120 of 255 Attachment 1 - Emergency Action Level Technical Bases Table 1[2]F-1 Fission Product Barrier Threshold Matrix Fuel Clad Barrier (FCB) Reactor Coolant System Barrier (RCB) Containment Barrier (CNB) Category Loss Potential Loss Loss Potential Loss Loss Potential Loss D CTMT Integrity or Bypass None None None None CNB4 Containment isolation is required AND EITHER: Containment integrity has been lost based on Emergency Director judgment UNISOLABLE pathway from Containment to the environment exists CNB5 Indications of RCS leakage outside of Containment CNB6 Containment pressure > 73.7 psia CNB7 Containment hydrogen concentration

> 3% CNB8 Containment pressure > 44.7 psia [23.3 psia] with < one full train of containment heat removal systems (Note 9) operating per design for 15 min. (Note 1) E Emergency Director Judgment FCB7 Any condition in the opinion of the Emergency Director that indicates loss of the Fuel Clad barrier FCB8 Any condition in the opinion of the Emergency Director that indicates potential loss of the Fuel Clad barrier RCB6 Any condition in the opinion of the Emergency Director that indicates loss of the RCS barrier RCB7 Any condition in the opinion of the Emergency Director that indicates potential loss of the RCS barrier CNB9 Any condition in the opinion of the Emergency Director that indicates loss of the Containment barrier CNB10 Any condition in the opinion of the Emergency Director that indicates potential loss of the Containment barrier to 0CAN031801 Page 121 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: A - RCS or S/G Tube Leakage Degradation Threat: Loss Threshold: None

to 0CAN031801 Page 122 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: A - RCS or S/G Tube Leakage Degradation Threat: Potential Loss Threshold: FCB1 RVLMS Levels 1 through 9 [1 through 7] indicate DRY Definition(s): None Basis: This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.

There is no Fuel Clad Barrier Loss threshold associated with RCS or S/G Tube Leakage.

Reference(s): 1. ULD SYS-24 Unit 1 Inadequate Core Cooling System 2. Calculation 84-EQ-0080-02 Loop Error Analysis for Reactor Vessel Level Monitoring System 3. ULD SYS-24 Unit 2 Inadequate Core Cooling Monitoring System

4. Calculation 90-E-0116-01 Unit 2 EOP Setpoint Document, Setpoint R.3 5. NEI 99-01 RCS or SG Tube Leakage Potential Loss 1.A to 0CAN031801 Page 123 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: B - Inadequate Heat Removal Degradation Threat: Loss Threshold: FCB2 CETs > 1200°F Definition(s): None Basis: This reading indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant. Reference(s): 1. NEI 99-01 Inadequate Heat Removal Loss 2.A to 0CAN031801 Page 124 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: B - Inadequate Heat Removal Degradation Threat: Potential Loss Threshold: FCB3 CETs > 700°F Definition(s): None Basis: This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage. Reference(s): 1. NEI 99-01 Inadequate Heat Removal Potential Loss 2.A to 0CAN031801 Page 125 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: B - Inadequate Heat Removal Degradation Threat: Potential Loss Threshold: FCB4 RCS heat removal cannot be established using steam generators AND HPI [Once Through] cooling initiated Definition(s): None Basis: This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the Fuel Clad Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using this threshold is not warranted. In combination with Potential Loss RCB4, meeting this threshold results in a Site Area Emergency. Reference(s): 1. OP-1202.004 Overheating 2. OP-1202.013 Figure 4, Core Exit Thermocouple for Inadequate Core Cooling

3. OP-2202.006 Loss of Feedwater 4. OP-2202.009 Functional Recovery, Safety Function Status Check 5 5. NEI 99-01 Inadequate Heat Removal Potential Loss 2.B to 0CAN031801 Page 126 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: C - CTMT Radiation / RCS Activity Degradation Threat: Loss Threshold: FCB5 Containment High Range Radiation Monitor RE-8060/8061 [2RE-8925-1/8925-2] > 750 [700] R/hr Definition(s): None Basis:

The containment radiation monitor reading (768[682] R/hr rounded to 750[700] R/hr for readability) corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300 µCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to 51.49[1.13]an approximate range of 2% to 3% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold C.1RCB5 since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the ECL to a Site Area Emergency.

There is no Potential Loss threshold associated with CTMT Radiation/RCS Activity /Containment Radiation. Basis Reference(s): 1. EP-CALC-ANO-1702 Containment High Range Radiation Monitor EAL Values 2. NEI 99-01 RCS Activity/Containment Radiation FC Loss 3.A to 0CAN031801 Page 127 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: C - CTMT Radiation / RCS Activity Degradation Threat: Loss Threshold: FCB6 Coolant activity > 300 Ci/gm dose equivalent I-131 Definition(s): None Basis: This threshold indicates that RCS radioactivity concentration is greater than 300 µCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier. It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample-related threshold is included as a backup to other indications. There is no Potential Loss threshold associated with CTMT Radiation/RCS Activity /Containment Radiation. Reference(s): 1. NEI 99-01 RCS Activity/Containment Radiation Fuel Clad Loss 3.B to 0CAN031801 Page 128 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: C - CTMT Radiation / RCS Activity Degradation Threat: Potential Loss Threshold: None

to 0CAN031801 Page 129 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: D - CTMT Integrity or Bypass Degradation Threat: Loss Threshold: None

to 0CAN031801 Page 130 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: D - CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold: None

to 0CAN031801 Page 131 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: E - Emergency Director Judgment Degradation Threat: Loss Threshold: FCB7 Any condition in the opinion of the Emergency Director that indicates loss of the Fuel Clad barrier Definition(s): None Basis:

This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is lost.

Reference(s):

1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A to 0CAN031801 Page 132 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: E - Emergency Director Judgment Degradation Threat: Potential Loss Threshold: FCB8 Any condition in the opinion of the Emergency Director that indicates potential loss of the Fuel Clad barrier Definition(s): None Basis:

This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Reference(s): 1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A

to 0CAN031801 Page 133 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: A - RCS or S/G Tube Leakage Degradation Threat: Loss Threshold: RCB1 An automatic or manual ESAS [ESFAS] actuation required by EITHER: UNISOLABLE RCS leakage S/G tube RUPTURE Definition(s): FAULTED - The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

RUPTURED - The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection (automatic or manual).

UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally. Basis: Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification. This threshold is based on an UNISOLABLE RCS leak of sufficient size to require an automatic or manual actuation of the Emergency Core Cooling System (ECCS). This condition clearly represents a loss of the RCS Barrier.

This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system.

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 a safety injection is considered to be RUPTURED. If a RUPTURED steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold 1.ACNB1 will also be met. to 0CAN031801 Page 134 of 255 Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. OP-1202.010 ESAS 2. OP-2202.003 Loss of Coolant Accident

3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Loss 1.A to 0CAN031801 Page 135 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: A - RCS or S/G Tube Leakage Degradation Threat: Potential Loss Threshold: RCB2 UNISOLABLE RCS leakage or S/G tube leakage > 50[44] gpm excluding normal reductions in RCS inventory (e.g., letdown, RCP seal leakoff) Definition(s): FAULTED - The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally. Basis: Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification. This threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizer level within specified limits by operation of a normally used makeup [charging] (makeup) pump, but an ECCS (SI)ESAS [ESFAS] actuation has not occurred. The threshold is met when letdown has been isolated and an operating procedure, or operating crew supervision, directs that a standby charging (makeup)makeup [charging] pump be placed in service to restore and maintain pressurizer level. This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. 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 declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold 1.ACNB1 will also be met. to 0CAN031801 Page 136 of 255 Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. 1SAR 9.1 Makeup and Purification System 2. 2SAR 9.3.4 Chemical and Volume Control System

3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.A to 0CAN031801 Page 137 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: A - RCS or S/G Tube Leakage Degradation Threat: Potential Loss Threshold: RCB3 Unit 1: PTS limits apply (RT14) AND RCS pressure and temperature are left of the NDTT/LTOP limit lines, on EOP Figure 3 (Note 12) Unit 2: Uncontrolled RCS cooldown (50°F step change which is below 500°F from NOT) AND RCS pressure and temperature are to the left of line B (200 degrees MTS), Standard Attachment 1, P-T Limits (Note 12) Note 12: Once PTS limits are first invoked, if RCS temperature and pressure are not brought within the limits within 15 minutes, this threshold is met and an immediate declaration is warranted. This threshold is met immediately upon exceeding the limits after this initial 15 minute period until PTS limits no longer apply. Definition(s): None Basis:

This condition indicates an extreme challenge to the integrity of the RCS pressure boundary due to pressurized thermal shock - a transient that causes rapid RCS cooldown while the RCS is in Mode 3 or higher (i.e., hot and pressurized).

to 0CAN031801 Page 138 of 255 Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. OP-1202.012 Repetitive Task 14 Control RCS Pressure 2. OP-1202.013 EOP Figures, Figure 3 RCS Pressure vs Temperature Limits

3. OP-1202.011 HPI Cooldown

4. Calculation No: 90-E-0116-01 ANO- EOP Setpoint Basis Document OP Setpoint P.2, RCS Pressure-Temperature 5. OP-2202.010 Standard Attachments, Attachment 1, P-T Limits 6. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.B to 0CAN031801 Page 139 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: B - Inadequate Heat Removal Degradation Threat: Loss Threshold: None

to 0CAN031801 Page 140 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: B - Inadequate Heat Removal Degradation Threat: Potential Loss Threshold: RCB4 RCS heat removal cannot be established using steam generators AND HPI [Once Through] cooling initiated Definition(s): None Basis: This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the RCS Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using this threshold is not warranted. In combination with Potential Loss FCB4, meeting this threshold results in a Site Area Emergency.

Meeting this threshold results in a Site Area Emergency because this threshold is identical to Fuel Clad Barrier Potential Loss threshold 2.BFCB4; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and increaseraise RCS pressure to the point where mass will be lost from the system. There is no RCS barrier Loss threshold associated with Inadequate Heat Removal. Reference(s): 1. OP-1202.004 Overheating 2. OP-1202.013 Figure 4, Core Exit Thermocouple for Inadequate Core Cooling 3. OP-2202.006 Loss of Feedwater

4. OP-2202.009 Functional Recovery, Safety Function Status Check 5

5. NEI 99-01 Inadequate Heat Removal RCS Potential Loss 2.B to 0CAN031801 Page 141 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: C - CTMT Radiation/ RCS Activity Degradation Threat: Loss Threshold: RCB5 Containment High Range Radiation Monitor RE-8060/8061 [2RE-8925-1/8925-2] > 40 [50] R/hr Definition(s): None Basis: NRC Information Notice 97-045, Supplement 1, identifies the potential for erratic indications from the high range radiation monitors (HRRMs) as a result of thermally induced currents (TIC) which may cause the HRRM to read falsely high (for approximately 15 minutes) on a rapid temperature rise, and fail low intermittently on a rapid temperature fall. Because of this phenomenon, any trends or alarms on the HRRM's should be validated by comparison to the containment low range/area radiation monitors and Air Monitoring Systems trends before actions are taken. The containment radiation monitor reading (42.8[50.4] R/hr rounded to 40[50] R/hr for readability) corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold 3.AFCB5 since it indicates a loss of the RCS Barrier only.

There is no Potential Loss threshold associated with RCS Activity / ContainmentCTMT Radiation/RCS Activity. Reference(s): 1. EP-CALC-ANO-1702 Containment High Range Radiation Monitor EAL Values 2. NEI 99-01 CMT Radiation / RCS Activity RCS Loss 3.A to 0CAN031801 Page 142 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: B - CTMT Radiation/ RCS Activity Degradation Threat: Potential Loss Threshold: None

to 0CAN031801 Page 143 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: D - CTMT Integrity or Bypass Degradation Threat: Loss Threshold: None

to 0CAN031801 Page 144 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: D - CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold: None

to 0CAN031801 Page 145 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: E - Emergency Director Judgment Degradation Threat: Loss Threshold: RCB6 Any condition in the opinion of the Emergency Director that indicates loss of the RCS barrier Definition(s): None Basis: This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is lost. Reference(s): 1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A to 0CAN031801 Page 146 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: E - Emergency Director Judgment Degradation Threat: Potential Loss Threshold: RCB7 Any condition in the opinion of the Emergency Director that indicates potential loss of the RCS barrier Definition(s): None Basis:

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Reference(s): 1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6.A

to 0CAN031801 Page 147 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: A - RCS or S/G Tube Leakage Degradation Threat: Loss Threshold: CNB1 A S/G that is leaking > 50[44] gpm (excluding normal reductions in RCS inventory) or that is RUPTURED S/G is also FAULTED outside of containment Definition(s): FAULTED - The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

RUPTURED - The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection (automatic or manual).

Basis:

This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outside of containment. The condition of the SG, whether leaking or RUPTURED, is determined in accordance with the thresholds for RCS Barrier Potential Loss 1.A RCB2 and Loss 1.ARCB1, respectively. This condition represents a bypass of the containment barrier. FAULTED is a defined term within the NEI 99-01 methodology; this determination is not necessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steam generator is droppecreasing uncontrollably ([part of the FAULTED definition)] and the FAULTED steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FAULTED for emergency classification purposes.

The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification. Steam releases of this size are readily observable with normal Control Room indications. The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU4 for the fuel clad barrier (i.e., RCS activity values) and IC SU5 for the RCS barrier (i.e., RCS leak rate values).

This threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown the plant, or to drive an auxiliary (emergency) feed water pump. These types of conditions will result in a significant and sustained release of radioactive to 0CAN031801 Page 148 of 255 Attachment 1 - Emergency Action Level Technical Bases steam to the environment (and are thus similar to a FAULTED condition). The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss of containment.

Steam releases associated with the expected operation of a SG power operated relief valve or safety relief valve do not meet the intent of this threshold. Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown. Steam releases associated with the unexpected operation of a valve (e.g., a stuck-open safety valve) do meet this threshold.

Following an SG tube leak or rupture, there may be minor radiological releases through a secondary-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 but should be evaluated using the Recognition Category A ICs.

The emergency classification levelECLs resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below. Affected SG is FAULTED Outside of Containment? P-to-S Leak Rate Yes No Less than or equal to 25 gpm No classification No classification Greater than 25 gpm Unusual Event per SU4SU5.1 Unusual Event per SU4SU5.1 Greater than 50[44] gpm (RCS Barrier Potential Loss) Site Area Emergency per FS1.1 Alert per FA1.1 Requires an automatic or manual ESAS [ESFAS]ECCS (SIAS) actuation (RCS Barrier Loss) Site Area Emergency per FS1.1 Alert per FA1.1 There is no Potential Loss threshold associated with RCS or S/G Tube Leakage. Reference(s): 1. NEI 99-01 RCS or SG Tube Leakage Containment Loss 1.A

to 0CAN031801 Page 149 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: A - RCS or S/G Tube Leakage Degradation Threat: Potential Loss Threshold: None

to 0CAN031801 Page 150 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: B - Inadequate Heat Removal Degradation Threat: Loss Threshold: None

to 0CAN031801 Page 151 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: B - Inadequate Heat Removal Degradation Threat: Potential Loss Threshold: CNB2 CETs > 1200°F AND Restoration procedures not effective within 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Definition(s): IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: The restoration procedure is considered "effective" if core exit thermocouple readings are droppdecreasing and/or if reactor vessel level is risincreasing. Whether or not the procedure(s) will be effective should be apparent within 15 minutes. The Emergency Director should escalate the emergency classification level as soon as it is determined that the procedure(s) will not be effective.

This condition represents an IMMINENT core melt sequence which, if not corrected, could lead to vessel failure and an increased potential for containment failure. For this condition to occur, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. If implementation of a procedure(s) to restore adequate core cooling is not effective (successful) within 15 minutes, it is assumed that the event trajectory will likely lead to core melting and a subsequent challenge of the Containment Barrier.

Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation in a significant fraction of core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence.

Reference(s): 1. NEI 99-01 Inadequate Heat Removal Containment Potential Loss 2.A to 0CAN031801 Page 152 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: C - CTMT Radiation/RCS Activity Degradation Threat: Loss Threshold: None

to 0CAN031801 Page 153 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: C - CTMT Radiation/RCS Activity Degradation Threat: Potential Loss Threshold: CNB3 Containment High Range Radiation Monitor RE-8060/8061 [2RE-8925-1/8925-2] > 10,000 [12,000] R/hr Definition(s): None Basis:

The containment radiation monitor reading (10,300[12,100] R/hr rounded to 10,000[12,000] R/hr for readability) corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds. NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency classification levelECL to a General Emergency. There is no Loss threshold associated with RCS Activity/ContainmentCTMT Radiation/RCS Activity. Reference(s): 1. EP-CALC-ANO-1702 Containment High Range Radiation Monitor EAL Values 2. NEI 99-01 CTMT Radiation / RCS Activity Containment Potential Loss 3.A to 0CAN031801 Page 154 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: D - CTMT Integrity or Bypass Degradation Threat: Loss Threshold: CNB4 Containment isolation is required AND EITHER: Containment integrity has been lost based on Emergency Director judgment UNISOLABLE pathway from Containment to the environment exists Definition(s):

UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally. Basis: Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification. The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold 1.ACNB1. These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. Users are reminded that there may be accident and release conditions that simultaneously meet both bulleted thresholds 4.A.1 and 4.A.2. 4.A.1First Threshold - Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage (or sometimes referred 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 integrity condition may (or may not) be accompanied by a noticeable drop in containment pressure.

Recognizing the inherent difficulties in determining a containment leak rate during accident conditions, it is expected that the Emergency Director will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g., containment pressure, readings on radiation monitors outside containment, operating status of containment pressure control equipment, etc.).

to 0CAN031801 Page 155 of 255 Attachment 1 - Emergency Action Level Technical Bases Refer to the middle piping run of Figure 9-F-41. Two simplified examples are provided. One is leakage from a penetration and the other is leakage from an in-service system valve. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure.

Another example would be a loss or potential loss of the RCS barrier, and the simultaneous occurrence of two FAULTED locations on a steam generator where one fault is located inside containment (e.g., on a steam or feedwater line) and the other outside of containment. In this case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.

Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category A ICs.

4.A.2Second Threshold - Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment. As used here, the term "environment" includes the atmosphere of a room or area, outside the containment, 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 variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure. Refer to the top piping run of Figure 9-F-41. In this simplified example, the inboard and outboard isolation valves remained open after a containment isolation was required (i.e., containment isolation was not successful). There is now an UNISOLABLE pathway from the containment to the environment.

The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.

Leakage between two interfacing liquid systems, by itself, does not meet this threshold.

Refer to the bottom piping run of Figure 9-F-41. In this simplified example, leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building. The radioactivity would be detected by the Process Monitor. If there is no leakage from the closed water cooling system to the Auxiliary Building, then no threshold has been met. If the pump developed a leak that allowed steam/water to enter the Auxiliary Building, then second threshold 4.B would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause the first threshold 4.A.1 to be met as well. to 0CAN031801 Page 156 of 255 Attachment 1 - Emergency Action Level Technical Bases Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable containment leakage through various penetrations or system components. Minor releases may also occur if a containment isolation valve(s) fails to close but the containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category A ICs.

Reference(s): 1. NEI 99-01 CTMT Integrity or Bypass Containment Loss 4.A

to 0CAN031801 Page 157 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: D - CTMT Integrity or Bypass Degradation Threat: Loss Threshold: CNB5 Indications of RCS leakage outside of Containment Definition(s): None Basis: To ensure proper escalation of the emergency classification, the RCS leakage outside of containment must be related to the mass loss that is causing the RCS Loss RCB1 and/or Potential Loss RCB2 threshold 1.A to be met. The status of the containment barrier during an event involving steam generator tube leakage is assessed using Containment Loss Threshold CNB1. Containment sump, temperature, pressure and/or radiation levels will riincrease if reactor coolant mass is leaking into the containment. If these parameters have not risenincreased, then the reactor coolant mass may be leaking outside of containment (i.e., a containment bypass sequence). RiIncreases in sump, temperature, pressure, flow and/or radiation level readings outside of the containment may indicate that the RCS mass is being lost outside of containment. Unexpected elevated readings and alarms on radiation monitors with detectors outside containment should be corroborated with other available indications to confirm that the source is 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 riincrease 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 occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure and cause threshold CNB4 to be met as well.

Reference(s): 1. NEI 99-01 CTMT Integrity or Bypass Containment Loss 4.B to 0CAN031801 Page 158 of 255 Attachment 1 - Emergency Action Level Technical Bases Figure 1: Containment Integrity or Bypass Examples

Open valveOpen valveOpen valve Open valve Open valve Open valvePenetrationDamperDamper Interface leakage RCP Seal Cooling Inside Reactor Building Auxiliary Building1st Threshold - Airborne 1st Threshold - Airborne release from penetration2nd Threshold - Airborne release from pathway2nd Threshold - RCS leakage outside ABAirborne MonitorArea MonitorProcess MonitorEffluent MonitorClosed CoolingPump to 0CAN031801 Page 159 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: D - CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold: CNB6 Containment pressure > 73.7 psia Definition(s): None Basis:

If containment pressure exceeds the design pressure, there exists a potential to lose the Containment Barrier. To reach this level, there must be an inadequate core cooling condition for an extended period of time; therefore, the RCS and Fuel Clad barriers would already be lost.

Thus, this threshold is a discriminator between a Site Area Emergency and General Emergency since there is now a potential to lose the third barrier. Reference(s): 1. 1SAR 1.4.43 Criterion 50 - Containment Design Basis 2. 2SAR Table 6.2-7 Principle Containment Design Parameters

3. NEI 99-01 CTMT Integrity or Bypass Containment Potential Loss 4.A to 0CAN031801 Page 160 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: D - CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold: CNB7 Containment hydrogen concentration > 3% Definition(s): None Basis: The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a potential loss of the Containment Barrier.

Reference(s): 1. Unit 1 SAMG Figure III-1B 2. Unit 2 SAMG Phase 1 Instructions, Containment Flowchart

3. NEI 99-01 CTMT Integrity or Bypass Containment Potential Loss 4.B to 0CAN031801 Page 161 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: D - CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold: CNB8 Containment pressure > 44.7 psia [23.3 psia] with < one full train of containment heat removal systems (Note 9) operating per design for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 9: One full train of containment heat removal systems consists of one train of RB [Containment] Spray and one train of RB [Containment] Cooling System. Definition(s): None Basis:

This threshold describes a condition where containment pressure is greater than the setpoint at 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-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. This threshold represents a potential loss of containment in that containment heat removal/depressurization systems (e.g., containment sprays, ice condenser fans, etc., but not including containment venting strategies) are either lost or performing in a degraded manner. Reference(s):

1. 1SAR 6.2 Reactor Building Spray System 2. 1SAR 6.3 Reactor Building Cooling System

3. OP-2202.003 Loss of Coolant Accident

4. OP-2202.010 Standard Attachments, Attachment 22

5. 2SAR 6.2.2 Containment Heat Removal Systems 6. 2SAR 7.3.1.1.11.2 Containment Spray System 7. NEI 99-01 CTMT Integrity or Bypass Containment Potential Loss 4.C to 0CAN031801 Page 162 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: E - Emergency Director Judgment Degradation Threat: Loss Threshold: CNB9 Any condition in the opinion of the Emergency Director that indicates loss of the Containment barrier Definition(s): None Basis:

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is lost.

Reference(s):

1. NEI 99-01 Emergency Director Judgment Containment Loss 6.A to 0CAN031801 Page 163 of 255 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: E - Emergency Director Judgment Degradation Threat: Potential Loss Threshold: CNB10 Any condition in the opinion of the Emergency Director that indicates potential loss of the Containment barrier Definition(s): None Basis:

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Reference(s): 1. NEI 99-01 Emergency Director Judgment Containment Potential Loss 6.A

to 0CAN031801 Page 164 of 255 Attachment 1 - Emergency Action Level Technical Bases Category H - Hazards and Other Conditions Affecting Plant Safety EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

Hazards are non-plant, system-related events that can directly or indirectly affect plant operation, reactor plant safety or personnel safety. 1. Security Unauthorized entry attempts into the PROTECTED AREA, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant. 2. Seismic Event Natural events such as earthquakes have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety. 3. Natural or Technological Hazard Other natural and non-naturally occurring events that can cause damage to plant facilities include tornados, FLOODING, hazardous material releases and events restricting site access warranting classification. 4. Fire FIRES can pose significant hazards to personnel and reactor safety. Appropriate for classification are FIRES within the plant PROTECTED AREA or which may affect operability of equipment needed for safe shutdown. 5. Hazardous Gas Toxic, corrosive, asphyxiant or flammable gas leaks can affect normal plant operations or preclude access to plant areas required to safely shutdown the plant. 6. Control Room Evacuation Events that are indicative of loss of Control Room habitability. If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities. 7. Emergency Director Judgment The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification.

While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the Emergency Director the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Director judgment. to 0CAN031801 Page 165 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL: HU1.1 Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by ANO Security Shift Supervision OR Notification of a credible security threat directed at the site OR A validated notification from the NRC providing information of an aircraft threat Mode Applicability:

All Definition(s): HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station.

HOSTILE ACTION - An act toward ANO or its personnel that includes the use of violent force 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 other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on ANO. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA).

OWNER CONTROLLED AREA - For the purposes of classification this is the Security Owner Controlled Area (SOCA). The SOCA is demarcated as a Vehicle Barrier System (VBS) and a detection fence on the outside and a delay fence on the inside of the passive and active barriers. The SOCA is the area inside the SOCA VBS up to the PROTECTED AREA fence line. Access to this area is controlled by the SOCA Personnel Access Control Point. PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

to 0CAN031801 Page 166 of 255 Attachment 1 - Emergency Action Level Technical Bases PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the 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 result in potential offsite exposures.

SECURITY CONDITION - Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a HOSTILE ACTION. Basis: This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR 73.71 or 10 CFR 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, and HS1 and HG1. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].

The first threshold EAL #1 references the Security Shift Supervision (site-specific security shift supervision)because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR 2.39 information.

The second threshold EAL #2 addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with OP-1203.048 Security Event (site-specific procedure). to 0CAN031801 Page 167 of 255 Attachment 1 - Emergency Action Level Technical Bases The third threshold EAL #3 addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with 11-S 1 Security Contingency Events (ref. 2)(site-specific procedure). Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Security Plan for ANO (ref. 1). Escalation of the emergency classification level would be via IC HA1. Reference(s): 1. ANO Security Plan 2. OP-1203.048 Security Event 3. NEI 99-01 HU1 to 0CAN031801 Page 168 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL: HA1.1 Alert A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by ANO Security Shift Supervision OR A validated notification from NRC of an aircraft attack threat within 30 min. of the site Mode Applicability: All Definition(s): HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION - An act toward ANO or its personnel that includes the use of violent force 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 other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on ANO. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA).

HOSTILE FORCE - One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

OWNER CONTROLLED AREA - For the purposes of classification this is the Security Owner Controlled Area (SOCA). The SOCA is demarcated as a Vehicle Barrier System (VBS) and a detection fence on the outside and a delay fence on the inside of the passive and active barriers. The SOCA is the area inside the SOCA VBS up to the PROTECTED AREA fence line. Access to this area is controlled by the SOCA Personnel Access Control Point.

PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. to 0CAN031801 Page 169 of 255 Attachment 1 - Emergency Action Level Technical Bases PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. Basis: This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact.

Timely and accurate communications between Security Shift Supervision and the 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, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program]. As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of Offsite Response Organizations (OROs), allowing them to be better prepared should it be necessary to consider further actions. This IC EAL 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. Examples include 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 the requirements of 10 CFR 73.71 or 10 CFR 50.72. The first threshold EAL #1 is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes any action directed against an ISFSI that is located outside the plant PROTECTED AREA. The second threshold EAL #2 addresses the threat from the impact of an aircraft on the plant, and the anticipated 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 a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with OP-1203.048 Security Event (ref. 2)(site-specific procedure). The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD through the NRC.

In some cases, it may not be readily apparent if an aircraft impact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The to 0CAN031801 Page 170 of 255 Attachment 1 - Emergency Action Level Technical Bases emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency. Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Security Plan for ANO (ref. 1). Escalation of the emergency classification level would be via IC HS1.

Reference(s): 1. ANO Security Plan 2. OP-1203.048 Security Event

3. NEI 99-01 HA1 to 0CAN031801 Page 171 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the PROTECTED AREA EAL: HS1.1 Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by ANO Security Shift Supervision Mode Applicability: All Definition(s):

HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station.

HOSTILE ACTION - An act toward ANO or its personnel that includes the use of violent force 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 other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on ANO. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA). HOSTILE FORCE - One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

OWNER CONTROLLED AREA - For the purposes of classification this is the Security Owner Controlled Area (SOCA). The SOCA is demarcated as a Vehicle Barrier System (VBS) and a detection fence on the outside and a delay fence on the inside of the passive and active barriers. The SOCA is the area inside the SOCA VBS up to the PROTECTED AREA fence line. Access to this area is controlled by the SOCA Personnel Access Control Point. PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. to 0CAN031801 Page 172 of 255 Attachment 1 - Emergency Action Level Technical Bases 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 plant equipment.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 1, 2). Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program]. As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Site Area Emergency declaration will mobilize Offsite Response Organization (ORO) resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions. This IC EAL does not apply to a HOSTILE ACTION directed at an ISFSI Protected Area located outside the PROTECTED AREA; such an attack should be assessed using IC HA1. 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. Examples include 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 the requirements of 10 CFR 73.71 or 10 CFR 50.72.

Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Security Plan for ANO (ref. 1). Escalation of the emergency classification level would be via IC HG1. Reference(s): 1. ANO Security Plan 2. OP-1203.048 Security Event

3. NEI 99-01 HS1 to 0CAN031801 Page 173 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 2 - Seismic Event Initiating Condition: Seismic event greater than OBE levels EAL: HU2.1 Unusual Event Seismic event > OBE as indicated by annunciation of the 0.10 g acceleration alarm Mode Applicability: All Definition(s): None Basis: This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE). An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE)Design Basis Earthquake (DBE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the 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 and inspections, and fully understand any impacts, this event represents a potential degradation of the 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 a seismic event (e.g., lateral accelerations in excess of 0.08g1g). The Shift Manager or Emergency Director may seek external verification if deemed appropriate (e.g., a call to the USGS, check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9. Two strong motion triaxial accelerometers, ACS-8001 and ACS-8003, located at the base slab provide alarms to the Unit 1 control room via the seismic network control center, C529-NCC. One alarm from C529-NCC is triggered when a setpoint of 0.01g has been exceeded. This alarm indicates that an earthquake has occurred and the seismic monitoring system is recording seismic data. Another alarm from C529-NCC is triggered when the pre-determined value of 0.1g, indicating the OBE has been exceeded (ref. 2, 3). to 0CAN031801 Page 174 of 255 Attachment 1 - Emergency Action Level Technical Bases To avoid inappropriate emergency classification resulting from spurious actuation of the seismic instrumentation or felt motion not attributable to seismic activity, an offsite agency (USGS, National Earthquake Information Center (NEIC)) can confirm that an earthquake has occurred in the area of the plant. Such confirmation should not, however, preclude a timely emergency declaration based on receipt of the OBE alarm. If requested, provide the analyst with the following ANO coordinates: 35º 18' 36" north latitude, 93º 13' 53" west longitude (ref. 4). Alternatively, near real-time seismic activity can be accessed via the NEIC website: Reference(s): 1. 1SAR 2.2.1 Location 2. 1SAR 2.7.2 Site Seismic Evaluation 3. 1SAR 2.7.6 Time-History Accelerograph

4. OP-1203.025 Natural Emergencies

5. OP-2203.008 Natural Emergencies

6. NEI 99-01 HU2 to 0CAN031801 Page 175 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.1 Unusual Event A tornado strike within the PROTECTED AREA Mode Applicability: All Definition(s): PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

This EALEAL #1 addresses a tornado striking (touching down) within the PROTECTED AREA. EAL #2 addresses flooding of a building room or area that results in operators isolating power 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 isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode. EAL #3 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA. EAL #4 addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal 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 blocking the access road. This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011. to 0CAN031801 Page 176 of 255 Attachment 1 - Emergency Action Level Technical Bases EAL #5 addresses (site-specific description). Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, S or C.

If damage is confirmed visually or by other in-plant indications, the event may be escalated to an Alert under EAL CA6.1 or SA9.1. A tornado striking (touching down) within the PROTECTED AREA warrants declaration of an Unusual Event regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base of a thunderstorm. Reference(s): 1. OP-1203.025 Natural Emergencies 2. OP-2203.008 Natural Emergencies 3. NEI 99-01 HU3 to 0CAN031801 Page 177 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.2 Unusual Event Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component required by Technical Specifications for the current operating mode Mode Applicability: All Definition(s): FLOODING - A condition where water is entering a room or area faster than installed equipment 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/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the 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 result in potential offsite exposures.

Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

EAL #1 addresses a tornado striking (touching down) within the PROTECTED AREA. This EAL addresses FLOODING of a building room or area that results in operators isolating power 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 isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To to 0CAN031801 Page 178 of 255 Attachment 1 - Emergency Action Level Technical Bases warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode. EAL #3 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA. EAL #4 addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal 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 blocking the access road. This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011. EAL #5 addresses (site-specific description). Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, S or C.

Refer to EAL CA6.1 or SA9.1 for internal FLOODING affecting more than one SAFETY SYSTEM train. Reference(s): 1. OP-1203.025 Natural Emergencies 2. OP-2203.008 Natural Emergencies

3. NEI 99-01 HU3 to 0CAN031801 Page 179 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.3 Unusual Event Movement of personnel within the PROTECTED AREA is IMPEDED due to an event external to the PROTECTED AREA involving hazardous materials (e.g., an offsite chemical spill or toxic gas release) Mode Applicability: All Definition(s): IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures 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 (i.e., the security fence) and to which access is controlled.

Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

This EAL EAL #1 addresses a tornado striking (touching down) within the PROTECTED AREA. This EAL addresses flooding of a building room or area that results in operators isolating power 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 isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode. EAL #3 addresses a hazardous materials event originating at an offsite location outside the PROTECTED AREA and of sufficient magnitude to IMPEDE the movement of personnel within the PROTECTED AREA.

EAL #4 addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles. to 0CAN031801 Page 180 of 255 Attachment 1 - Emergency Action Level Technical Bases 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 blocking the access road. This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011. EAL #5 addresses (site-specific description). Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, S or C. Reference(s): 1. NEI 99-01 HU3

to 0CAN031801 Page 181 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.4 Unusual Event A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing 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 or accidents. Mode Applicability: All Definition(s):

FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.

Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. EAL #1 addresses a tornado striking (touching down) within the PROTECTED AREA. This EAL addresses flooding of a building room or area that results in operators isolating power 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 isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode. EAL #3 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA. This EAL EAL #4 addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal 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 blocking the access road.

to 0CAN031801 Page 182 of 255 Attachment 1 - Emergency Action Level Technical Bases This EAL is not intended to apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the FLOODING around the Cooper Station during the Midwest floods of 1993, or the FLOODING around Ft. Calhoun Station in 2011.

EAL #5 addresses (site-specific description). Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, S or C. Reference(s): 1. NEI 99-01 HU3 to 0CAN031801 Page 183 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.1 Unusual Event A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (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 alarm AND The FIRE is located within any Table 1[2]H-1 area Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table 1H-1 Unit 1 Fire Areas Reactor Building All elevations Auxiliary Building All elevations including: Penthouse/MSIV Room Exceptions: Boric Acid Mix Tank Room (Chem Add Area), 404' (157-B), EDG Exhaust Fan area on 386' (1-E and 2-E)

Turbine Building All elevations including: Pipechase under ICW Coolers, CRD Pump Pit/T-28 Room/Area under ICW Pumps Outside Areas Manholes adjacent to Startup #2 XFMR (MH-03/MH-04) Manholes adjacent to Intake Structure (MH-05/MH-06) Intake Structure (354' and 366') Diesel Fuel Vault Diesel Fuel Vault Pump Manholes (MH-09 and MH-10) to 0CAN031801 Page 184 of 255 Attachment 1 - Emergency Action Level Technical Bases Table 2H-1 Unit 2 Fire Areas Reactor Building All elevations Auxiliary Building All elevations including: Aux Extension Turbine Building All elevations Outside Areas Intake Structure (354' and 366') Concrete Manhole East, NE of intake (2MH-01) Concrete Manhole East of Turbine Building next to train bay (2MH-03) Diesel Fuel Vault Diesel Fuel Vault Pump Manholes (MH-09 and MH-10) Mode Applicability: All

Definition(s): FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. EAL #1 The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc. to 0CAN031801 Page 185 of 255 Attachment 1 - Emergency Action Level Technical Bases 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 at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also 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 take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed. A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress. If an actual FIRE is verified by a report from the field, then EAL #1 is immediately applicable, 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 spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted. EAL #3 In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. This basis extends to a FIRE occurring within the PROTECTED AREA of an ISFSI located outside the plant PROTECTED AREA. [Sentence for plants with an ISFSI outside the plant Protected Area] EAL #4 If a FIRE within the plant or ISFSI [for plants with an ISFSI outside the plant Protected Area] 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. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions. Basis-Related Requirements from Appendix R Appendix R to 10 CFR 50, states in part: to 0CAN031801 Page 186 of 255 Attachment 1 - Emergency Action Level Technical Bases Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions." When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage 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 of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents. In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in EAL #2, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC EAL CA6.1 or SA9SA9.1. The 15 minute requirement begins with a credible notification that a FIRE is occurring, or receipt of multiple VALID fire detection system alarms or field validation of a single fire alarm. The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field. Table 1[2]H-1 Fire Areas are those areas that contain equipment necessary for safe operation and shutdown of the plant (ref. 1, 2). Reference(s): 1. OP-1203.049 Fires in Areas Affecting Safe Shutdown 2. OP- 2203.049 Fires in Areas Affecting Safe Shutdown 3. NEI 99-01 HU4 to 0CAN031801 Page 187 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.2 Unusual Event Receipt of a single fire alarm (i.e., no other indications of a FIRE) AND The fire alarm is indicating a FIRE within any Table 1[2]H-1 area AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table 1H-1 Unit 1 Fire Areas Reactor Building All elevations Auxiliary Building All elevations including: Penthouse/MSIV Room Exceptions: Boric Acid Mix Tank Room (Chem Add Area), 404' (157-B), EDG Exhaust Fan area on 386' (1-E and 2-E) Turbine Building All elevations including: Pipechase under ICW Coolers, CRD Pump Pit/T-28 Room/Area under ICW Pumps Outside Areas Manholes adjacent to Startup #2 XFMR (MH-03/MH-04) Manholes adjacent to Intake Structure (MH-05/MH-06) Intake Structure (354' and 366')

Diesel Fuel Vault Diesel Fuel Vault Pump Manholes (MH-09 and MH-10) to 0CAN031801 Page 188 of 255 Attachment 1 - Emergency Action Level Technical Bases Table 2H-1 Unit 2 Fire Areas Reactor Building All elevations Auxiliary Building All elevations including: Aux Extension Turbine Building All elevations Outside Areas Intake Structure (354' and 366') Concrete Manhole East, NE of intake (2MH-01) Concrete Manhole East of Turbine Building next to train bay (2MH-03) Diesel Fuel Vault Diesel Fuel Vault Pump Manholes (MH-09 and MH-10) Mode Applicability: All Definition(s): FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.

EAL #1 The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc. to 0CAN031801 Page 189 of 255 Attachment 1 - Emergency Action Level Technical Bases 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 at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also 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 take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed.

A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.

If an actual FIRE is verified by a report from the field, then HU4.1 EAL #1 is immediately applicable, 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 spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted. EAL #3 In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. This basis extends to a FIRE occurring within the PROTECTED AREA of an ISFSI located outside the plant PROTECTED AREA. [Sentence for plants with an ISFSI outside the plant Protected Area] EAL #4 Basis-Related Fire Protection Requirements from Appendix R Appendix R to 10 CFR 50, states in part: Criterion 3 of 10 CFR 50, Appendix A, states, in part: "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions." In this respect, noncombustible and heat resistant materials are used wherever practical throughout the unit, particularly in locations such as the containment and Control Room. Fire detection and fighting systems of appropriate capacity and capability are provided and designed to minimize the adverse effects of fires on SSCs important to safety. Firefighting systems are to 0CAN031801 Page 190 of 255 Attachment 1 - Emergency Action Level Technical Bases designed to assure that the rupture or inadvertent operation of a fire system does not significantly impair the safety capability of these structures, systems, and components. When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage 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 of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents. In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train is employed(G.2.c). As used in HU4.2EAL #2, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC EAL CA6.1 or SA9SA9.1. 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 it is not spurious, or by reports from the field. Actual field reports must be made within the 30-minute time limit or a classification must be made. If a fire is verified to be occurring by field report, classification shall be made based on EAL HU4.1, with the 15-minute requirement beginning with the verification of the fire by field report. Table 1[2]H-1 Fire Areas are those areas that contain equipment necessary for safe operation and shutdown of the plant (ref. 1, 2). Reference(s): 1. OP-1203.049 Fires in Areas Affecting Safe Shutdown 2. OP- 2203.049 Fires in Areas Affecting Safe Shutdown 3. NEI 99-01 HU4 to 0CAN031801 Page 191 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.3 Unusual Event A FIRE within the PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: All Definition(s): FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled.

Basis:

This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.

EAL #1 The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation 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 at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report. to 0CAN031801 Page 192 of 255 Attachment 1 - Emergency Action Level Technical Bases 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 take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed. A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress. If an actual FIRE is verified by a report from the field, then HU4.1 EAL #1 is immediately applicable, 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 spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted. EAL #3 In addition to a FIRE addressed by EAL HU4.1 #1 or HU4.2EAL #2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety.

This basis extends to a FIRE occurring within the PROTECTED AREA of an ISFSI located outside the plant PROTECTED AREA. [Sentence for plants with an ISFSI outside the plant Protected Area] EAL #4 If a FIRE within the plant or ISFSI [for plants with an ISFSI outside the plant Protected Area] 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. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions. Basis-Related Requirements from Appendix R Appendix R to 10 CFR 50, states in part: Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions." to 0CAN031801 Page 193 of 255 Attachment 1 - Emergency Action Level Technical Bases When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage 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 of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents. In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in EAL #2, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC EAL CA6.1 or SA9SA8.1. Reference(s): 1. NEI 99-01 HU4

to 0CAN031801 Page 194 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.4 Unusual Event A FIRE within the PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Mode Applicability: All Definition(s):

FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.

PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. EAL #1 The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation 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 at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report. EAL #2 to 0CAN031801 Page 195 of 255 Attachment 1 - Emergency Action Level Technical Bases 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 take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed. A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress. If an actual FIRE is verified by a report from the field, then HU4.1 EAL #1 is immediately applicable, 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 spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted. EAL #3 In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. This basis extends to a FIRE occurring within the PROTECTED AREA of an ISFSI located outside the plant PROTECTED AREA. [Sentence for plants with an ISFSI outside the plant Protected Area] EAL #4 If a FIRE within the plant or ISFSI [for plants with an ISFSI outside the plant Protected Area] 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. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions. Basis-Related Requirements from Appendix R Appendix R to 10 CFR 50, states in part: Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions." When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage 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 of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not to 0CAN031801 Page 196 of 255 Attachment 1 - Emergency Action Level Technical Bases per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents. In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in EAL #2, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC EAL CA6.1 or SA9SA9.1. Reference(s): 1. NEI 99-01 HU4

to 0CAN031801 Page 197 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 5 - Hazardous Gas Initiating Condition: Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown EAL: HA5.1 Alert Release of a toxic, corrosive, asphyxiant or flammable gas into any Table 1[2]H-2 room or area AND Entry 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, then no emergency classification is warranted. Table 1H-2 Unit 1 Safe Operation & Shutdown Rooms/Areas Room/Area Mode A-4 Switchgear Room 3, 4 Upper North Electrical Penetration Room 3, 4 Lower South Electrical Equipment Room 3, 4 Table 2H-2 Unit 2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Aux Building 317' Emergency Core Cooling Rooms 3, 4 Aux Building 317' Tendon Gallery Access 3, 4 Aux Building 335' Charging Pumps / MCC 2B-52 3, 4 Aux Building 354' MCC 2B-62 Area 3, 4 Emergency Diesel Generator Corridor 3, 4 Lower South Piping Penetration Room 3, 4 Aux Building 386' Containment Hatch 3, 4 to 0CAN031801 Page 198 of 255 Attachment 1 - Emergency Action Level Technical Bases Mode Applicability:

3 - Hot Standby, 4 - Hot Shutdown Definition(s): IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures 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). Basis:

This IC addresses an event involving a release of a hazardous gas that precludes or IMPEDES access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown. This condition represents an actual or potential substantial degradation of the level of safety of the plant.

An Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the gaseous release. The emergency classification is not contingent upon whether entry is actually necessary at the time of the release.

Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the Emergency Director's judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly IMPEDE procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as IMPEDED if extraordinary measures 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).

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 affected room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous 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 require entry into the affected room until Mode 4. The gas release is a planned activity that includes compensatory measures which address the 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 keeping nature (e.g., normal rounds or routine inspections). to 0CAN031801 Page 199 of 255 Attachment 1 - Emergency Action Level Technical Bases The access control measures are of a conservative or precautionary nature, and would not actually prevent or IMPEDE a required action. If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event. An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels.

Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death.

This EAL does not apply to firefighting activities that generate smoke and that automatically or manually activate a fire suppression system in an area , or to intentional inerting of containment. (BWR only). The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1). Escalation of the emergency classification level would be via Recognition Category A, C or F ICs. EAL HA5.1 mode applicability has been limited to the mode limitations of Table 1[2]H-2 (Modes 3 and 4 only). Reference(s):

1. Attachment 2 Safe Operation & Shutdown Areas Tables 1[2]A-3 & 1[2]H-2 Bases 2. NEI 99-01 HA5 to 0CAN031801 Page 200 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations EAL: HA6.1 Alert An event has resulted in plant control being transferred from the Control Room to alternate locations Mode Applicability: All Definition(s): None Basis:

This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety.

Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown 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, will present challenges to plant operators and other on-shift personnel. Activation of the ERO and emergency response facilities will assist in responding to these challenges.

Transfer of plant control begins when the last licensed operator leaves the Control Room.

Escalation of the emergency classification level would be via IC HS6.

Reference(s):

1. OP-1203.002 Alternate Shutdown 2. OP- 2203.014 Alternate Shutdown

3. NEI 99-01 HA6 to 0CAN031801 Page 201 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Inability to control a key safety function from outside the Control Room EAL: HS6.1 Site Area Emergency An event has resulted in plant control being transferred from the Control Room to alternate locations AND Control of any of the following key safety functions is not re-established within 15 min. (Note 1): Reactivity (Modes 1, 2 and 3 only) Core cooling RCS heat removal Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling

Definition(s): None Basis:

This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time. The determination of whether or not "control" is established at the remote safe shutdown location(s) is based on Emergency Director judgment. The Emergency Director is expected to make a reasonable, informed judgment within 15 (the site-specific time for transfer) minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s). to 0CAN031801 Page 202 of 255 Attachment 1 - Emergency Action Level Technical Bases Transfer of plant control and the time period to establish control begins when the last licensed operator leaves the Control Room. Escalation of the emergency classification level would be via IC FG1 or CG1 Reference(s):

1. OP-1203.002 Alternate Shutdown 2. OP-2203.014 Alternate Shutdown

3. NEI 99-01 HS6 to 0CAN031801 Page 203 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a UE EAL: HU7.1 Unusual Event Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which indicate a potential degradation of the level of 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 expected unless further degradation of SAFETY SYSTEMS occurs. Mode Applicability:

All Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the 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 result in potential offsite exposures.

Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an UNUSUAL EVENTNOUE. Reference(s): 1. NEI 99-01 HU7 to 0CAN031801 Page 204 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of an ALERT EAL: HA7.1 Alert Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. Mode Applicability: All Definition(s):

HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station.

HOSTILE ACTION - An act toward ANO or its personnel that includes the use of violent force 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 other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on ANO. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA). OWNER CONTROLLED AREA - For the purposes of classification this is the Security Owner Controlled Area (SOCA). The SOCA is demarcated as a Vehicle Barrier System (VBS) and a detection fence on the outside and a delay fence on the inside of the passive and active barriers. The SOCA is the area inside the SOCA VBS up to the PROTECTED AREA fence line. Access to this area is controlled by the SOCA Personnel Access Control Point.

PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. to 0CAN031801 Page 205 of 255 Attachment 1 - Emergency Action Level Technical Bases Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an ALERT.

Reference(s): 1. NEI 99-01 HA7

to 0CAN031801 Page 206 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a SITE AREA EMERGENCY EAL: HS7.1 Site Area Emergency Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY. Mode Applicability: All Definition(s): HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station.

HOSTILE ACTION - An act toward ANO or its personnel that includes the use of violent force 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 other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on ANO. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA).

OWNER CONTROLLED AREA - For the purposes of classification this is the Security Owner Controlled Area (SOCA). The SOCA is demarcated as a Vehicle Barrier System (VBS) and a detection fence on the outside and a delay fence on the inside of the passive and active barriers. The SOCA is the area inside the SOCA VBS up to the PROTECTED AREA fence line. Access to this area is controlled by the SOCA Personnel Access Control Point.

PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

to 0CAN031801 Page 207 of 255 Attachment 1 - Emergency Action Level Technical Bases PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant. Basis:

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a SITE AREA EMERGENCY.

Reference(s): 1. NEI 99-01 HS7

to 0CAN031801 Page 208 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a GENERAL EMERGENCY EAL: HG7.1 General Emergency Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area. Mode Applicability: All Definition(s):

HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station.

HOSTILE ACTION - An act toward ANO or its personnel that includes the use of violent force 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 other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on ANO. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA). IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

OWNER CONTROLLED AREA - For the purposes of classification this is the Security Owner Controlled Area (SOCA). The SOCA is demarcated as a Vehicle Barrier System (VBS) and a detection fence on the outside and a delay fence on the inside of the passive and active barriers. The SOCA is the area inside the SOCA VBS up to the PROTECTED AREA fence line. Access to this area is controlled by the SOCA Personnel Access Control Point.

PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. to 0CAN031801 Page 209 of 255 Attachment 1 - Emergency Action Level Technical Bases PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a GENERAL EMERGENCY.

Reference(s): 1. NEI 99-01 HG7

to 0CAN031801 Page 210 of 255 Attachment 1 - Emergency Action Level Technical Bases Category S - System Malfunction EAL Group: Hot Conditions (RCS temperature > 200°F); EALs in this category are applicable only in one or more hot operating modes.

Numerous system-related equipment failure events that warrant emergency classification have been 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 Vital AC Power Loss of vital electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite sources for vital 4.16 KV buses. 2. Loss of Vital DC Power Loss of emergency electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of vital plant 125V DC power sources. 3. Loss of Control Room Indications Certain events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification. Losses of indicators are in this subcategory. 4. RCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant rise from these base-line levels (2% - 5% clad failures) is indicative of fuel failures and is covered under the Fission Product Barrier Degradation category. However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits. These fission products will be circulated with the reactor coolant and can be detected by coolant sampling. 5. RCS Leakage The reactor vessel provides a volume for the coolant that covers the reactor core. The reactor pressure vessel and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail. Excessive RCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and containment integrity. to 0CAN031801 Page 211 of 255 Attachment 1 - Emergency Action Level Technical Bases 6. RPS Failure This 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 the RPS to complete a reactor trip comprise a specific set of analyzed events referred to as Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, ATWS is intended to mean any trip failure event that does not achieve reactor shutdown. If RPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could cause a threat to fuel clad, RCS and containment integrity. 7. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification. 8. Containment Failure Failure of containment isolation capability (under conditions in which the containment is not currently challenged) warrants emergency classification. Failure of containment pressure control capability also warrants emergency classification. 9. Hazardous Event Affecting Safety Systems Various natural and technological events that result in degraded plant safety system performance or significant VISIBLE DAMAGE warrant emergency classification under this subcategory.

to 0CAN031801 Page 212 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of Vital AC Power Initiating Condition: Loss of all offsite AC power capability to vital buses for 15 minutes or longer EAL: SU1.1 Unusual Event Loss of all offsite AC power capability, Table 1[2]S-1, to vital 4.16 KV buses A3 [2A3] and A4 [2A4] for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table 1S-1 Unit 1 AC Power Sources Offsite Startup Transformer No. 1 Startup Transformer No. 2 Unit Auxiliary Transformer (from 22 KV switchyard) Onsite Unit Auxiliary Transformer (main generator via main transformer) DG1 DG2 AAC Gen to 0CAN031801 Page 213 of 255 Attachment 1 - Emergency Action Level Technical Bases Table 2S-1 Unit 2 AC Power Sources Offsite Startup Transformer No. 3 Startup Transformer No. 2 Unit Auxiliary Transformer (backfed from main transformer) Onsite Unit Auxiliary Transformer (main generator via main transformer) 2DG1 2DG2 AAC Gen Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s): None

Basis: This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC emergency vital buses. This condition represents 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 vital buses, whether or not the buses are powered from it. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power. Escalation of the emergency classification level would be via IC SA1.

Reference(s): 1. 1SAR Figure 8-1 Station Single Line Diagram 2. OP-1202.007 Degraded Power 3. OP-1202.008 Blackout

4. OP-2104.037 Alternate AC Diesel Generator Operations to 0CAN031801 Page 214 of 255 Attachment 1 - Emergency Action Level Technical Bases 5. 2SAR Figure 8.3-1 Station Single Line Diagram 6. OP-2202.007 Loss of Off-Site Power 7. OP-2202.008 Station Blackout 8. OP-2107.006 Backfeed of Unit Auxiliary Transformer

9. NEI 99-01 SU1 to 0CAN031801 Page 215 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of Vital AC Power Initiating Condition: Loss of all but one AC power source to vital buses for 15 minutes or longer EAL: SA1.1 Alert AC power capability, Table 1[2]S-1, to vital 4.16 KV buses A3 [2A3] and A4 [2A4] reduced to a single power source for 15 min. (Note 1) AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table 1S-1 Unit 1 AC Power Sources Offsite Startup Transformer No. 1 Startup Transformer No. 2 Unit Auxiliary Transformer (from 22 KV switchyard) Onsite Unit Auxiliary Transformer (main generator via main transformer) DG1 DG2 AAC Gen to 0CAN031801 Page 216 of 255 Attachment 1 - Emergency Action Level Technical Bases Table 2S-1 Unit 2 AC Power Sources Offsite Startup Transformer No. 3 Startup Transformer No. 2 Unit Auxiliary Transformer (backfed from main transformer) Onsite Unit Auxiliary Transformer (main generator via main transformer) 2DG1 2DG2 AAC Gen Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the 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 result in potential offsite exposures. Basis: This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC SU1. An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to a emergency vital bus. Some examples of this condition are presented below. to 0CAN031801 Page 217 of 255 Attachment 1 - Emergency Action Level Technical Bases A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator). A loss of all offsite power and loss of all vital emergency power sources (e.g., onsite diesel generators) with a single train of emergency vital buses being back-fed from the unit main generator. A loss of vital emergency power sources (e.g., onsite diesel generators) with a single train of vital emergency buses being back-fed from an offsite power source. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.

Escalation of the emergency classification level would be via IC SS1. This EAL is the hot condition equivalent of the cold condition EAL CU2.1. Reference(s): 1. 1SAR Figure 8-1 Station Single Line Diagram 2. OP-1202.007 Degraded Power

3. OP-1202.008 Blackout

4. OP-2104.037 Alternate AC Diesel Generator Operations

5. 2SAR Figure 8.3-1 Station Single Line Diagram 6. OP-2202.007 Loss of Off-Site Power 7. OP-2202.008 Station Blackout

8. OP-2107.006 Backfeed of Unit Auxiliary Transformer

9. NEI 99-01 SA1 to 0CAN031801 Page 218 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of Vital AC Power Initiating Condition: Loss of all offsite power and all onsite AC power to vital buses for 15 minutes or longer EAL: SS1.1 Site Area Emergency Loss of all offsite and all onsite AC power to vital 4.16 KV buses A3 [2A3] and A4 [2A4] for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown

Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2). Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the 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 result in potential offsite exposures. Basis:

Although the AAC may be considered available, it will not prevent declaration of this EAL unless it is powering a vital bus within the 15 minute time period of the EAL. This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. Mitigative strategies using non-safety related power sources (FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or emergency operating procedures, or beyond design basis to 0CAN031801 Page 219 of 255 Attachment 1 - Emergency Action Level Technical Bases accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures 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 AG1, FG1 or SG1. This EAL is the hot condition equivalent of the cold condition EAL CA2.1. Reference(s): 1. 1SAR Figure 8-1 Station Single Line Diagram 2. OP-1202.007 Degraded Power

3. OP-1202.008 Blackout

4. OP-2104.037 Alternate AC Diesel Generator Operations 5. 2SAR Figure 8.3-1 Station Single Line Diagram 6. OP-2202.007 Loss of Off-Site Power

7. OP-2202.008 Station Blackout

8. OP-2107.006 Backfeed of Unit Auxiliary Transformer

9. NEI 99-01 SS1 to 0CAN031801 Page 220 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of Vital AC Power Initiating Condition: Prolonged loss of all offsite and all onsite AC power to vital buses EAL: SG1.1 General Emergency Loss of all offsite and all onsite AC power to vital 4.16 KV buses A3 [2A3] and A4 [2A4] AND EITHER: Restoration of at least one vital 4.16 KV bus in < 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely (Note 1) CETs > 1200°F Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the 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 result in potential offsite exposures. Basis: This IC addresses a prolonged loss of all power sources to AC emergency vital buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. Mitigative strategies using non-safety related power sources (FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or to 0CAN031801 Page 221 of 255 Attachment 1 - Emergency Action Level Technical Bases emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. A prolonged loss of these buses will lead to a loss of one or more fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these conditions.

The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1. 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 projected that power cannot be restored to at least one AC emergency vital bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased greater likelihood of challenges to multiple fission product barriers. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is the site-specific SBO coping analysis time (ref. 4, 5). The estimate for restoring at least one emergency vital bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.

The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.

Reference(s): 1. OP-1202.005 Inadequate Core Cooling 2. OP-2202.009 Functional Recovery

3. OP-2202.011 Lower Mode Functional Recovery

4. Unit 1 Calculation 85-E-0072-02 Time from Loss of All AC Power to Loss of Subcooling 5. Unit 2 Calculation 85-E-0072-01 Time from Loss of All AC Power to Loss of Subcooling 6. 1SAR Figure 8-1 Station Single Line Diagram

7. OP-1202.007 Degraded Power

8. OP-1202.008 Blackout

9. OP-2104.037 Alternate AC Diesel Generator Operations 10. 2SAR Figure 8.3-1 Station Single Line Diagram 11. OP-2202.007 Loss of Off-Site Power

12. OP-2202.008 Station Blackout

13. OP-2107.006 Backfeed of Unit Auxiliary Transformer

14. NEI 99-01 SG1 to 0CAN031801 Page 222 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of Vital AC Power Initiating Condition: Loss of all vital AC and vital DC power sources for 15 minutes or longer EAL: SG1.2 General Emergency Loss of all offsite and all onsite AC power to vital 4.16 KV buses A3 [2A3] and A4 [2A4] for 15 min. (Note 1) AND Indicated voltage is < 105 VDC on D01 [2D01] and D02 [2D02] vital 125 VDC buses for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the 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 result in potential offsite exposures. Basis: Unit 1 batteries D06 and D07 and Unit 2 batteries 2D11 and 2D12 contain 58 cells each with a minimum cell voltage of 1.81 V or 105 VDC (ref. 9, 10). This IC addresses a concurrent and prolonged loss of both vital AC and Vital DC power. A loss of all vital AC power compromises the performance of all SAFETY SYSTEMS requiring electric to 0CAN031801 Page 223 of 255 Attachment 1 - Emergency Action Level Technical Bases power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. Mitigative strategies using non-safety related power sources (FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. A loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both vital AC and vital DC power will lead to multiple challenges to fission 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 are met. Reference(s): 1. 1SAR Figure 8-1 Station Single Line Diagram 2. OP-1202.007 Degraded Power 3. OP-1202.008 Blackout 4. OP-2104.037 Alternate AC Diesel Generator Operations

5. 2SAR Figure 8.3-1 Station Single Line Diagram

6. OP-2202.007 Loss of Off-Site Power

7. OP-2202.008 Station Blackout 8. OP-2107.006 Backfeed of Unit Auxiliary Transformer 9. 1SAR 8.3.2.1.1 Batteries

10. 2SAR 8.3.2.1.1 Batteries

11. OP-1203.036 Loss of 125V DC

12. OP-2203.037 Loss of 125V DC

13. 2SAR Figure 8.3-6 Low Voltage Safety System Power Supplies 14. NEI 99-01 SG8 to 0CAN031801 Page 224 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 2 - Loss of Vital DC Power Initiating Condition: Loss of all vital DC power for 15 minutes or longer EAL: SS2.1 Site Area Emergency Indicated voltage is < 105 VDC on D01 [2D01] and D02 [2D02] vital 125 VDC buses for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the 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 result in potential offsite exposures. Basis: Unit 1 batteries D06 and D07 and Unit 2 batteries 2D11 and 2D12 contain 58 cells each with a minimum cell voltage of 1.81 V or 105 VDC (ref. 2, 3). This IC addresses a loss of vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure 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 AG1, FG1 or SG1SG8. to 0CAN031801 Page 225 of 255 Attachment 1 - Emergency Action Level Technical Bases This EAL is the hot condition equivalent of the cold condition EAL CU4.1. Reference(s): 1. 1SAR Figure 8-1 Station Single Line Diagram 2. 1SAR 8.3.2.1.1 Batteries 3. 2SAR 8.3.2.1.1 Batteries 4. OP-1203.036 Loss of 125V DC

5. OP-2203.037 Loss of 125V DC

6. 2SAR Figure 8.3-6 Low Voltage Safety System Power Supplies

7. NEI 99-01 SS8 to 0CAN031801 Page 226 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer EAL: SU3.1 Unusual Event An UNPLANNED event results in the inability to monitor one or more Table 1[2]S-2 parameters from within the Control Room for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table 1[2]S-2 Safety System Parameters Reactor power RCS level RCS pressure CET temperature Level in at least one S/G EFW flow to at least one S/G Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the 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 result in potential offsite exposures. to 0CAN031801 Page 227 of 255 Attachment 1 - Emergency Action Level Technical Bases UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis:

This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.

As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling [PWR] / RPV level [BWR] and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level [PWR] / RPV water level [BWR] cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level would be via IC EAL SA3.1SA2. Reference(s): 1. 1SAR 7.5 Safety-Related Display Instrumentation 2. 2SAR 7.5 Safety-Related Display Instrumentation 3. NEI 99-01 SU2 to 0CAN031801 Page 228 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress EAL: SA3.1 Alert An UNPLANNED event results in the inability to monitor one or more Table 1[2]S-2 parameters from within the Control Room for 15 min. (Note 1) AND Any significant transient is in progress, Table 1[2]S-3 Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table 1[2]S-2 Safety System Parameters Reactor power RCS level RCS pressure CET temperature Level in at least one S/G EFW flow to at least one S/G Table 1[2]S-3 Significant Transients Reactor trip Runback > 25% thermal power Electrical load rejection > 25% electrical load Safety injection actuation Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown to 0CAN031801 Page 229 of 255 Attachment 1 - Emergency Action Level Technical Bases Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the 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 result in potential offsite exposures.

UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis: This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant.

As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling [PWR] / RPV level [BWR] and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level [PWR] / RPV water level [BWR] cannot be determined from the to 0CAN031801 Page 230 of 255 Attachment 1 - Emergency Action Level Technical Bases indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level would be via ICs FS1 or IC AS1. Reference(s): 1. 1SAR 7.1.3 Engineered Safeguards Actuation System 2. 2SAR 7.3 Engineered Safety Features Systems 3. NEI 99-01 SA2 to 0CAN031801 Page 231 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 4 - RCS Activity Initiating Condition: RCS activity greater than Technical Specification allowable limits EAL: SU4.1 Unusual Event Failed Fuel Iodine radiation monitor RI-1237S [2RITS-4806B] > 9.0 E5 cpm Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s): None Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.

Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category A ICs. Unit 1 RE-1237S, Failed Fuel Monitor, is in the letdown system to monitor the letdown line for evidence of fuel damage. Unit 2 specific activity monitor 2RITS-4806B monitors the Letdown fluid for the presence of Iodine-131. A monitor reading corresponding to the instantaneous dose equivalent I-131 value of 60 uCi/gm is determined by multiplying by 30 the monitor reading listed in the table in OP-1203.019[OP-2203.020] that represents a projected 2.0 uCi/gm I-131 RCS activity in order to correlate to a Tech Spec instantaneous limit of 60 uCi/gm dose equivalent I-131 for the EAL (ref. 2, 5). This yields values of 3.1E6 cpm for Unit 1 and 3.9E6 cpm for Unit 2. The top of scale of the monitor is 1E6. The EAL value is set at 9.0 E5 cpm for both units which is 90% of the top of the scale. to 0CAN031801 Page 232 of 255 Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. 1SAR Table 11-7 2. OP-1203-019 High Activity in Reactor Coolant

3. Unit 1 Technical Specifications LCO 3.4.12 RCS Specific Activity 4. 2SAR 9.3.5 Failed Fuel Detection System

5. OP-2203.020 High Activity in RCS 6. OP- 2203.012L ANNUNCIATOR 2K12 CORRECTIVE ACTION, A-1 7. Unit 2 Technical Specifications LCO 3.4.8 Reactor Coolant System Specific Activity 8. NEI 99-01 SU3 to 0CAN031801 Page 233 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 4 - RCS Activity Initiating Condition: RCS activity greater than Technical Specification allowable limits EAL: SU4.2 Unusual Event RCS sample activity > 1.0 µCi/gm dose equivalent I-131 for > 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> (Note 1) OR RCS sample activity > 60 µCi/gm dose equivalent I-131 OR RCS sample activity > 2200[3100] µCi/gm dose equivalent Xe-133 for > 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s): None Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.

Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category A ICs. Reference(s): 1. Unit 1 Technical Specifications LCO 3.4.12 RCS Specific Activity 2. Unit 2 Technical Specifications LCO 3.4.8 Reactor Coolant System Specific Activity 3. NEI 99-01 SU3 to 0CAN031801 Page 234 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 5 - RCS Leakage Initiating Condition: RCS leakage for 15 minutes or longer EAL: SU5.1 Unusual Event RCS unidentified or pressure boundary leakage > 10 gpm for 15 min. (Note 1) OR RCS identified leakage > 25 gpm for 15 min. (Note 1) OR Reactor coolant leakage to a location outside containment > 25 gpm for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s): UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally. Basis: Failure to isolate the leak (from the Control Room or locally) within 15 minutes, or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification. Steam generator tube leakage is identified RCS leakage. This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant.

The first and second EAL conditions EAL #1 and EAL #2 are focused on a loss of mass from the RCS due to "unidentified leakage", "pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the plant Technical Specifications). The third condition EAL #3 addresses an RCS mass loss caused by an UNISOLABLE leak through an interfacing to 0CAN031801 Page 235 of 255 Attachment 1 - Emergency Action Level Technical Bases system. These conditions EALs thus apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage in a PWR) or a location outside of containment. The leak rate values for each condition EAL were selected because they are usually observable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). The first condition EAL #1 uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage. The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. For PWRs, aAn emergency classification would be required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated). For BWRs, a stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL. The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible. Escalation of the emergency classification level would be via ICs of Recognition Category A or F.

Reference(s):

1. Unit 1 and Unit 2 Technical Specifications Section 1.1 Definitions 2. NEI 99-01 SU4 to 0CAN031801 Page 236 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor EAL: SU6.1 Unusual Event An automatic trip did not shut down the reactor as indicated by reactor power > 5% after any RPS setpoint is exceeded AND A subsequent automatic trip or manual trip action taken at the reactor control console (C03 [2C03/2C14]) (manual reactor trip pushbuttons or DROPS[DSS]) is successful in shutting down the reactor as indicated by reactor power 5% (Note 8) Note 8: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Mode Applicability: 1 - Power Operation Definition(s):

IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor (trip [PWR] / scram [BWR]) that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic (trip [PWR] / scram [BWR]) is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.

In the event that the operator identifies a reactor trip is IMMINENT and initiates a successful manual reactor trip before the automatic trip setpoint is reached, no declaration is required. The successful manual trip of the reactor before it reaches its automatic trip setpoint or reactor trip signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss.

Following the failure onf an automatic reactor (trip[PWR] / scram [BWR]), operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor (trip[PWR] / scram [BWR])). If these manual actions are successful in shutting to 0CAN031801 Page 237 of 255 Attachment 1 - Emergency Action Level Technical Bases down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. If an initial manual reactor (trip [PWR] / scram [BWR]) is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor (trip[PWR] / scram [BWR])) using a different switch). Depending upon several factors, the initial or subsequent effort to manually (trip [PWR] / scram [BWR]) the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor (trip [PWR] / scram [BWR]) signal. If a subsequent manual or automatic (trip [PWR] / scram [BWR]) is successful in shutting down the reactor, core heat generation will quickly fall to 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, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor (trip[PWR] / scram [BWR])). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. [BWR] The 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, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA5SA6. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA5 SA6 or FA1, an Unusual Event declaration is appropriate for this event. A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria. Should a reactor (trip [PWR] / scram [BWR]) signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied. If the signal generated as a result of plant work causes a plant transient that results in a condition that should have included an automatic reactor trip [PWR] / scram [BWR]) and the RPS fails to automatically shutdown the reactor, then this IC and associated the EALs are applicable, and should be evaluated. If the signal generated as a result of plant work does not cause a plant transient and the (trip [PWR] / scram [BWR]) failure is determined through other means (e.g., assessment of test results), then this IC and associated the EALs are not applicable and no classification is warranted. to 0CAN031801 Page 238 of 255 Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. Unit 1 Technical Specifications Table 3.3.1-1 Reactor Protection System Instrumentation 2. Unit 2 Technical Specifications Table 3.3-1 Reactor Protective Instrumentation

3. Unit 1 and Unit 2 Technical Specifications Table 1.1-1 Modes

4. OP-1202.001 Reactor Trip

5. OP-2202.001 Standard Post Trip Actions 6. NEI 99-01 SU5 to 0CAN031801 Page 239 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor EAL: SU6.2 Unusual Event A manual trip did not shut down the reactor as indicated by reactor power > 5% after any manual trip action was initiated AND A subsequent automatic trip or manual trip action taken at the reactor control console (C03 [2C03/2C14]) (manual reactor trip pushbuttons or DROPS[DSS]) is successful in shutting down the reactor as indicated by reactor power 5% (Note 8) Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Mode Applicability: 1 - Power Operation Definition(s):

None Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor (trip [PWR] / scram [BWR]) that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic (trip [PWR] / scram [BWR]) is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.

This EAL addresses a failure of a manually initiated trip in the absence of having exceeded an automatic RPS trip setpoint and a subsequent automatic or manual trip is successful in shutting down the reactor. Following the failure on an automatic reactor (trip [PWR] / scram [BWR]), operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor (trip [PWR] / scram [BWR])). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.

to 0CAN031801 Page 240 of 255 Attachment 1 - Emergency Action Level Technical Bases If an initial manual reactor (trip [PWR] / scram [BWR]) is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor (trip [PWR] / scram [BWR])) using a different switch). Depending upon several factors, the initial or subsequent effort to manually (trip [PWR] / scram [BWR]) the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor (trip [PWR] / scram [BWR]) signal. If a subsequent manual or automatic (trip [PWR] / scram [BWR]) is successful in shutting down the reactor, core heat generation will quickly fall to 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, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor (trip [PWR] / scram [BWR])). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles."

Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. [BWR] The 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, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA5SA6. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA5 SA6 or FA1, an Unusual Event declaration is appropriate for this event.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria. Should a reactor (trip [PWR] / scram [BWR]) signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied. If the signal generated as a result of plant work causes a plant transient that results in a condition that should have included an automatic reactor (trip [PWR] / scram [BWR]) and the RPS fails to automatically shutdown the reactor, then this IC and associated the EALs are applicable, and should be evaluated. If the signal generated as a result of plant work does not cause a plant transient and the (trip [PWR] / scram [BWR]) failure is determined through other means (e.g., assessment of test results), then this IC and associated the EALs are not applicable and no classification is warranted. to 0CAN031801 Page 241 of 255 Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. Unit 1 Technical Specifications Table 3.3.1-1 Reactor Protection System Instrumentation 2. Unit 2 Technical Specifications Table 3.3-1 Reactor Protective Instrumentation

3. Unit 1 and Unit 2 Technical Specifications Table 1.1-1 Modes

4. OP-1202.001 Reactor Trip

5. OP-2202.001 Standard Post Trip Actions 6. NEI 99-01 SU5 to 0CAN031801 Page 242 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor EAL: SA6.1 Alert An automatic or manual trip fails to shut down the reactor as indicated by reactor power > 5% AND Manual trip actions taken at the reactor control console (C03 [2C03/2C14]) (manual reactor trip pushbuttons or DROPS[DSS]) are not successful in shutting down the reactor as indicated by reactor power > 5% (Note 8) Note 8: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Mode Applicability: 1 - Power Operation Definition(s): None Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor (trip [PWR] / scram [BWR]) that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS.

A manual action at the reactor control console is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor (trip [PWR] / scram [BWR])). This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers). Actions taken at back panels or other locations within to 0CAN031801 Page 243 of 255 Attachment 1 - Emergency Action Level Technical Bases the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles." Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. [BWR] The 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, other concurrent plant conditions, etc. If the failure to shut down the reactor is prolonged enough to cause a challenge to the core cooling [PWR] / RPV water level [BWR] or RCS RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS65. Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent the plant conditions needed to meet either IC SS65 or FS1, an Alert declaration is appropriate for this event.

It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration. A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Reference(s): 1. Unit 1 Technical Specifications Table 3.3.1-1 Reactor Protection System Instrumentation 2. Unit 2 Technical Specifications Table 3.3-1 Reactor Protective Instrumentation

3. Unit 1 and Unit 2 Technical Specifications Table 1.1-1 Modes

4. OP-1202.001 Reactor Trip 5. OP-2202.001 Standard Post Trip Actions 6. NEI 99-01 SA5 to 0CAN031801 Page 244 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Inability to shut down the reactor causing a challenge to core cooling or RCS heat removal EAL: SS6.1 Site Area Emergency An automatic or manual trip fails to shut down the reactor as indicated by reactor power > 5% AND All actions to shut down the reactor are not successful as indicated by reactor power > 5% AND EITHER: CETs >1200°F RCS heat removal cannot be established using steam generators and HPI [Once Through] cooling initiated. Mode Applicability: 1 - Power Operation

Definition(s): None Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor (trip [PWR] / scram [BWR]) that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.

In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shut down the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria. to 0CAN031801 Page 245 of 255 Attachment 1 - Emergency Action Level Technical Bases Escalation of the emergency classification level would be via IC AG1 or FG1. Reference(s): 1. Unit 1 Technical Specifications Table 3.3.1-1 Reactor Protection System Instrumentation 2. Unit 2 Technical Specifications Table 3.3-1 Reactor Protective Instrumentation 3. Unit 1 and Unit 2 Technical Specifications Table 1.1-1 Modes 4. OP-1202.001 Reactor Trip

5. OP-2202.001 Standard Post Trip Actions

6. OP-1202.004 Overheating

7. OP-2202.006 Loss of Feedwater 8. OP-1202.013 Figure 1, Saturation and Adequate SCM 9. Calculation 90-E-0116-07 Unit 1 EOP Setpoint Document, Setpoint B.19

10. OP-2202.009 Functional Recovery

11. Calculation 90-E-0116-01 Unit 2 EOP Setpoint Document

12. NEI 99-01 SS5 to 0CAN031801 Page 246 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 7 - Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL: SU7.1 Unusual Event Loss of all Table 1[2]S-4 onsite communication methods OR Loss of all Table 1[2]S-4 State and local agency communication methods OR Loss of all Table 1[2]S-4 NRC communication methods Table 1[2]S-4 Communication Methods System Onsite State / Local NRC Station radio system X ANO plant phone system X Gaitronics X Telephone Systems: Commercial Microwave Satellite VOIP X X INFORM Notification System X Emergency Notification System (ENS) X Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown to 0CAN031801 Page 247 of 255 Attachment 1 - Emergency Action Level Technical Bases Definition(s):

None Basis: This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs State and local agencies and the NRC. This IC should be assessed only when extraordinary means are being utilized to make communications 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 to offsite locations, etc.).

EAL #1The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations. EAL #2The second EAL condition addresses a total loss of the communications methods used to notify all OROs State and local agencies of an emergency declaration. The OROs State and local agencies referred to here are the Arkansas Department of Health, Arkansas Department of Emergency Management, Pope, Yell, Johnson, and Logan County agencies.(see Developer Notes) EAL #3The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. This EAL is the hot condition equivalent of the cold condition EAL CU5.1. Reference(s): 1. OP-1903.062 Communications System Operating Procedure 2. NEI 99-01 SU6 to 0CAN031801 Page 248 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 8 - Containment Failure Initiating Condition: Failure to isolate containment or loss of containment pressure control EAL: SU8.1 Unusual Event Any penetration is not closed within 15 min. of an ESAS [CIAS] actuation signal OR Containment pressure > 44.7 psia [23.3 psia] with < one full train of containment heat removal systems (Note 9) operating per design for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 9: One full train of containment heat removal systems consists of one train of RB [Containment] Spray and one train of RB [Containment] Cooling System. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown

Definition(s): None Basis:

A penetration is closed for this EAL if either side of the penetration has a closed valve or a check valve is intact (for penetrations that only have one automatic valve and a check valve). This IC EAL addresses a failure of one or more containment penetrations to automatically isolate (close) when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems. Absent challenges to another fission product barrier, either condition represents potential degradation of the level of safety of the plant.

For EAL #1the first condition, the containment isolation signal must be generated as the result on an off-normal/accident condition (e.g., a safety injection or high containment pressure); a failure resulting from testing or maintenance does not warrant classification. The determination of containment and penetration status - isolated or not isolated - should be made in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute to 0CAN031801 Page 249 of 255 Attachment 1 - Emergency Action Level Technical Bases criterion is included to allow operators time to manually isolate the required penetrations, if possible. EAL #2The second condition addresses a condition where containment pressure is greater than the setpoint at 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-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. The inability to start the required equipment 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 a concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers. Reference(s): 1. OP-1202.010 ESAS 2. 1SAR 6.2 Reactor Building Spray System 3. 1SAR 6.3 Reactor Building Cooling System 4. OP-2202.003 Loss of Coolant Accident

5. OP-2202.010 Standard Attachments, Attachment 22

6. 2SAR 6.2.2 Containment Heat Removal Systems

7. 2SAR 7.3.1.1.11.2 Containment Spray System 8. NEI 99-01 SU7 to 0CAN031801 Page 250 of 255 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 9 - Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode EAL: SA9.1 Alert The occurrence of any Table 1[2]S-5 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER: Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 10, 11) Note 10: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted. Note 11: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted. Table 1[2]S-5 Hazardous Events Seismic event (earthquake) Internal or external FLOODING event High winds or tornado strike FIRE EXPLOSION Other events with similar hazard characteristics as determined by the Shift Manager to 0CAN031801 Page 251 of 255 Attachment 1 - Emergency Action Level Technical Bases Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s): EXPLOSION - A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines 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 drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred 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 installed equipment 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/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the 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 result in potential offsite exposures.

VISIBLE DAMAGE - Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train. Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. to 0CAN031801 Page 252 of 255 Attachment 1 - Emergency Action Level Technical Bases Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events. Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make a determination of VISIBLE DAMAGE based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure 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 product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant. EAL 1.b.1 addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. EAL 1.b.2 addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. Escalation of the emergency classification level would be via IC FS1 or AS1. This EAL is the hot condition equivalent of the cold condition EAL CA6.1. Reference(s): 1. EP FAQ 2016-002 2. NEI 99-01 SA9 to 0CAN031801 Page 253 of 255 Attachment 2 - Safe Operation & Shutdown Areas Tables 1[2]A-3 & 1[2]H-2 Bases Background NEI 99-01 Revision 6 ICs AA3 and HA5 prescribe declaration of an Alert based on impeded access 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 Developers Notes 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/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Do not include rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations). In addition, the list should specify the plant mode(s) during which entry would be required for each room or area. The list should not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections). Further, as specified in IC HA5:

The list need not include the Control Room if adequate engineered safety/design features are 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 air intakes at different and separate locations, inner and outer atmospheric boundaries, or the capability to acquire and maintain positive pressure within the Control Room envelope.

to 0CAN031801 Page 254 of 255 Attachment 2 - Safe Operation & Shutdown Areas Tables 1[2]A-3 & 1[2]H-2 Bases ANO Table 1[2]A-3 and 1[2]H-2 Bases A review of station operating procedures identified the following mode dependent in-plant actions and associated areas that are required for normal plant operation, cooldown or shutdown:

Unit 1 AREA MODES PURPOSE REFERENCE A-4 Switchgear Room 3, 4 Core flood tank valves, decay heat removal (DHR) OP-1102.010 OP-1104.004 Upper North Electrical Penetration Room 3, 4 DHR alignment OP-1104.004 Lower South Electrical Equipment Room 3, 4 DHR alignment OP-1104.004 Unit 2 AREA MODES PURPOSE REFERENCE Aux Building 317' Emergency Core Cooling Rooms 3, 4 Shutdown Cooling (SDC) venting and alignment OP-2104.004 Aux Building 317' Tendon Gallery Access 3, 4 SDC alignment OP-2104.004 Aux Building 335' Charging Pumps / Motor Control Center (MCC) 2B-52 3, 4 Charging low pressure operation, T-Hot injection valves, and SDC alignment OP-2102.010 OP-2104.004 Auxiliary Building 354' MCC 2B-62 Area 3, 4 SDC alignment and T-Hot injection valves at MCC 2B-62 OP-2102.010 OP-2104.004 Emergency Diesel Generator Corridor 3, 4 Close Safety Injection Tank (SIT) valves and SDC / Low Temperature Overpressure (LTOP) valve alignment at MCC 2B-51 OP-2102.010 Lower South Piping Penetration Room 3, 4 SDC alignment OP-2104.004 Aux Building 386' Containment Hatch 3, 4 Close SIT valves at MCC 2B-61 OP-2102.010 Mode 3 is included above for DHR- and SDC-related activities because the procedures begin alignment in Mode 3; however, these actions could be delayed until Mode 4, if necessary. In order to ensure adequate guidance to emergency response personnel, the above areas are added to the EAL in order to provide prompt operator guidance for EAL declaration. to 0CAN031801 Page 255 of 255 Attachment 2 - Safe Operation & Shutdown Areas Tables 1[2]A-3 & 1[2]H-2 Bases Both ANO-1 and ANO-2 Control Room ventilation systems have adequate engineered safety/design features in place to preclude a Control Room evacuation due to the release of a hazardous gas. Therefore the Control Room is not included in this assessment or in Tables 1[2]H-2. Table 1[2]A-3 & 1[2]H-2 Results Table 1[2]A-3 & 1[2]H-2 Safe Operation & Shutdown Rooms/Areas Unit 1 Room/Area Mode Applicability A-4 Switchgear Room 3, 4 Upper North Electrical Penetration Room 3, 4 Lower South Electrical Equipment Room 3, 4 Unit 2 Room/Area Mode Applicability Aux Building 317' Emergency Core Cooling Rooms 3, 4 Aux Building 317' Tendon Gallery Access 3, 4 Aux Building 335' Charging Pumps / MCC 2B-52 3, 4 Auxiliary Building 354' MCC 2B-62 Area 3, 4 Emergency Diesel Generator Corridor 3, 4 Lower South Piping Penetration Room 3, 4 Aux Building 386' Containment Hatch 3, 4

Enclosure 3 to 0CAN031801 Proposed EAL Technical Basis Document (Clean) to 0CAN031801 Page 1 of 240 Table of Contents Section Page

1.0 INTRODUCTION

.................................................................................................................2 2.0 DISCUSSION ......................................................................................................................2 2.1 Background ................................................................................................................2 2.2 Fission Product Barriers .............................................................................................3 2.3 Fission Product Barrier Classification Criteria ............................................................3 2.4 EAL Organization .......................................................................................................4 2.5 Technical Basis Information .......................................................................................5 2.6 Operations Mode Applicability ....................................................................................7 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS ..........................................8 3.1 General Considerations .............................................................................................8 3.2 Classification Methodology ......................................................................................10

4.0 REFERENCES

..................................................................................................................13 4.1 Developmental .........................................................................................................13 4.2 Implementing............................................................................................................13 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS ...........................................................13 5.1 Definitions (ref. 4.1.1 except as noted) ....................................................................13 5.2 Abbreviations/Acronyms ..........................................................................................18 6.0 ANO-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE ..................................................21

7.0 ATTACHMENTS ...............................................................................................................24 7.1 Attachment 1, Emergency Action Level Technical Bases ........................................24 Category A - Abnormal Rad Levels / Rad Effluents ................................................25 Category C - Cold Shutdown / Refueling System Malfunction ................................65 Category E - Independent Spent Fuel Storage Installation (ISFSI) .......................104 Category F - Fission Product Barrier Degradation ................................................107 Table 1[2]F-1, Fission Product Barrier Threshold Matrix & Bases ...114 Category H - Hazards and Other Conditions Affecting Plant Safety .....................158 Category S - System Malfunction ..........................................................................196 7.2 Attachment 2, Safe Operation & Shutdown Areas Tables 1[2]A-3 & 1[2]H-2 Bases .....................................................238 to 0CAN031801 Page 2 of 240

1.0 INTRODUCTION

This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for Arkansas Nuclear One (ANO). It should be used to facilitate review of the ANO EALs and provide historical documentation for future reference. Decision-makers responsible for implementation of 1903.010, Emergency Action Level Classification, may use this document as a technical reference in support of EAL interpretation.

This information may assist the Emergency Director in making classifications, particularly those involving judgment or multiple events. The basis information may also be useful in training and for explaining event classifications to off-site officials.

The expectation is that emergency classifications are to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes or less in all cases of conditions present. Use of this document for assistance is not intended to delay the emergency classification.

Because the information in a basis document can affect emergency classification decision-making (e.g., the Emergency Director refers to it during an event), the NRC staff expects that changes to the basis document will be evaluated in accordance with the provisions of 10 CFR 50.54(q). 2.0 DISCUSSION 2.1 Background EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the ANO Emergency Plan.

In 1992, the NRC endorsed NUMARC/NESP-007 "Methodology for Development of Emergency Action Levels" as an alternative to NUREG-0654 EAL guidance. NEI 99-01 (NUMARC/NESP-007) Revisions 4 and 5 were subsequently issued for industry implementation. Enhancements over earlier revisions included:

Consolidating the system malfunction initiating conditions and example emergency action levels which address conditions that may be postulated to occur during plant shutdown conditions. Initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent 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 to numerous implementation issues including the NRC EAL Frequently Asked Questions (FAQs). Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors," November 2012 (ref. 4.1.1), ANO conducted an EAL implementation upgrade project that produced the EALs discussed herein. to 0CAN031801 Page 3 of 240 2.2 Fission Product Barriers Fission product barrier thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment. This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment.

Many of the EALs derived from the NEI methodology are fission product barrier threshold based. That is, the conditions that define the EALs are based upon thresholds that represent the 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 a greater probability of barrier loss and reduced certainty of maintaining the barrier. The primary fission product barriers are:

A. Fuel Clad Barrier (FCB): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets. B. Reactor Coolant System Barrier (RCB): The Reactor Coolant System Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves. C. Containment Barrier (CNB): The Containment Barrier includes the Reactor Building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the Reactor Building up to and including the outermost secondary side isolation valve.

Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (ECL) from an Alert to a Site Area Emergency or a General Emergency. 2.3 Fission Product Barrier Classification Criteria

The following criteria are the bases for event classification related to fission product barrier loss or potential loss: Alert: Any loss or any potential loss of either Fuel Clad or RCS Barrier Site Area Emergency: Loss or potential loss of any two barriers General Emergency: Loss of any two barriers and loss or potential loss of the third barrier to 0CAN031801 Page 4 of 240 2.4 EAL Organization The ANO EAL scheme includes the following features:

Division of the EAL set into three broad groups: o EALs applicable under any plant operating modes - This group would be reviewed by the EAL-user any time emergency classification is considered. o EALs applicable only under hot operating modes - This group would only be reviewed 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 be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling or Defueled mode. The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition. This approach significantly minimizes the total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency. Within each group, assignment of EALs to categories and subcategories: Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. The ANO EAL categories are aligned to and represent the NEI 99-01, "Recognition Categories." Subcategories are used in the ANO scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The ANO EAL categories and subcategories are listed below.

The primary tool for determining the emergency classification level is the EAL Classification Matrix. The user of the EAL Classification Matrix may (but is not required to) consult the EAL technical bases in order to obtain additional information concerning the EALs under classification consideration. The user should consult Section 3.0 and Attachment 1 of this document for such information.

to 0CAN031801 Page 5 of 240 EAL Groups, Categories and Subcategories EAL Group/Category EAL Subcategory Any Operating Mode: A - Abnormal Rad Levels / Rad Effluent 1 - Radiological Effluent 2 - Irradiated Fuel Event 3 - Area Radiation Levels H - Hazards and Other Conditions Affecting Plant Safety 1 - Security 2 - Seismic Event 3 - Natural or Technological Hazard 4 - Fire 5 - Hazardous Gas 6 - Control Room Evacuation 7 - Emergency Director Judgment E - Independent Spent Fuel Storage Installation (ISFSI) 1 - Confinement Boundary Hot Conditions: S - System Malfunction 1 - Loss of Essential AC Power 2 - Loss of Vital DC Power 3 - Loss of Control Room Indications 4 - RCS Activity 5 - RCS Leakage 6 - RPS Failure 7 - Loss of Communications 8 - Containment Failure 9 - Hazardous Event Affecting Safety Systems F - Fission Product Barrier Degradation None Cold Conditions: C - Cold Shutdown / Refueling System Malfunction 1 - RCS Level 2 - Loss of Essential AC Power 3 - RCS Temperature 4 - Loss of Vital DC Power 5 - Loss of Communications 6 - Hazardous Event Affecting Safety Systems 2.5 Technical Bases Information EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (Any, Hot, Cold), EAL category (A, C, E, F, H and S) and EAL subcategory. A summary explanation of each category and subcategory is given at the beginning of the technical bases discussions of the EALs included in the category. For each EAL, the following information is provided:

to 0CAN031801 Page 6 of 240 Category Letter & Title Subcategory Number & Title Initiating 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 the emergency classification to onsite and offsite personnel. Four characters define each EAL identifier: 1. First character (letter): Corresponds to the EAL category as described above (A, C, E, F, H or S) 2. Second character (letter): The emergency classification (G, S, A or U) G = General Emergency S = Site Area Emergency A = Alert U = Unusual Event 3. Third character (number): Subcategory number within the given category. Subcategories are sequentially numbered beginning with the number one (1). If a category 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 EAL subcategory. 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 Matrix. If an ANO Unit 2 EAL threshold value differs from Unit 1, the Unit 2 threshold is enclosed in brackets. For example, in the EAL threshold "RVLMS Levels 1 through 8 indicate DRY [RVLMS Levels 1 through 5 indicate DRY]", "RVLMS Levels 1 through 5 indicate DRY" apply only to Unit 2. Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling, DEF - Defueled, or Any. (See Section 2.6 for operating mode definitions).

to 0CAN031801 Page 7 of 240 Definitions: If the EAL wording contains a defined term, the definition of the term is included in this section. These definitions can also be found in Section 5.1. Basis: An EAL basis section that provides ANO-relevant information concerning the EAL as well as a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6. Reference(s): Source documentation from which the EAL is derived. 2.6 Operating Mode Applicability Unit 1 (ref. 4.1.6):

1 Power Operation Keff 0.99, reactor power > 5% 2 Startup Keff 0.99, reactor power 5% 3 Hot Standby Keff < 0.99, reactor coolant temperature 280°F 4 Hot Shutdown Keff < 0.99, reactor coolant temperature 280°F > Tavg > 200°F and all reactor vessel head closure bolts fully tensioned 5 Cold Shutdown Keff < 0.99, reactor coolant temperature 200°F and all reactor vessel head closure bolts fully tensioned 6 Refueling One or more reactor vessel head closure bolts less than fully tensioned DEF Defueled All fuel assemblies have been removed from Containment and placed in the spent fuel pool.

to 0CAN031801 Page 8 of 240 Unit 2 (ref. 4.1.6): 1 Power Operation Keff 0.99, reactor power > 5%, average coolant temperature 300°F 2 Startup Keff 0.99, reactor power 5%, average coolant temperature 300°F 3 Hot Standby Keff < 0.99, average coolant temperature 300°F 4 Hot Shutdown Keff < 0.99, average coolant temperature 300°F > Tavg > 200°F 5 Cold Shutdown Keff < 0.99, average coolant temperature 200°F 6 Refueling Keff 0.95, average coolant temperature 140°F, reactor vessel head unbolted or removed, and fuel in the vessel. DEF Defueled All fuel assemblies have been removed from Containment and placed in the spent fuel pool.

The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action being initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred.

3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 3.1 General Considerations When making an emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes, and the informing basis information. In the Recognition Category F matrices, EALs are based on loss or potential loss of Fission Product Barrier Thresholds.

EAL matrices shall be read from left to right, from General Emergency to Unusual Event, and top to bottom. Declaration decisions shall be independently verified before declaration is made except when gaining this verification would exceed the 15 minute declaration requirement. Place keeping shall be used on all EAL matrices.

to 0CAN031801 Page 9 of 240 3.1.1 Classification Timeliness NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (ref. 4.1.8). 3.1.2 Valid Indications

All emergency classification assessments shall be based upon valid indications, reports or conditions. A valid indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy. For example, verification could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel.

An indication, report, or condition is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

3.1.3 Imminent Conditions For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.

3.1.4 Planned vs. Unplanned Events A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that: 1) the activity proceeds as planned, and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 CFR 50.72 (ref. 4.1.4).

3.1.5 Classification Based on Analysis The assessment of some EALs is based on the results of analyses that are necessary to ascertain 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 the to 0CAN031801 Page 10 of 240 associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift).

3.1.6 Emergency Director Judgment

While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-01 EAL scheme provides the Emergency Director with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to Category H). The Emergency Director will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL 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 Methodology To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, the associated IC is likewise met, the emergency classification process "clock" starts, and the ECL must be declared in accordance 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. 4.1.8).

3.2.1 Classification of Multiple Events and Conditions When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared. For example: If an Alert EAL and a Site Area Emergency EAL are met, whether at one unit or at two 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 units, an Alert should be declared. 3.2.2 Consideration of Mode Changes During Classification The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event to 0CAN031801 Page 11 of 240 or 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 emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition. For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher. 3.2.3 Classification of Imminent Conditions

Although EALs provide specific thresholds, the Emergency Director must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT). If, in the judgment of the Emergency Director, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met. While applicable to all emergency classification levels, this approach is particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures.

3.2.4 Emergency Classification Level Upgrading and Termination An ECL may be terminated when the event or condition that meets the classified IC and EAL no longer exists, and other site-specific termination requirements are met.

3.2.5 Classification of Short-Lived Events Event-based ICs and EALs define a variety of specific occurrences that have potential or actual 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 occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration. Examples of such events include an earthquake or a failure of the reactor protection system to automatically trip the reactor followed by a successful manual trip.

3.2.6 Classification of Transient Conditions

Many of the ICs and/or EALs employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions.

EAL momentarily met during expected plant response - In instances in which an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures.

to 0CAN031801 Page 12 of 240 EAL momentarily met but the condition is corrected prior to an emergency declaration - If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required. For illustrative purposes, consider the following example: An ATWS occurs and the high pressure ECCS systems fail to automatically start. The plant enters an inadequate core cooling condition (a potential loss of both the Fuel Clad and RCS Barriers). If an operator manually starts a high pressure ECCS system in accordance with an EOP step and clears the inadequate core cooling condition prior to an emergency declaration, then the classification should be 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 allow the 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 an operator is able to take a successful corrective action prior to the Emergency Director completing the review and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.

3.2.7 After-the-Fact Discovery of an Emergency Event or Condition

In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process.

In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 (ref. 4.1.3) is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50.72 (ref. 4.1.4) within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements. 3.2.8 Retraction of an Emergency Declaration

Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022 (ref. 4.1.3).

to 0CAN031801 Page 13 of 240

4.0 REFERENCES

4.1 Developmental 4.1.1 NEI 99-01 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML12326A805 4.1.2 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007. 4.1.3 NUREG-1022 Event Reporting Guidelines: 10CFR50.72 and 50.73 4.1.4 10 § CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors 4.1.5 10 § CFR 50.73 License Event Report System 4.1.6 Unit 1[2] Technical Specifications Table 1.1-1[1.1], Modes[Operational Modes]

4.1.7 Arkansas Nuclear One Offsite Dose Calculation Manual (ODCM) 4.1.8 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.9 Arkansas Nuclear One Emergency Plan 4.1.10 1015.008 Unit 2 SDC Control 4.2 Implementing 4.2.1 1903.010 Emergency Action Level Classification 4.2.2 NEI 99-01 Rev. 6 to ANO EAL Comparison Matrix 4.2.3 ANO EAL Matrix 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS 5.1 Definitions (ref. 4.1.1 except as noted)

Selected terms used in Initiating Condition, Emergency Action Level statements and EAL bases are set in all capital letters (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document. The definitions of these terms are provided below. Alert Events are in progress, or have occurred, which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION.

Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

to 0CAN031801 Page 14 of 240 Confinement Boundary The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the ANO ISFSI, the Confinement Boundary is comprised of either the Multi-assembly Sealed Basket (MSB) (SNC System) or Multi-Purpose Canister (MPC) (Holtec System).

Containment Closure The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions (ref. 4.1.10).

As applied to ANO, Containment Closure must be capable of being set within 30 minutes. Containment Closure is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

Emergency Action Level (EAL) A pre-determined, site-specific, observable threshold for an INITIATING CONDITION that, when met or exceeded, places the plant in a given emergency classification level. Emergency Classification Level (ECL) One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are:

Unusual Event (UE) Alert Site Area Emergency (SAE) General Emergency (GE) Explosion A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines 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.

Faulted The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized. to 0CAN031801 Page 15 of 240 Fire Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. Fission Product Barrier Threshold A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier. Flooding A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. General Emergency Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

Hostage A person(s) held as leverage against the station to ensure that demands will be met by the station.

Hostile Action An act toward ANO or its personnel that includes the use of violent force 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 other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on ANO. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA).

Hostile Force One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction. Imminent The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. to 0CAN031801 Page 16 of 240 Impede(d) Personnel access to a room or area is hindered to an extent that extraordinary measures 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). Independent Spent Fuel Storage Installation (ISFSI) A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. Initiating Condition (IC) An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences.

Owner Controlled Area (OCA) For the purposes of classification this is the Security Owner Controlled Area (SOCA). The SOCA is demarcated as a Vehicle Barrier System (VBS) and a detection fence on the outside and a delay fence on the inside of the passive and active barriers. The SOCA is the area inside the SOCA VBS up to the PROTECTED AREA fence line. Access to this area is controlled by the SOCA Personnel Access Control Point (ref. 4.1.9).

Projectile An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

Protected Area An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled (ref. 4.1.9).

RCS Intact The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams).

Refueling Pathway All the cavities, tubes, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel.

Ruptured The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection (automatic or manual).

to 0CAN031801 Page 17 of 240 Safety System A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the 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 result in potential offsite exposures.

Security Condition Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A Security Condition does not involve a HOSTILE ACTION.

Site Area Emergency Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; (1) toward site personnel or equipment that could lead to the likely failure of or; (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the SITE BOUNDARY.

Site Boundary That boundary defined by a 1046 meter (0.65 mile) radius around the plant (ref. 4.1.7). 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 intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Unusual Event Events are in progress or have occurred which indicate a potential degradation in the level of 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 expected unless further degradation of SAFETY SYSTEMS occurs. to 0CAN031801 Page 18 of 240 Valid An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

Visible Damage Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train. 5.2 Abbreviations/Acronyms

°F .................................................................................................................... Degrees Fahrenheit ° ......................................................................................................................................... Degrees AC ..................................................................................................................... Alternating Current ANO ............................................................................................................ Arkansas Nuclear One AOP .............................................................................................. Abnormal Operating Procedure ATWS .................................................................................... Anticipated Transient Without Scram BMS ................................................................................................... Boron Management System BWST ................................................................................................ Borated Water Storage Tank CDE ................................................................................................... Committed Dose Equivalent CET .......................................................................................................... Core Exit Thermocouple CFR ................................................................................................... Code of Federal Regulations CIAS .................................................................................. Containment Isolation Actuation Signal CMT, CNTMT, CTMT .................................................................................................. Containment CNB ................................................................................................................ Containment Barrier DBA ............................................................................................................. Design Basis Accident DBE ......................................................................................................... Design Basis Earthquake DC ............................................................................................................................. Direct Current DEF ................................................................................................................................... Defueled D/G ....................................................................................................................... Diesel Generator DHR .............................................................................................................. Decay Heat Removal DROPS ............................................................ Diverse Reactor Overpressure Protection System DSC ............................................................................................................. Dry Shielded Canister to 0CAN031801 Page 19 of 240 DSS ............................................................................................................ Diverse Scram System EAL .......................................................................................................... Emergency Action Level ECCS ......................................................................................... Emergency Core Cooling System ECL ............................................................................................... Emergency Classification Level DEF ................................................................................................................................... Defueled ENS ............................................................................................... Emergency Notification System EOF ................................................................................................ Emergency Operations Facility EOP ........................................................................................... Emergency Operating Procedure EPA ............................................................................................ Environmental Protection Agency ERG ............................................................................................ Emergency Response Guideline EPIP ............................................................................. Emergency Plan Implementing Procedure ESAS ........................................................................... Engineered Safeguards Actuation System ESF ...................................................................................................... Engineered Safety Feature ESFAS .................................................................. Engineered Safety Features Actuation System FAA ............................................................................................... Federal Aviation Administration FBI ................................................................................................ Federal Bureau of Investigation FCB ...................................................................................................................... Fuel Clad Barrier FEMA ............................................................................ Federal Emergency Management Agency GE ................................................................................................................... General Emergency HPI ............................................................................................................. High Pressure Injection IC ...................................................................................................................... Initiating Condition IPEEE ............................. Individual Plant Examination of External Events (Generic Letter 88-20)

ISFSI ......................................................................... Independent Spent Fuel Storage Installation Keff ...................................................................................... Effective Neutron Multiplication Factor LCO ................................................................................................ Limiting Condition of Operation LER ............................................................................................................. Licensee Event Report LOCA ...................................................................................................... Loss of Coolant Accident LRW .................................................................................................................... Liquid Rad Waste LTOP ............................................................................................ Low Temperature Overpressure LWR ................................................................................................................ Light Water Reactor MCC .............................................................................................................. Motor Control Center MPC ................................................ Maximum Permissible Concentration/Multi-Purpose Canister mR, mRem, mrem, mREM ............................................................ milli-Roentgen Equivalent Man MSL ...................................................................................................................... Main Steam Line to 0CAN031801 Page 20 of 240 MTS ................................................................................................................ Margin to Saturation MW .................................................................................................................................. Megawatt NDTT ...................................................................................... Nil Ductility Transition Temperature NEI ............................................................................................................ Nuclear Energy Institute NEIC ................................................................................ National Earthquake Information Center NESP ................................................................................ National Environmental Studies Project NORAD ................................................................ North American Aerospace Defense Command NOT .............................................................................................. Normal Operating Temperature (NO)UE ............................................................................................. Notification of Unusual Event NPP ................................................................................................................ Nuclear Power Plant NRC ............................................................................................ Nuclear Regulatory Commission NSSS ............................................................................................. Nuclear Steam Supply System OBE ................................................................................................... Operating Basis Earthquake OCA ........................................................................................................... Owner Controlled Area ODCM ........................................................................................ Off-site Dose Calculation Manual ORO ............................................................................................... Offsite Response Organization PA ........................................................................................................................... Protected Area PAG ..................................................................................................... Protective Action Guideline PRA/PSA .................................. Probabilistic Risk Assessment / Probabilistic Safety Assessment P-T .............................................................................................................. Pressure-Temperature PTS ..................................................................................................... Pressurized Thermal Shock PWR ..................................................................................................... Pressurized Water Reactor PSIG ............................................................................................ Pounds per Square Inch Gauge R ..................................................................................................................................... Roentgen RB ......................................................................................................................... Reactor Building RCC ......................................................................................................... Reactor Control Console RCB ............................................................................................. Reactor Coolant System Barrier RCP ............................................................................................................ Reactor Coolant Pump RCS ......................................................................................................... Reactor Coolant System Rem, rem, REM ..................................................................................... Roentgen Equivalent Man Rep CET ....................................................................... Representative Core Exit Thermocouples RETS ...................................................................... Radiological Effluent Technical Specifications RPS ...................................................................................................... Reactor Protection System RV ........................................................................................................................... Reactor Vessel to 0CAN031801 Page 21 of 240 RVLMS ........................................................................... Reactor Vessel Level Monitoring System RWT ............................................................................................................. Refueling Water Tank SAR ............................................................................................................. Safety Analysis Report SBO ...................................................................................................................... Station Blackout SCBA ................................................................................... Self-Contained Breathing Apparatus SDC ................................................................................................................... Shutdown Cooling SOCA ........................................................................................... Security Owner Controlled Area SG ....................................................................................................................... Steam Generator SI ............................................................................................................................ Safety Injection SPDS ........................................................................................ Safety Parameter Display System SPING ..................................................................................... Super Particulate Iodine Noble Gas SRO ......................................................................................................... Senior Reactor Operator TEDE ............................................................................................ Total Effective Dose Equivalent TOAF ................................................................................................................. Top of Active Fuel TSC ........................................................................................................ Technical Support Center USGS .......................................................................................... United States Geological Survey VBS ............................................................................................................ Vehicle Barrier System 6.0 ANO-TO-NEI 99-01 REV. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of an ANO EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the ANO EALs based on the NEI guidance can be found in the EAL Comparison Matrix.

ANO NEI 99-01 Rev. 6 EAL IC Example EAL AU1.1 AU1 1, 2 AU1.2 AU1 3 AU2.1 AU2 1 AA1.1 AA1 1 AA1.2 AA1 2 AA1.3 AA1 3 AA1.4 AA1 4 AA2.1 AA2 1 to 0CAN031801 Page 22 of 240 ANO NEI 99-01 Rev. 6 EAL IC Example EAL AA2.2 AA2 2 AA2.3 AA2 3 AA3.1 AA3 1 AA3.2 AA3 2 AS1.1 AS1 1 AS1.2 AS1 2 AS1.3 AS1 3 AS2.1 AS2 1 AG1.1 AG1 1 AG1.2 AG1 2 AG1.3 AG1 3 AG2.1 AG2 1 CU1.1 CU1 1 CU1.2 CU1 2 CU2.1 CU2 1 CU3.1 CU3 1 CU3.2 CU3 2 CU4.1 CU4 1 CU5.1 CU5 1, 2, 3 CA1.1 CA1 1 CA1.2 CA1 2 CA2.1 CA2 1 CA3.1 CA3 1, 2 CA6.1 CA6 1 CS1.1 CS1 1 CS1.2 CS1 2 CS1.3 CS1 3 CG1.1 CG1 1 CG1.2 CG1 2 EU1.1 EU1 1 FA1.1 FA1 1 to 0CAN031801 Page 23 of 240 ANO NEI 99-01 Rev. 6 EAL IC Example EAL FS1.1 FS1 1 FG1.1 FG1 1 HU1.1 HU1 1, 2, 3 HU2.1 HU2 1 HU3.1 HU3 1 HU3.2 HU3 2 HU3.3 HU3 3 HU3.4 HU3 4 HU4.1 HU4 1 HU4.2 HU4 2 HU4.3 HU4 3 HU4.4 HU4 4 HU7.1 HU7 1 HA1.1 HA1 1, 2 HA5.1 HA5 1 HA6.1 HA6 1 HA7.1 HA7 1 HS1.1 HS1 1 HS6.1 HS6 1 HS7.1 HS7 1 HG7.1 HG7 1 SU1.1 SU1 1 SU3.1 SU2 1 SU4.1 SU3 1 SU4.2 SU3 2 SU5.1 SU4 1, 2, 3 SU6.1 SU5 1 SU6.2 SU5 2 SU7.1 SU6 1, 2, 3 SU8.1 SU7 1, 2 SA1.1 SA1 1 to 0CAN031801 Page 24 of 240 ANO NEI 99-01 Rev. 6 EAL IC Example EAL SA3.1 SA2 1 SA6.1 SA5 1 SA9.1 SA9 1 SS1.1 SS1 1 SS2.1 SS8 1 SS6.1 SS5 1 SG1.1 SG1 1 SG1.2 SG8 1 7.0 ATTACHMENTS 7.1 Attachment 1, Emergency Action Level Technical Bases 7.2 Attachment 2, Safe Operation & Shutdown Areas Tables 1[2]A-3 & 1[2]H-2 Bases

to 0CAN031801 Page 25 of 240 Attachment 1 - Emergency Action Level Technical Bases Category A - Abnormal Rad Levels / Rad Effluent EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via non-radiological symptoms. Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emergency classification.

At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety.

Events of this category pertain to the following subcategories: 1. Radiological Effluent Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits. 2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification. 3. Area Radiation Levels Sustained general area radiation levels which may preclude access to areas requiring continuous occupancy also warrant emergency classification.

to 0CAN031801 Page 26 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer EAL: AU1.1 Unusual Event Reading on any Table 1[2]A-1 effluent radiation monitor > column "UE" for 60 min. (Notes 1, 2, 3) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Table 1A-1 Unit 1 Effluent Monitor Classification Thresholds (2 min. avg reading) Release Point Monitor GE SAE Alert UE Gaseous Containment Purge RX-9820 (SPING 1) 4.15E+01 Ci/cc 4.15E+00 µCi/cc4.15E-01 µCi/cc 1.21E-03 µCi/cc Radwaste Area RX-9825 (SPING 2) 2.67E+01 Ci/cc 2.67E+00 µCi/cc2.67E-01 µCi/cc 4.94E-04 µCi/cc Fuel Handling Area RX-9830 (SPING 3) 6.20E+02 Ci/cc 6.20E+01 µCi/cc6.20E+00 µCi/cc 5.44E-04 µCi/cc Emergency Penetration Room RX-9835 (SPING 4) 6.55E+02 Ci/cc 6.55E+01 µCi/cc6.55E+00 µCi/cc 1.21E-02 µCi/cc Liquid Liquid Radwaste RE-4642 ---- ---- ---- 2.46E+05 cpm to 0CAN031801 Page 27 of 240 Attachment 1 - Emergency Action Level Technical Bases Table 2A-1 Unit 2 Effluent Monitor Classification Thresholds (2 min. avg reading) Release Point Monitor GE SAE Alert UE Gaseous Containment Purge 2RX-9820 (SPING 5) 1.88E+01 Ci/cc 1.88E+00 µCi/cc1.88E-01 µCi/cc 5.48E-04 µCi/cc Radwaste Area 2RX-9825 (SPING 6) 2.35E+01 Ci/cc 2.35E+00 µCi/cc2.35E-01 µCi/cc 4.35E-04 µCi/cc Fuel Handling Area 2RX-9830 (SPING 7) 6.86E+02 Ci/cc 6.86E+01 µCi/cc6.86E+00 µCi/cc 6.04E-04 µCi/cc Emergency Penetration Room 2RX-9835 (SPING 8) 5.88E+02 Ci/cc 5.88E+01 µCi/cc5.88E+00 µCi/cc 1.09E-02 µCi/cc Liquid BMS Liquid Discharge 2RE-2330 ---- ---- ---- 2.45E+04 cpm Regenerative Waste Discharge 2RE-4423 ---- ---- ---- 2.45E+05 cpm Mode Applicability: All Definition(s): VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This IC addresses a potential reduction 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 is normally prepared.

Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.

to 0CAN031801 Page 28 of 240 Attachment 1 - Emergency Action Level Technical Bases Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways as well as radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. Such releases are typically associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas). Escalation of the emergency classification level would be via IC AA1.

Reference(s):

1. OP-1604.051 Eberline Radiation Monitor System 2. Offsite Dose Calculation Manual

3. EP-CALC-ANO-1701 Radiological Effluent EAL Values

4. NEI 99-01 AU1 to 0CAN031801 Page 29 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer EAL: AU1.2 Unusual Event Sample analysis for a gaseous or liquid release indicates a concentration or release rate 2 x ODCM limits for 60 min. (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All

Definition(s): None Basis: This IC addresses a potential reduction 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 is normally prepared. Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

to 0CAN031801 Page 30 of 240 Attachment 1 - Emergency Action Level Technical Bases Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.).

Escalation of the emergency classification level would be via IC AA1. Reference(s): 1. Offsite Dose Calculation Manual 2. NEI 99-01 AU1 to 0CAN031801 Page 31 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL: AA1.1 Alert Reading on any Table 1[2]A-1 effluent radiation monitor > column "ALERT" for 15 min. (Notes 1, 2, 3, 4) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Note 4 The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Table 1A-1 Unit 1 Effluent Monitor Classification Thresholds (2 min. avg reading) Release Point Monitor GE SAE Alert UE Gaseous Containment Purge RX-9820 (SPING 1) 4.15E+01 Ci/cc 4.15E+00 µCi/cc4.15E-01 µCi/cc 1.21E-03 µCi/cc Radwaste Area RX-9825 (SPING 2) 2.67E+01 Ci/cc 2.67E+00 µCi/cc2.67E-01 µCi/cc 4.94E-04 µCi/cc Fuel Handling Area RX-9830 (SPING 3) 6.20E+02 Ci/cc 6.20E+01 µCi/cc6.20E+00 µCi/cc 5.44E-04 µCi/cc Emergency Penetration Room RX-9835 (SPING 4) 6.55E+02 Ci/cc 6.55E+01 µCi/cc6.55E+00 µCi/cc 1.21E-02 µCi/cc Liquid Liquid Radwaste RE-4642 ---- ---- ---- 2.46E+05 cpm to 0CAN031801 Page 32 of 240 Attachment 1 - Emergency Action Level Technical Bases Table 2A-1 Unit 2 Effluent Monitor Classification Thresholds (2 min. avg reading) Release Point Monitor GE SAE Alert UE Gaseous Containment Purge 2RX-9820 (SPING 5) 1.88E+01 Ci/cc 1.88E+00 µCi/cc1.88E-01 µCi/cc 5.48E-04 µCi/cc Radwaste Area 2RX-9825 (SPING 6) 2.35E+01 Ci/cc 2.35E+00 µCi/cc2.35E-01 µCi/cc 4.35E-04 µCi/cc Fuel Handling Area 2RX-9830 (SPING 7) 6.86E+02 Ci/cc 6.86E+01 µCi/cc6.86E+00 µCi/cc 6.04E-04 µCi/cc Emergency Penetration Room 2RX-9835 (SPING 8) 5.88E+02 Ci/cc 5.88E+01 µCi/cc5.88E+00 µCi/cc 1.09E-02 µCi/cc Liquid BMS Liquid Discharge 2RE-2330 ---- ---- ---- 2.45E+04 cpm Regenerative Waste Discharge 2RE-4423 ---- ---- ---- 2.45E+05 cpm Mode Applicability: All Definition(s): VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. to 0CAN031801 Page 33 of 240 Attachment 1 - Emergency Action Level Technical Bases The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Escalation of the emergency classification level would be via IC AS1.

Reference(s): 1. OP-1604.051 Eberline Radiation Monitor System 2. EP-CALC-ANO-1701 Radiological Effluent EAL Values

3. NEI 99-01 AA1 to 0CAN031801 Page 34 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL: AA1.2 Alert Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4) Note 4: The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Mode Applicability: All

Definition(s): SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant. Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release). Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the 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 CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Escalation of the emergency classification level would be via IC AS1. Reference(s): 1. OP-1904.002 Offsite Dose Projections 2. NEI 99-01 AA1 to 0CAN031801 Page 35 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL: AA1.3 Alert Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All Definition(s): SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant. Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release). Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the 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 CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. This EAL is assessed per the ODCM (ref. 2). Escalation of the emergency classification level would be via IC AS1. to 0CAN031801 Page 36 of 240 Attachment 1 - Emergency Action Level Technical Bases Reference(s): 1. OP-1904.002 Offsite Dose Projections 2. Offsite Dose Calculation Manual

3. NEI 99-01 AA1 to 0CAN031801 Page 37 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL: AA1.4 Alert Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: Closed window dose rates > 10 mR/hr expected to continue for 60 min. Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All Definition(s):

SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant. Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

to 0CAN031801 Page 38 of 240 Attachment 1 - Emergency Action Level Technical Bases The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Escalation of the emergency classification level would be via IC AS1.

Reference(s):

1. OP-1905.002 Offsite Emergency Monitoring 2. NEI 99-01 AA1 to 0CAN031801 Page 39 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL: AS1.1 Site Area Emergency Reading on any Table A-1 effluent radiation monitor > column "SAE" for 15 min. (Notes 1, 2, 3, 4) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Note 4: The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Table 1A-1 Unit 1 Effluent Monitor Classification Thresholds (2 min. avg reading) Release Point Monitor GE SAE Alert UE Gaseous Containment Purge RX-9820 (SPING 1) 4.15E+01 Ci/cc 4.15E+00 µCi/cc4.15E-01 µCi/cc 1.21E-03 µCi/cc Radwaste Area RX-9825 (SPING 2) 2.67E+01 Ci/cc 2.67E+00 µCi/cc2.67E-01 µCi/cc 4.94E-04 µCi/cc Fuel Handling Area RX-9830 (SPING 3) 6.20E+02 Ci/cc 6.20E+01 µCi/cc6.20E+00 µCi/cc 5.44E-04 µCi/cc Emergency Penetration Room RX-9835 (SPING 4) 6.55E+02 Ci/cc 6.55E+01 µCi/cc6.55E+00 µCi/cc 1.21E-02 µCi/cc Liquid Liquid Radwaste RE-4642 ---- ---- ---- 2.46E+05 cpm to 0CAN031801 Page 40 of 240 Attachment 1 - Emergency Action Level Technical Bases Table 2A-1 Unit 2 Effluent Monitor Classification Thresholds (2 min. avg reading) Release Point Monitor GE SAE Alert UE Gaseous Containment Purge 2RX-9820 (SPING 5) 1.88E+01 Ci/cc 1.88E+00 µCi/cc1.88E-01 µCi/cc 5.48E-04 µCi/cc Radwaste Area 2RX-9825 (SPING 6) 2.35E+01 Ci/cc 2.35E+00 µCi/cc2.35E-01 µCi/cc 4.35E-04 µCi/cc Fuel Handling Area 2RX-9830 (SPING 7) 6.86E+02 Ci/cc 6.86E+01 µCi/cc6.86E+00 µCi/cc 6.04E-04 µCi/cc Emergency Penetration Room 2RX-9835 (SPING 8) 5.88E+02 Ci/cc 5.88E+01 µCi/cc5.88E+00 µCi/cc 1.09E-02 µCi/cc Liquid BMS Liquid Discharge 2RE-2330 ---- ---- ---- 2.45E+04 cpm Regenerative Waste Discharge 2RE-4423 ---- ---- ---- 2.45E+05 cpm Mode Applicability: All Definition(s): VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

to 0CAN031801 Page 41 of 240 Attachment 1 - Emergency Action Level Technical Bases The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was 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 the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Escalation of the emergency classification level would be via IC AG1.

Reference(s):

1. OP-1604.051 Eberline Radiation Monitor System 2. EP-CALC-ANO-1701 Radiological Effluent EAL Values

3. NEI 99-01 AS1 to 0CAN031801 Page 42 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL: AS1.2 Site Area Emergency Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4) Note 4 The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Mode Applicability: All

Definition(s): SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant). Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the 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 CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Escalation of the emergency classification level would be via IC AG1.

Reference(s): 1. OP-1904.002 Offsite Dose Projections 2. NEI 99-01 AS1 to 0CAN031801 Page 43 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL: AS1.3 Site Area Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: Closed window dose rates > 100 mR/hr expected to continue for 60 min. Analyses of field survey samples indicate thyroid CDE > 500 mrem for 60 min. of inhalation (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All Definition(s):

SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the 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 CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Escalation of the emergency classification level would be via IC AG1. to 0CAN031801 Page 44 of 240 Attachment 1 - Emergency Action Level Technical Bases Reference(s): 1. OP-1905.002 Offsite Emergency Monitoring

2. NEI 99-01 AS1 to 0CAN031801 Page 45 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL: AG1.1 General Emergency Reading on any Table A-1 effluent radiation monitor > column "GE" for 15 min. (Notes 1, 2, 3, 4) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Note 4: The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Table 1A-1 Unit 1 Effluent Monitor Classification Thresholds (2 min. avg reading) Release Point Monitor GE SAE Alert UE Gaseous Containment Purge RX-9820 (SPING 1) 4.15E+01 Ci/cc 4.15E+00 µCi/cc4.15E-01 µCi/cc 1.21E-03 µCi/cc Radwaste Area RX-9825 (SPING 2) 2.67E+01 Ci/cc 2.67E+00 µCi/cc2.67E-01 µCi/cc 4.94E-04 µCi/cc Fuel Handling Area RX-9830 (SPING 3) 6.20E+02 Ci/cc 6.20E+01 µCi/cc6.20E+00 µCi/cc 5.44E-04 µCi/cc Emergency Penetration Room RX-9835 (SPING 4) 6.55E+02 Ci/cc 6.55E+01 µCi/cc6.55E+00 µCi/cc 1.21E-02 µCi/cc Liquid Liquid Radwaste RE-4642 ---- ---- ---- 2.46E+05 cpm to 0CAN031801 Page 46 of 240 Attachment 1 - Emergency Action Level Technical Bases Table 2A-1 Unit 2 Effluent Monitor Classification Thresholds (2 min. avg reading) Release Point Monitor GE SAE Alert UE Gaseous Containment Purge 2RX-9820 (SPING 5) 1.88E+01 Ci/cc 1.88E+00 µCi/cc1.88E-01 µCi/cc 5.48E-04 µCi/cc Radwaste Area 2RX-9825 (SPING 6) 2.35E+01 Ci/cc 2.35E+00 µCi/cc2.35E-01 µCi/cc 4.35E-04 µCi/cc Fuel Handling Area 2RX-9830 (SPING 7) 6.86E+02 Ci/cc 6.86E+01 µCi/cc6.86E+00 µCi/cc 6.04E-04 µCi/cc Emergency Penetration Room 2RX-9835 (SPING 8) 5.88E+02 Ci/cc 5.88E+01 µCi/cc5.88E+00 µCi/cc 1.09E-02 µCi/cc Liquid BMS Liquid Discharge 2RE-2330 ---- ---- ---- 2.45E+04 cpm Regenerative Waste Discharge 2RE-4423 ---- ---- ---- 2.45E+05 cpm Mode Applicability: All Definition(s): VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

to 0CAN031801 Page 47 of 240 Attachment 1 - Emergency Action Level Technical Bases The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was 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 the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Reference(s): 1. OP-1604.051 Eberline Radiation Monitor System 2. EP-CALC-ANO-1701 Radiological Effluent EAL Values 3. NEI 99-01 AG1 to 0CAN031801 Page 48 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL: AG1.2 General Emergency Dose assessment using actual meteorology indicates doses > 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4) Note 4 The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Mode Applicability: All

Definition(s): SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant). Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the 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 was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Reference(s):

1. OP-1904.002 Offsite Dose Projections 2. NEI 99-01 AG1 to 0CAN031801 Page 49 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL: AG1.3 General Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: Closed window dose rates > 1,000 mR/hr expected to continue for 60 min. Analyses of field survey samples indicate thyroid CDE > 5,000 mrem for 60 min. of inhalation (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All Definition(s):

SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the 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 was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. to 0CAN031801 Page 50 of 240 Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. OP-1905.002 Offsite Emergency Monitoring 2. NEI 99-01 AG1 to 0CAN031801 Page 51 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: UNPLANNED loss of water level above irradiated fuel EAL: AU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by low water level alarm, visual observation, or BWST[RWT] level drop due to makeup demands AND UNPLANNED rise in corresponding area radiation levels as indicated by any of the following radiation monitors: Unit 1 o RE-8009 Spent Fuel Area o RE-8017 Fuel Handling Area Unit 2 o 2RE-8914 Spent Fuel Area o 2RE-8915 Spent Fuel Area o 2RE-8916 Spent Fuel Area o 2RE-8912 Containment Incore Instrumentation Mode Applicability: All

Definition(s): UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. REFUELING PATHWAY - All the cavities, tubes, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel. to 0CAN031801 Page 52 of 240 Attachment 1 - Emergency Action Level Technical Bases Basis:

This IC addresses a drop in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant.

A water level drop will be primarily determined by indications from available level instrumentation. 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 in the water level may also cause a rise in the radiation levels of adjacent areas that can be detected by monitors in those locations. The effects of planned evolutions should be considered. For example, a refueling bridge area radiation monitor reading may rise due to planned evolutions such as lifting of the reactor vessel head or movement of a fuel assembly. Note that this EAL is applicable only in cases where 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 in accordance Recognition Category C during the Cold Shutdown and Refueling modes.

Escalation of the emergency classification level would be via IC AA2.

Reference(s): 1. OP-1203.050 Unit 1 Spent Fuel Pool Emergencies 2. OP-2203.002 Spent Fuel Pool Emergencies

3. 1SAR 11.2.5 Area Radiation Monitoring Systems Table 11-15 Area Radiation Monitors

4. 2SAR 12.1.4 Area Radiation Monitoring System 5. NEI 99-01 AU2 to 0CAN031801 Page 53 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL: AA2.1 Alert IMMINENT uncovery of irradiated fuel in the REFUELING PATHWAY. Mode Applicability: All Definition(s): CONFINEMENT BOUNDARY - The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the ANO ISFSI, the Confinement Boundary is comprised of either the Multi-assembly Sealed Basket (MSB) (SNC System) or Multi-Purpose Canister (MPC) (Holtec System).

IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

REFUELING PATHWAY - All the cavities, tubes, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel. Basis: This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the REFUELING PATHWAY. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant. This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC EU1.

This EAL escalates from AU2.1 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. 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 and radiation 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 available indications, reports and observations. to 0CAN031801 Page 54 of 240 Attachment 1 - Emergency Action Level Technical Bases While an area radiation monitor could detect a rise in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should 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 in accordance with Recognition Category C during the Cold Shutdown and Refueling modes.

Escalation of the emergency classification level would be via IC AS1.

Reference(s): 1. NEI 99-01 AA2

to 0CAN031801 Page 55 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL: AA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity AND High alarm on any Table 1[2]A-2 radiation monitor. Table 1A-2 Unit 1 Fuel Damage Radiation Monitors RE-8009 Spent Fuel Area RE-8017 Fuel Handling RE-8060 Containment High Range Radiation Monitors RE-8061 Containment High Range Radiation Monitors RX-9820 (SPING 1) Containment Purge RX-9825 (SPING 2) Radwaste Area RX-9830 (SPING 3) Fuel Handling Area Table 2A-2 Unit 2 Fuel Damage Radiation Monitors 2RE-8905 Containment Equipment Hatch Area 2RE-8909 Containment Personnel Access Area 2RE-8912 Containment Incore Inst. 2RE-8914 Spent Fuel Area 2RE-8915 Spent Fuel Area 2RE-8916 Spent Fuel Area 2RE-8925-1 Containment High Range Radiation Monitors 2RE-8925-2 Containment High Range Radiation Monitors 2RX-9820 (SPING 5) Containment Purge 2RX-9825 (SPING 6) Radwaste Area 2RX-9830 (SPING 7) Fuel Handling Area to 0CAN031801 Page 56 of 240 Attachment 1 - Emergency Action Level Technical Bases Mode Applicability:

All Definition(s): CONFINEMENT BOUNDARY - The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the ANO ISFSI, the Confinement Boundary is comprised of either the Multi-assembly Sealed Basket (MSB) (SNC System) or Multi-Purpose Canister (MPC) (Holtec System).

VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis: This EAL addresses events that have caused actual damage to an irradiated fuel assembly.

These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant. This EAL applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC EU1.

This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident).

Escalation of the emergency classification level would be via IC AS1. Reference(s): 1. OP-1203.050 Unit 1 Spent Fuel Pool Emergencies 2. OP-1305.001 Radiation Monitoring System Check and Test 3. OP-2203.002 Spent Fuel Pool Emergencies 4. OP-1604.051 Eberline Radiation Monitoring System

5. OP-2304.133 Containment High Range Radiation Monitor Calibration

6. Offsite Dose Calculation Manual

7. NEI 99-01 AA2 to 0CAN031801 Page 57 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL: AA2.3 Alert Lowering of spent fuel pool level to 387.0 ft.[389.5 ft.] (Alarm 2) on LIT-2020-3(4) [2LIT-2020-1(2)] Mode Applicability: All Definition(s):

None Basis: This EAL addresses events that have caused a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.

Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool.

Escalation of the emergency classification level would be via IC AS1 or AS2. Post-Fukushima order EA 051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3) (ref. 1, 2).

Reference(s): 1. MOHR-ANO-1, ANO-1 SFPI (Level) Configuration, Sheet 1, Revision 0 2. MOHR-ANO-2, ANO-2 SFPI (Level) Configuration, Sheet 1, Revision 0

3. NEI 99-01 AA2 to 0CAN031801 Page 58 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level at the top of the fuel racks EAL: AS2.1 Site Area Emergency Lowering of spent fuel pool level to 377.0 ft.[379.5 ft.] (Alarm 3) on LIT-2020-3(4) [2LIT-2020-1(2)] Mode Applicability: All Definition(s):

IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

Basis:

This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity. Escalation of the emergency classification level would be via IC AG1 or AG2.

Post-Fukushima order EA 051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3) (ref. 1, 2). Reference(s): 1. MOHR-ANO-1, ANO-1 SFPI (Level) Configuration, Sheet 1, Revision 0 2. MOHR-ANO-2, ANO-2 SFPI (Level) Configuration, Sheet 1, Revision 0 3. NEI 99-01 AS2 to 0CAN031801 Page 59 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer EAL: AG2.1 General Emergency Spent fuel pool level cannot be restored to at least 377.0 ft.[379.5 ft.] (Alarm 3) on LIT-2020-3(4)[2LIT-2020-1(2)] for 60 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: All

Definition(s): None Basis: This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment.

It is recognized that this IC would likely not be met until well after another General Emergency IC was met; however, it is included to provide classification diversity.

Post-Fukushima order EA 051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3) (ref. 1, 2). Reference(s): 1. MOHR-ANO-1, ANO-1 SFPI (Level) Configuration, Sheet 1, Revision 0 2. MOHR-ANO-2, ANO-2 SFPI (Level) Configuration, Sheet 1, Revision 0 3. NEI 99-01 AG2 to 0CAN031801 Page 60 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 3 - Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL: AA3.1 Alert Dose rate > 15 mR/hr in EITHER of the following areas: Control Room Central Alarm Station (by survey) Mode Applicability:

All Definition(s): IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures 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). Basis: Areas that meet this threshold include the Control Room (CR) and the Central Alarm Station (CAS). The Control Room envelope (Unit 1 and Unit 2) is monitored for excessive radiation by five detectors. These radiation detectors are RE-8001, 2RE-8001A, 2RE-8001B, 2RE-8750-1A, and 2RE-8750-1B (ref. 1). The CAS is included in this EAL because of its importance to permitting access to areas required to assure safe plant operations. There are no permanently installed area radiation monitors in CAS that may be used to assess this EAL threshold. Therefore, this threshold is evaluated using local radiation survey for this area.

This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or IMPEDE personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the rise in radiation levels and determine if another IC may be applicable.

Escalation of the emergency classification level would be via Recognition Category A, C or F ICs.

to 0CAN031801 Page 61 of 240 Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. STM 1-62 Radiation Monitoring 2. NEI 99-01 AA3 to 0CAN031801 Page 62 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 3 - Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL: AA3.2 Alert An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table 1[2]A-3 room or area (Note 5) Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. Table 1A-3 Unit 1 Safe Operation & Shutdown Rooms/Areas Room/Area Mode A-4 Switchgear Room 3, 4 Upper North Electrical Penetration Room 3, 4 Lower South Electrical Equipment Room 3, 4 Table 2A-3 Unit 2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Aux Building 317' Emergency Core Cooling Rooms 3, 4 Aux Building 317' Tendon Gallery Access 3, 4 Aux Building 335' Charging Pumps / MCC 2B-52 3, 4 Aux Building 354' MCC 2B-62 Area 3, 4 Emergency Diesel Generator Corridor 3, 4 Lower South Piping Penetration Room 3, 4 Aux Building 386' Containment Hatch 3, 4 to 0CAN031801 Page 63 of 240 Attachment 1 - Emergency Action Level Technical Bases Mode Applicability:

3 - Hot Standby, 4 - Hot Shutdown Definition(s): IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures 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 intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or IMPEDE personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the rise in radiation levels and determine if another IC may be applicable.

For AA3.2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the higher radiation levels. Access should be considered as IMPEDED if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding, requiring use of non-routine protective equipment, 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 affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation rise occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 3. The higher radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., radiography, spent filter or resin transfer, etc.). The access control measures are of a conservative or precautionary nature, and would not actually prevent or IMPEDE a required action.

Escalation of the emergency classification level would be via Recognition Category A, C or F ICs. to 0CAN031801 Page 64 of 240 Attachment 1 - Emergency Action Level Technical Bases If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact 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 those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1).

EAL AA3.2 mode applicability has been limited to the mode limitations of Table 1[2]A-3 (Modes 3 and 4 only). Reference(s): 1. Attachment 2 Safe Operation & Shutdown Areas Tables 1[2]A-3 & 1[2]H-2 Bases 2. NEI 99-01 AA3 to 0CAN031801 Page 65 of 240 Attachment 1 - Emergency Action Level Technical Bases Category C - Cold Shutdown / Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature 200°F); EALs in this category are applicable only in one or more cold operating modes. Category C EALs are directly associated with cold shutdown or refueling system safety functions. Given the variability of plant configurations (e.g., systems out-of-service for maintenance, containment open, reduced AC power redundancy, time since shutdown) during these periods, the consequences of any given initiating event can vary greatly. For example, a loss of decay heat removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable. The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, CONTAINMENT CLOSURE, and fuel clad integrity for the applicable operating modes (5 - Cold Shutdown, 6 - Refueling, DEF - Defueled). The events of this category pertain to the following subcategories: 1. RCS Level RCS water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity. 2. Loss of Vital AC Power Loss of vital plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite power sources for 4.16 KV vital buses. 3. RCS Temperature Uncontrolled or inadvertent temperature or pressure rises are indicative of a potential loss of safety functions. 4. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the 125V DC vital buses. 5. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification. 6. Hazardous Event Affecting Safety Systems Certain hazardous natural and technological events may result in VISIBLE DAMAGE to or degraded performance of safety systems warranting classification. to 0CAN031801 Page 66 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: UNPLANNED loss of RCS inventory EAL: CU1.1 Unusual Event UNPLANNED loss of reactor coolant results in RCS water level less than a required lower limit for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 5 - Cold Shutdown, 6 - Refueling Definition(s): UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis: This EAL 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 RCS level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.

Refueling evolutions that lower RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level lowering below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered. This EAL recognizes that the minimum required RCS level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document. to 0CAN031801 Page 67 of 240 Attachment 1 - Emergency Action Level Technical Bases The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level.

Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.

Reference(s): 1. OP-1015.002 Decay Heat Removal and LTOP System 2. OP-1015.008 Unit 2 SDC Control

3. NEI 99-01 CU1 to 0CAN031801 Page 68 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: UNPLANNED loss of RCS inventory EAL: CU1.2 Unusual Event RCS level cannot be monitored AND EITHER UNPLANNED rise in any Table 1[2]C-1 sump/tank level due to loss of RCS inventory Visual observation of UNISOLABLE RCS leakage Table 1C-1 Unit 1 Sumps / Tanks Reactor Building Sump Reactor Drain Tank Aux. Building Equipment Drain Tank Aux. Building Sump Quench Tank Table 2C-1 Unit 2 Sumps / Tanks CNTMT Sump Reactor Drain Tank LRW Waste Tank (2T-20) Holdup Tank Aux. Building Sump Quench Tank Mode Applicability:

5 - Cold Shutdown, 6 - Refueling to 0CAN031801 Page 69 of 240 Attachment 1 - Emergency Action Level Technical Bases Definition(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 intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. 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 RCS level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.

Refueling evolutions that lower RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level lowering below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered.

This EAL addresses a condition where all means to determine RCS level have been lost. In this condition, operators may determine that an inventory 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 water flow to ensure they are indicative of leakage from the RCS.

Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.

Reference(s): 1. OP-1203.039 Excess RCS Leakage 2. OP-2203.016 Excess RCS Leakage

3. NEI 99-01 CU1 to 0CAN031801 Page 70 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Significant Loss of RCS inventory EAL: CA1.1 Alert Loss of RCS inventory as indicated by EITHER: RVLMS Levels 1 through 8[1 through 5] indicate DRY Reactor vessel level 368.5 ft. (LT-1195/LT-1196)[0 in. (L4791/L4792)] (bottom of hot leg) Mode Applicability: 5 - Cold Shutdown, 6 - Refueling Definition(s): None Basis: This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.

For this EAL, a lowering of RPV water level below the specified level indicates that operator actions have not been successful in restoring and maintaining RCS water level. The heat-up rate of the coolant will rise as the available water inventory is reduced. A continuing drop in water level will lead to core uncovery.

Although related, this EAL is concerned with the loss of RPV inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Decay Heat Removal suction point). A rise in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.

If water level continues to lower, then escalation to Site Area Emergency would be via IC CS1. The bottom of the RCS hot leg penetration into the reactor vessel is approximately RLVMS Level 8 (Unit 1) or RVLMS Level 5 (Unit 2). However, RVLMS may not be available in the cold shutdown modes. Redundant means of level indication is provided in these modes and included in this EAL. The bottom of the RCS hot leg penetration into the reactor vessel is 368 ft., 0 in. (Unit 1) or 369 ft., 1.5 in. (Unit 2). Below this level, reactor vessel level indication may be lost and loss of suction to decay heat removal systems will occur (ref. 1, 2, 3). Where to 0CAN031801 Page 71 of 240 Attachment 1 - Emergency Action Level Technical Bases redundant means of level indication cannot read below this level, the lowest indication that can be read on scale is used for this EAL. The inability to restore and maintain level after reaching this setpoint would be indicative of a failure of the RCS barrier. Reference(s): 1. OP-1105.008 Inadequate Core Cooling Monitor and Display 2. OP-2105.003 Reactor Vessel Level Monitoring System Operations 3. Calculation No. 90-E-0116-01 ANO-2 EOP Setpoint Basis Document, Setpoints R.3 and R.9

4. NEI 99-01 CA1 to 0CAN031801 Page 72 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Significant Loss of RCS inventory EAL: CA1.2 Alert RCS level cannot be monitored for 15 min. (Note 1) AND EITHER UNPLANNED rise in any Table 1[2]C-1 sump/tank level due to a loss of RCS inventory Visual observation of UNISOLABLE RCS leakage Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table 1C-1 Unit 1 Sumps / Tanks Reactor Building Sump Reactor Drain Tank Aux. Building Equipment Drain Tank Aux. Building Sump Quench Tank Table 2C-1 Unit 2 Sumps / Tanks CNTMT Sump Reactor Drain Tank LRW Waste Tank (2T-20) Holdup Tank Aux. Building Sump Quench Tank to 0CAN031801 Page 73 of 240 Attachment 1 - Emergency Action Level Technical Bases Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Definition(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 intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis: This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.

For this EAL, the inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine 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 water flow to ensure they are indicative of leakage from the RCS.

The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1.

If the RCS inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1. Reference(s): 1. OP-1203.039 Excess RCS Leakage 2. OP-2203.016 Excess RCS Leakage 3. NEI 99-01 CA1 to 0CAN031801 Page 74 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability EAL: CS1.1 Site Area Emergency CONTAINMENT CLOSURE not established AND RVLMS Levels 1 through 9[1 through 6] indicate DRY Mode Applicability: 5 - Cold Shutdown, 6 - Refueling Definition(s): CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: This IC addresses a significant and prolonged loss of RCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection 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 will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.

Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions.

to 0CAN031801 Page 75 of 240 Attachment 1 - Emergency Action Level Technical Bases This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; 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 United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

Escalation of the emergency classification level would be via IC CG1 or AG1.

Reference(s): 1. OP-1105.008 Inadequate Core Cooling Monitor and Display 2. OP-2105.003 Reactor Vessel Level Monitoring System Operations

3. NEI 99-01 CS1 to 0CAN031801 Page 76 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability EAL: CS1.2 Site Area Emergency [RVLMS Levels 1 through 7 indicate DRY OR] RCS level cannot be monitored for 30 min. (Note 1) AND Core uncovery is indicated by any of the following: UNPLANNED rise in any Table 1[2]C-1 sump/tank level of sufficient magnitude to indicate core uncovery Containment high range radiation monitor RE-8060/8061[2RE-8925-1/8925-2] reading > 10 R/hr Erratic Source Range Monitor indication Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table 1C-1 Unit 1 Sumps / Tanks Reactor Building Sump Reactor Drain Tank Aux. Building Equipment Drain Tank Aux. Building Sump Quench Tank to 0CAN031801 Page 77 of 240 Attachment 1 - Emergency Action Level Technical Bases Table 2C-1 Unit 2 Sumps / Tanks CNTMT Sump Reactor Drain Tank LRW Waste Tank (2T-20) Holdup Tank Aux. Building Sump Quench Tank Mode Applicability: 5 - Cold Shutdown, 6 - Refueling

Definition(s): CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: When in service, the Unit 2 RVLMS can measure RCS level below the top of active fuel.

Level 7 DRY on this system is an indication of core uncovery.

This IC addresses a significant and prolonged loss of reactor vessel/RCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. to 0CAN031801 Page 78 of 240 Attachment 1 - Emergency Action Level Technical Bases Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.

Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions.

The difference in the specified RCS levels of EALs CS1.1 and CS1.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.

The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

The inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine 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 water flow to ensure they are indicative of leakage from the RCS.

This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; 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 United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

Containment High Range Radiation Monitors RE-8060/8061 [2RE-8925-1/8925-2] are the site-specific radiation monitors that would be indicative of possible core uncovery in the Refueling mode. The dose rate due to core shine when the top of the core becomes uncovered should result in dose rates > 10 R/hr.

Escalation of the emergency classification level would be via IC CG1 or AG1. Reference(s): 1. OP-1203.039 Excess RCS Leakage 2. OP-2203.016 Excess RCS Leakage

3. OP-2105.003 Reactor Vessel Level Monitoring System Operations 4. 1SAR Table 7-11 5. 2SAR 12.1.4.2

6. NEI 99-01 CS1 to 0CAN031801 Page 79 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting fuel clad integrity with containment challenged EAL: CG1.1 General Emergency - UNIT 2 ONLY RVLMS Levels 1 through 7 indicate DRY AND Any Containment Challenge indication, Table 2C-2 Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required. Table 1[2]C-2 Containment Challenge Indications CONTAINMENT CLOSURE not established (Note 6) Containment hydrogen concentration > 3% UNPLANNED rise in containment pressure Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Definition(s): CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent. IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. to 0CAN031801 Page 80 of 240 Attachment 1 - Emergency Action Level Technical Bases Basis: When in service, the Unit 2 RVLMS can measure RCS level below the top of active fuel.

Level 7 DRY on this system is an indication of core uncovery.

This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS level cannot be restored, fuel damage is probable.

With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity.

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged. This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; 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 United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

to 0CAN031801 Page 81 of 240 Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. OP-1203.039 Excess RCS Leakage 2. OP-2203.016 Excess RCS Leakage

3. OP-2105.003 Reactor Vessel Level Monitoring System Operations

4. 1SAR Table 7-11

5. 2SAR 12.1.4.2 6. Unit 1 SAMG Figure III-1B 7. Unit 2 SAMG Phase 1 Instructions, Containment Flowchart

8. NEI 99-01 CG1 to 0CAN031801 Page 82 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting fuel clad integrity with containment challenged EAL: CG1.2 General Emergency RCS level cannot be monitored for 30 min. (Note 1) AND Core uncovery is indicated by any of the following: UNPLANNED rise in any Table 1[2]C-1 sump/tank level of sufficient magnitude to indicate core uncovery Containment High Range Radiation Monitor RE-8060/8061[2RE-8925-1/8925-2] reading > 10 R/hr Erratic Source Range Monitor indication AND Any Containment Challenge indication, Table 1[2]C-2 Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required. Table 1C-1 Unit 1 Sumps / Tanks Reactor Building Sump Reactor Drain Tank Aux. Building Equipment Drain Tank Aux. Building Sump Quench Tank to 0CAN031801 Page 83 of 240 Attachment 1 - Emergency Action Level Technical Bases Table 2C-1 Unit 2 Sumps / Tanks CNTMT Sump Reactor Drain Tank LRW Waste Tank (2T-20) Holdup Tank Aux. Building Sump Quench Tank Table 1[2]C-2 Containment Challenge Indications CONTAINMENT CLOSURE not established (Note 6) Containment hydrogen concentration > 3% UNPLANNED rise in containment pressure Mode Applicability: 5 - Cold Shutdown, 6 - Refueling Definition(s): CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

to 0CAN031801 Page 84 of 240 Attachment 1 - Emergency Action Level Technical Bases Basis:

When in service, the Unit 2 RVLMS can measure RCS level below the top of active fuel. Level 7 DRY on this system is an indication of core uncovery.

This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS level cannot be restored, fuel damage is probable.

With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity.

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged.

The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine 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 water flow to ensure they are indicative of leakage from the RCS. to 0CAN031801 Page 85 of 240 Attachment 1 - Emergency Action Level Technical Bases Containment High Range Radiation Monitor RE-8060/8061 [2RE-8925-1/8925-2] are the site-specific radiation monitors that would be indicative of possible core uncovery in the Refueling mode. The dose rate due to core shine when the top of the core becomes uncovered should result in dose rates > 10 R/hr.

This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; 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 United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Reference(s):

1. OP-1203.039 Excess RCS Leakage 2. OP-2203.016 Excess RCS Leakage

3. OP-2105.003 Reactor Vessel Level Monitoring System Operations

4. 1SAR Table 7-11

5. 2SAR 12.1.4.2

6. Unit 1 SAMG Figure III-1B 7. Unit 2 SAMG Phase 1 Instructions, Containment Flowchart 8. NEI 99-01 CG1

to 0CAN031801 Page 86 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Vital AC Power Initiating Condition: Loss of all but one AC power source to vital buses for 15 minutes or longer EAL: CU2.1 Unusual Event AC power capability, Table 1[2]C-3, to vital 4.16 KV buses A3[2A3] and A4[2A4] reduced to a single power source for 15 min. (Note 1) AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table 1C-3 Unit 1 AC Power Sources Offsite Startup Transformer No. 1 Startup Transformer No. 2 Unit Auxiliary Transformer (from 22 KV switchyard) Onsite DG1 DG2 AAC Gen to 0CAN031801 Page 87 of 240 Attachment 1 - Emergency Action Level Technical Bases Table 2C-3 Unit 2 AC Power Sources Offsite Startup Transformer No. 3 Startup Transformer No. 2 Unit Auxiliary Transformer (backfed from main transformer) Onsite 2DG1 2DG2 AAC Gen Mode Applicability:

5 - Cold Shutdown, 6 - Refueling, DEF - Defueled Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the 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 result in potential offsite exposures. Basis: This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.

When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the greater time available to restore another power source to service.

Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant. to 0CAN031801 Page 88 of 240 Attachment 1 - Emergency Action Level Technical Bases An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to a vital bus. Some examples of this condition are presented below. A loss of all offsite power with a concurrent failure of all but one vital power source. A loss of all offsite power and loss of all vital power sources with a single train of vital buses being back-fed from the unit main generator. A loss of vital power sources (e.g., onsite diesel generators) with a single train of vital buses being back-fed from an offsite power source.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.

The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2.

This EAL is the cold condition equivalent of the hot condition EAL SA1.1.

Reference(s): 1. 1SAR Figure 8-1 Station Single Line Diagram 2. OP-1202.007 Degraded Power

3. OP-1202.008 Blackout

4. OP-2104.037 Alternate AC Diesel Generator Operations

5. 2SAR Figure 8.3-1 Station Single Line Diagram 6. OP-2202.007 Loss of Off-Site Power 7. OP-2202.008 Station Blackout

8. OP-2107.006 Backfeed of Unit Auxiliary Transformer

9. NEI 99-01 CU2 to 0CAN031801 Page 89 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Vital AC Power Initiating Condition: Loss of all offsite and all onsite AC power to vital buses for 15 minutes or longer EAL: CA2.1 Alert Loss of all offsite and all onsite AC power to vital 4.16 KV buses A3[2A3] and A4[2A4] for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 5 - Cold Shutdown, 6 - Refueling, DEF - Defueled

Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the 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 result in potential offsite exposures. Basis:

Although the AAC may be considered available, it will not prevent declaration of this EAL unless it is powering a vital bus within the 15-minute time period of the EAL.

This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. Mitigative strategies using non-safety related power sources (FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in to 0CAN031801 Page 90 of 240 Attachment 1 - Emergency Action Level Technical Bases accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems.

When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the greater time available to restore a vital bus to service.

Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.

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 AS1.

This EAL is the cold condition equivalent of the hot condition EAL SS1.1.

Reference(s): 1. 1SAR Figure 8-1 Station Single Line Diagram 2. OP-1202.007 Degraded Power

3. OP-1202.008 Blackout

4. OP-2104.037 Alternate AC Diesel Generator Operations 5. 2SAR Figure 8.3-1 Station Single Line Diagram 6. OP-2202.007 Loss of Off-Site Power

7. OP-2202.008 Station Blackout

8. OP-2107.006 Backfeed of Unit Auxiliary Transformer

9. NEI 99-01 CU2 to 0CAN031801 Page 91 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: UNPLANNED rise in RCS temperature EAL: CU3.1 Unusual Event UNPLANNED rise in RCS temperature to > 200°F due to loss of decay heat removal capability Mode Applicability: 5 - Cold Shutdown, 6 - Refueling Definition(s): CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis: This IC addresses an UNPLANNED rise in RCS temperature above the Technical Specification cold shutdown temperature limit and represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to EAL CA3.1.

A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. This EAL involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications.

During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.

to 0CAN031801 Page 92 of 240 Attachment 1 - Emergency Action Level Technical Bases During an outage, the level in the reactor vessel will normally be maintained at or above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled. A loss of forced decay heat removal at lowered inventory may result in a rapid rise in reactor coolant temperature depending on the time after shutdown.

Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria. Reference(s): 1. Unit 1 and Unit 2 Technical Specifications Table 1.1-1 2. NEI 99-01 CU3 to 0CAN031801 Page 93 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: UNPLANNED rise in RCS temperature EAL: CU3.2 Unusual Event Loss of all RCS temperature and RCS level indication for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 5 - Cold Shutdown, 6 - Refueling Definition(s):

CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: This EAL addresses the inability to determine RCS temperature and level, and represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to EAL CA3.1.

This EAL reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.

to 0CAN031801 Page 94 of 240 Attachment 1 - Emergency Action Level Technical Bases Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.

Reference(s): 1. NEI 99-01 CU3

to 0CAN031801 Page 95 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: Inability to maintain plant in cold shutdown EAL: CA3.1 Alert UNPLANNED rise in RCS temperature to > 200°F for > Table 1[2]C-4 duration (Note 1) OR UNPLANNED RCS pressure rise > 10 psig due to a loss of RCS cooling (this EAL does not apply during water-solid plant conditions) Note 1: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table 1[2]C-4 RCS Heat-up Duration Thresholds RCS Status CONTAINMENT CLOSURE Status Heat-up Duration Intact (but not lowered inventory) N/A 60 min.* Not intact OR lowered inventory established 20 min.* not established 0 min.

• If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable. Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Definition(s): CONTAINMENT CLOSURE - The action to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

to 0CAN031801 Page 96 of 240 Attachment 1 - Emergency Action Level Technical Bases As applied to ANO, CONTAINMENT CLOSURE must be capable of being set within 30 minutes. CONTAINMENT CLOSURE is set when the penetrations are isolated by manual or automatic isolation valve, blind flange, or equivalent.

UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on the RCS pressure rise criteria when the RCS is intact in Mode 5 or based on time to boil data when in Mode 6 or the RCS is not intact in Mode 5.

This EAL addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant.

A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. The RCS Heat-up Duration Thresholds table addresses a rise in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS inventory is reduced (e.g., lowered inventory operation). The 20-minute criterion was included to allow time for operator action to address the temperature rise. The RCS Heat-up Duration Thresholds table also addresses a rise in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature rise without a substantial degradation in plant safety.

Finally, in the case where there is a rise in RCS temperature, the RCS is not intact or is at lowered inventory and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel.

The RCS pressure rise threshold provides a pressure-based indication of RCS heat-up in the absence of RCS temperature monitoring capability.

Escalation of the emergency classification level would be via IC CS1 or AS1. Reference(s): 1. Unit 1 and Unit 2 Technical Specifications Table 1.1-1 2. NEI 99-01 CA3 to 0CAN031801 Page 97 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 4 - Loss of Vital DC Power Initiating Condition: Loss of Vital DC power for 15 minutes or longer EAL: CU4.1 Unusual Event Indicated voltage is < 105 VDC on vital 125 VDC buses for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 5 - Cold Shutdown, 6 - Refueling Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the 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 result in potential offsite exposures. Basis: Unit 1 batteries D06 and D07 and Unit 2 batteries 2D11 and 2D12 contain 58 cells each with a minimum cell voltage of 1.81 V or 105 VDC. This IC addresses a loss of vital DC power which compromises the ability to monitor and control 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 system temperatures and pressures are lower; these conditions raise the time available to restore a vital DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant.

to 0CAN031801 Page 98 of 240 Attachment 1 - Emergency Action Level Technical Bases As used in this EAL, "required" means the vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency 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 IC CA1 or CA3, or an IC in Recognition Category A.

This EAL is the cold condition equivalent of the hot condition EAL SS2.1. Reference(s): 1. 1SAR 8.3.2.1.1 Batteries 2. 2SAR 8.3.2.1.1 Batteries 3. NEI 99-01 CU4 to 0CAN031801 Page 99 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 5 - Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL: CU5.1 Unusual Event Loss of all Table 1[2]C-5 onsite communication methods OR Loss of all Table 1[2]C-5 State and local agency communication methods OR Loss of all Table 1[2]C-5 NRC communication methods Table 1[2]C-5 Communication Methods System Onsite ORO NRC Station radio system X ANO plant phone system X Gaitronics X Telephone Systems: Commercial Microwave Satellite VOIP X X INFORM Notification System X Emergency Notification System (ENS) X Mode Applicability: 5 - Cold Shutdown, 6 - Refueling, DEF - Defueled Definition(s): None to 0CAN031801 Page 100 of 240 Attachment 1 - Emergency Action Level Technical Bases Basis: This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to State and local agencies and the NRC.

This IC should be assessed only when extraordinary means are being utilized to make communications 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 to offsite locations, etc.).

The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations.

The second EAL condition addresses a total loss of the communications methods used to notify all State and local agencies of an emergency declaration. The State and local agencies referred to here are the Arkansas Department of Health, Arkansas Department of Emergency Management, Pope, Yell, Johnson, and Logan County offsite agencies.

The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

This EAL is the cold condition equivalent of the hot condition EAL SU7.1. Reference(s): 1. OP-1903.062 Communications System Operating Procedure 2. NEI 99-01 CU5 to 0CAN031801 Page 101 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 6 - Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode EAL: CA6.1 Alert The occurrence of any Table 1[2]C-6 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER: Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 10, 11) Note 10: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted. Note 11: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted. Table 1[2]C-6 Hazardous Events Seismic event (earthquake) Internal or external FLOODING event High winds or tornado strike FIRE EXPLOSION Other events with similar hazard characteristics as determined by the Shift Manager to 0CAN031801 Page 102 of 240 Attachment 1 - Emergency Action Level Technical Bases Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Definition(s): EXPLOSION - A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines 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 drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred 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 installed equipment 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/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the 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 result in potential offsite exposures.

VISIBLE DAMAGE - Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train. Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM to 0CAN031801 Page 103 of 240 Attachment 1 - Emergency Action Level Technical Bases train is from the same SAFETY SYSTEM that has indications of degraded performance; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.

Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

Escalation of the emergency classification level would be via IC CS1 or AS1.

This EAL is the cold condition equivalent of the hot condition EAL SA9.1. Reference(s): 1. EP FAQ 2016-002 2. NEI 99-01 CA6 to 0CAN031801 Page 104 of 240 Attachment 1 - Emergency Action Level Technical Bases Category E - Independent Spent Fuel Storage Installation (ISFSI) EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

An independent spent fuel storage installation (ISFSI) is a complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. A significant amount of the radioactive material contained within a canister must escape its packaging and enter the biosphere for there to be a significant environmental effect resulting from an accident involving the dry storage of spent nuclear fuel.

An Unusual Event is declared on the basis of the occurrence of an event of sufficient magnitude that a loaded cask CONFINEMENT BOUNDARY is damaged or violated.

The ANO ISFSI is located wholly within the plant PROTECTED AREA. Therefore any security event related to the ISFSI is classified under Category H1 security event related EALs.

to 0CAN031801 Page 105 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: E - ISFSI Subcategory: Confinement Boundary Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY EAL: EU1.1 Unusual Event Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask (VSC-24 VCC or HI-STORM overpack) > any Table 1[2]E-1 value Table 1[2]E-1 ISFSI Dose Rates VSC-24 VCC HI-STORM 200 mrem/hr on the sides 400 mrem/hr on the top 700 mrem/hr at the air inlet 200 mrem/hr at the air outlet 60 mrem/hr (gamma + neutron) on the top or outlet vent 600 mrem/hr (gamma + neutron on the side of the side of the overpack (excluding inlet and outlet ducts)

Mode Applicability: All Definition(s): CONFINEMENT BOUNDARY - The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the ANO ISFSI, the Confinement Boundary is comprised of either the Multi-assembly Sealed Basket (MSB) (SNC System) or Multi-Purpose Canister (MPC) (Holtec System).

INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) - A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.

Basis: This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the to 0CAN031801 Page 106 of 240 Attachment 1 - Emergency Action Level Technical Bases creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.

The existence of "damage" is determined by radiological survey. The specified EAL threshold values correspond to 2 times the cask technical specification values (ref. 1, 2). The technical specification (licensing bases document) multiple of "2 times", which is also used in Recognition Category A IC AU1, is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.

Security-related events for ISFSIs are covered under ICs HU1 and HA1.

Reference(s):

1. Certificate of Compliance Appendix A Technical Specifications for the HI-STORM 100 Cask System Section 5.7.4 2. VSC-24 Storage Cask Final Safety Analysis Report Section 1.2.4 Maximum External Surface Dose Rate 3. NEI 99-01 E-HU1 to 0CAN031801 Page 107 of 240 Attachment 1 - Emergency Action Level Technical Bases Category F - Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature > 200°F); EALs in this category are applicable only in one or more hot operating modes.

EALs in this category represent threats to the defense in depth design concept that precludes the release of highly radioactive fission products to the environment. This concept relies on multiple physical barriers any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. The primary fission product barriers are:

A. Fuel Clad Barrier (FCB): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets. B. Reactor Coolant System Barrier (RCB): The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves. C. Containment Barrier (CNB): The Containment Barrier includes the Reactor Building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the Reactor Building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (ECL) from an Alert to a Site Area Emergency or a General Emergency.

The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-1. "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. "Loss" means the barrier no longer assures containment of radioactive materials. "Potential Loss" means integrity of the barrier is threatened and could be lost if conditions continue to degrade. The number of barriers that are lost or potentially lost and the following criteria determine the appropriate emergency classification level: Alert: Any loss or any potential loss of either Fuel Clad or RCS Barrier Site Area Emergency: Loss or potential loss of any two barriers General Emergency: Loss of any two barriers and loss or potential loss of third barrier to 0CAN031801 Page 108 of 240 Attachment 1 - Emergency Action Level Technical Bases The logic used for emergency classification based on fission product barrier monitoring should reflect the following considerations: The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Containment Barrier. Unusual Event ICs associated with RCS and Fuel Clad Barriers are addressed under System Malfunction ICs. For accident conditions involving a radiological release, evaluation of the fission product barrier thresholds will need to be performed in conjunction with dose assessments to ensure correct and timely escalation of the emergency classification. For example, an evaluation of the fission product barrier thresholds may result in a Site Area Emergency classification while a dose assessment may indicate that an EAL for General Emergency IC AG1 has been exceeded. The fission product barrier thresholds specified within a scheme reflect plant-specific ANO design and operating characteristics. As used in this category, the term RCS leakage encompasses not just those types defined in Technical Specifications but also includes the loss of RCS mass to any location - inside the containment, an interfacing system, or outside of the 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 far present conditions are from meeting a threshold that would require a General Emergency declaration. For example, if the Fuel Clad and RCS fission product barriers were both lost, then there should be frequent assessments of containment radioactive inventory and integrity. Alternatively, if both the Fuel Clad and RCS fission product barriers were potentially lost, the Emergency Director would have more assurance that there was no immediate need to escalate to a General Emergency.

to 0CAN031801 Page 109 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: F - Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of either Fuel Clad or RCS EAL: FA1.1 Alert Any loss or any potential loss of either Fuel Clad or RCS barrier (Table 1[2]F-1) Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s): None Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table 1[2]F-1 lists the fission product barrier thresholds, bases and references. At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavily than the Containment barrier. Unlike the Containment barrier, loss or potential loss of either the Fuel Clad or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability. Note that the loss or potential loss of Containment barrier in combination with loss or potential loss of either Fuel Clad or RCS barrier results in declaration of a Site Area Emergency under EAL FS1.1. Reference(s): 1. NEI 99-01 FA1 to 0CAN031801 Page 110 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: F - Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss or potential loss of any two barriers EAL: FS1.1 Site Area Emergency Loss or potential loss of any two barriers (Table 1[2]F-1) Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s): IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table 1[2]F-1 lists the fission product barrier thresholds, bases and references.

At the Site Area Emergency classification level, each barrier is weighted equally. A Site Area Emergency is therefore appropriate for any combination of the following conditions:

One barrier loss and a second barrier loss (i.e., loss - loss) One barrier loss and a second barrier potential loss (i.e., loss - potential loss) One barrier potential loss and a second barrier potential loss (i.e., potential loss - potential loss)

At the Site Area Emergency classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a General Emergency is important. For example, the existence of Fuel Clad and RCS Barrier loss thresholds in addition to offsite dose assessments would require continual assessments of radioactive inventory and Containment integrity in anticipation of reaching a General Emergency classification. Alternatively, if both Fuel Clad and RCS potential loss thresholds existed, the Emergency Director would have greater assurance that escalation to a General Emergency is less IMMINENT. Reference(s): 1. NEI 99-01 FS1 to 0CAN031801 Page 111 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: F - Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss of any two barriers and loss or potential loss of third barrier EAL: FG1.1 General Emergency Loss of any two barriers AND Loss or potential loss of the third barrier (Table 1[2]F-1) Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s): None Basis:

Fuel Clad, RCS and Containment comprise the fission product barriers. Table 1[2]F-1 lists the fission product barrier thresholds, bases and references.

At the General Emergency classification level each barrier is weighted equally. A General Emergency 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 Barrier Reference(s): 1. NEI 99-01 FG1 to 0CAN031801 Page 112 of 240 Attachment 1 - Emergency Action Level Technical Bases Table 1[2]F-1 Fission Product Barrier Threshold Matrix & Bases Table 1[2]F-1 lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment). The table is structured so that each of the three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential Loss thresholds. The first column of the table (to the left of the Fuel Clad Loss column) lists the categories (types) of fission product barrier thresholds. The fission product barrier categories are:

A. RCS or S/G Tube Leakage B. Inadequate Heat removal C. Containment Radiation / RCS Activity D. Containment Integrity or Bypass E. Emergency Director Judgment Each category occupies a row in Table 1[2]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 or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word "None" is entered in the cell. Thresholds are assigned sequential numbers within each barrier column beginning with number one (ex., FCB1, FCB2-FCB9). If a cell in Table 1[2]F-1 contains more than one numbered threshold, each of the numbered thresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessary to exceed all of the thresholds in a category before declaring a barrier Loss/Potential Loss. Subdivision of Table 1[2]F-1 by category facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss thresholds. This structure promotes a systematic approach to assessing the classification status of the fission product barriers. When equipped with knowledge of plant conditions related to the fission product barriers, the EAL-user first scans down the category column of Table 1[2]F-1, locates the likely category and then reads across the fission product barrier Loss and Potential Loss thresholds in that category 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 the new category. If the EAL-user determines that any threshold has been exceeded, by definition, the barrier is lost 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 three fission product barriers to determine if other barrier thresholds in the category are lost or potentially lost. For example, if containment radiation is sufficiently high, a Loss of the Fuel to 0CAN031801 Page 113 of 240 Attachment 1 - Emergency Action Level Technical Bases Clad and RCS Barriers and a Potential Loss of the Containment Barrier can occur. Barrier Losses and Potential Losses are then applied to the algorithms given in EALs FG1.1, FS1.1, 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 each barrier, the bases are given according category Loss followed by category Potential Loss beginning with Category A, then B,-, E. to 0CAN031801 Page 114 of 240 Attachment 1 - Emergency Action Level Technical Bases Table 1[2]F-1 Fission Product Barrier Threshold Matrix Fuel Clad Barrier (FCB) Reactor Coolant System Barrier (RCB) Containment Barrier (CNB) Category Loss Potential Loss Loss Potential Loss Loss Potential Loss A RCS or S/G Tube Leakage None FCB1 RVLMS Levels 1 through 9 [1 through 7] indicate DRY RCB1 An automatic or manual ESAS [ESFAS] actuation required by EITHER: UNISOLABLE RCS leakage S/G tube RUPTURE RCB2 UNISOLABLE RCS leakage or S/G tube leakage > 50[44] gpm excluding normal reductions in RCS inventory (e.g., letdown, RCP seal leakoff) RCB3 Unit 1: PTS limits apply (RT14) AND RCS pressure and temperature are left of the NDTT/LTOP limit lines on EOP Figure 3 (Note 12) Unit 2: Uncontrolled RCS cooldown (50°F step change which is below 500°F from NOT) AND RCS pressure and temperature are to the left of line B (200 degrees MTS),

Standard Attachment 1, P-T Limits (Note 12) CNB1 A S/G that is leaking > 50[44] gpm (excluding normal reductions in RCS inventory) or that is RUPTURED is also FAULTED outside of containment None B Inadequate Heat Removal FCB2 CETs > 1200°F FCB3 CETs > 700°F FCB4 RCS heat removal cannot be established using steam generators AND HPI [Once Through] cooling initiated None RCB4 RCS heat removal cannot be established using steam generators AND HPI [Once Through] cooling initiated None CNB2 CETs > 1200°F AND Restoration procedures not effective within 15 min. (Note 1) C CTMT Radiation / RCS Activity FCB5 Containment High Range Radiation Monitor RE-8060/8061

[2RE-8925-1/ 8925-2]

> 750 [700] R/hr FCB6 Coolant activity

> 300 Ci/gm dose equivalent I-131 None RCB5 Containment High Range Radiation Monitor RE-8060/8061 [2RE-8925-1/ 8925-2] > 40[50] R/hr None None CNB3 Containment High Range Radiation Monitor RE-8060/8061 [2RE-8925-1/ 8925-2] > 10,000[12,000] R/hr to 0CAN031801 Page 115 of 240 Attachment 1 - Emergency Action Level Technical Bases Table 1[2]F-1 Fission Product Barrier Threshold Matrix Fuel Clad Barrier (FCB) Reactor Coolant System Barrier (RCB) Containment Barrier (CNB) Category Loss Potential Loss Loss Potential Loss Loss Potential Loss D CTMT Integrity or Bypass None None None None CNB4 Containment isolation is required AND EITHER: Containment integrity has been lost based on Emergency Director judgment UNISOLABLE pathway from Containment to the environment exists CNB5 Indications of RCS leakage outside of Containment CNB6 Containment pressure > 73.7 psia CNB7 Containment hydrogen concentration

> 3% CNB8 Containment pressure > 44.7 psia [23.3 psia] with < one full train of containment heat removal systems (Note 9) operating per design for 15 min. (Note 1) E Emergency Director Judgment FCB7 Any condition in the opinion of the Emergency Director that indicates loss of the Fuel Clad barrier FCB8 Any condition in the opinion of the Emergency Director that indicates potential loss of the Fuel Clad barrier RCB6 Any condition in the opinion of the Emergency Director that indicates loss of the RCS barrier RCB7 Any condition in the opinion of the Emergency Director that indicates potential loss of the RCS barrier CNB9 Any condition in the opinion of the Emergency Director that indicates loss of the Containment barrier CNB10 Any condition in the opinion of the Emergency Director that indicates potential loss of the Containment barrier to 0CAN031801 Page 116 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: A - RCS or S/G Tube Leakage Degradation Threat: Loss Threshold: None

to 0CAN031801 Page 117 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: A - RCS or S/G Tube Leakage Degradation Threat: Potential Loss Threshold: FCB1 RVLMS Levels 1 through 9[1 through 7] indicate DRY Definition(s): None Basis: This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.

There is no Fuel Clad Barrier Loss threshold associated with RCS or S/G Tube Leakage.

Reference(s): 1. ULD SYS-24 Unit 1 Inadequate Core Cooling System 2. Calculation 84-EQ-0080-02 Loop Error Analysis for Reactor Vessel Level Monitoring System 3. ULD SYS-24 Unit 2 Inadequate Core Cooling Monitoring System

4. Calculation 90-E-0116-01 Unit 2 EOP Setpoint Document, Setpoint R.3 5. NEI 99-01 RCS or SG Tube Leakage Potential Loss 1.A to 0CAN031801 Page 118 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: B - Inadequate Heat Removal Degradation Threat: Loss Threshold: FCB2 CETs > 1200°F Definition(s): None Basis: This reading indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant. Reference(s): 1. NEI 99-01 Inadequate Heat Removal Loss 2.A to 0CAN031801 Page 119 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: B - Inadequate Heat Removal Degradation Threat: Potential Loss Threshold: FCB3 CETs > 700°F Definition(s): None Basis: This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage. Reference(s): 1. NEI 99-01 Inadequate Heat Removal Potential Loss 2.A to 0CAN031801 Page 120 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: B - Inadequate Heat Removal Degradation Threat: Potential Loss Threshold: FCB4 RCS heat removal cannot be established using steam generators AND HPI[Once Through] cooling initiated Definition(s): None Basis: This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the Fuel Clad Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using this threshold is not warranted.

In combination with Potential Loss RCB4, meeting this threshold results in a Site Area Emergency.

Reference(s): 1. OP-1202.004 Overheating 2. OP-1202.013 Figure 4, Core Exit Thermocouple for Inadequate Core Cooling

3. OP-2202.006 Loss of Feedwater 4. OP-2202.009 Functional Recovery, Safety Function Status Check 5 5. NEI 99-01 Inadequate Heat Removal Potential Loss 2.B to 0CAN031801 Page 121 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: C - CTMT Radiation / RCS Activity Degradation Threat: Loss Threshold: FCB5 Containment High Range Radiation Monitor RE-8060/8061[2RE-8925-1/8925-2] > 750[700] R/hr Definition(s): None Basis:

The containment radiation monitor reading (768[682] R/hr rounded to 750[700] R/hr for readability) corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300 µCi/gm dose equivalent I-131.

Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to 1.49[1.13]% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold RCB5 since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier.

Note that a combination of the two monitor readings appropriately escalates the ECL to a Site Area Emergency. There is no Potential Loss threshold associated with CTMT Radiation/RCS Activity.

Reference(s):

1. EP-CALC-ANO-1702 Containment High Range Radiation Monitor EAL Values 2. NEI 99-01 RCS Activity/Containment Radiation FC Loss 3.A to 0CAN031801 Page 122 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: C - CTMT Radiation / RCS Activity Degradation Threat: Loss Threshold: FCB6 Coolant activity > 300 µCi/gm dose equivalent I-131 Definition(s): None Basis: This threshold indicates that RCS radioactivity concentration is greater than 300 µCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample-related threshold is included as a backup to other indications.

There is no Potential Loss threshold associated with CTMT Radiation/RCS Activity.

Reference(s): 1. NEI 99-01 RCS Activity/Containment Radiation Fuel Clad Loss 3.B

to 0CAN031801 Page 123 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: C - CTMT Radiation / RCS Activity Degradation Threat: Potential Loss Threshold: None

to 0CAN031801 Page 124 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: D - CTMT Integrity or Bypass Degradation Threat: Loss Threshold: None

to 0CAN031801 Page 125 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: D - CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold: None

to 0CAN031801 Page 126 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: E - Emergency Director Judgment Degradation Threat: Loss Threshold: FCB7 Any condition in the opinion of the Emergency Director that indicates loss of the Fuel Clad barrier Definition(s): None Basis:

This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is lost.

Reference(s):

1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A to 0CAN031801 Page 127 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: E - Emergency Director Judgment Degradation Threat: Potential Loss Threshold: FCB8 Any condition in the opinion of the Emergency Director that indicates potential loss of the Fuel Clad barrier Definition(s): None Basis:

This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Reference(s): 1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A

to 0CAN031801 Page 128 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: A - RCS or S/G Tube Leakage Degradation Threat: Loss Threshold: RCB1 An automatic or manual ESAS[ESFAS] actuation required by EITHER: UNISOLABLE RCS leakage S/G tube RUPTURE Definition(s): FAULTED - The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

RUPTURED - The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection (automatic or manual).

UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally. Basis: Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

This threshold is based on an UNISOLABLE RCS leak of sufficient size to require an automatic or manual actuation of the Emergency Core Cooling System (ECCS). This condition clearly represents a loss of the RCS Barrier.

This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system.

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 a safety injection is considered to be RUPTURED. If a RUPTURED steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold CNB1 will also be met. to 0CAN031801 Page 129 of 240 Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. OP-1202.010 ESAS 2. OP-2202.003 Loss of Coolant Accident

3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Loss 1.A to 0CAN031801 Page 130 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: A - RCS or S/G Tube Leakage Degradation Threat: Potential Loss Threshold: RCB2 UNISOLABLE RCS leakage or S/G tube leakage > 50[44] gpm excluding normal reductions in RCS inventory (e.g., letdown, RCP seal leakoff) Definition(s): FAULTED - The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally. Basis: Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

This threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizer level within specified limits by operation of a normally used makeup [charging]

pump, but an ESAS [ESFAS] actuation has not occurred. This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system.

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 declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold CNB1 will also be met.

Reference(s): 1. 1SAR 9.1 Makeup and Purification System 2. 2SAR 9.3.4 Chemical and Volume Control System

3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.A to 0CAN031801 Page 131 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: A - RCS or S/G Tube Leakage Degradation Threat: Potential Loss Threshold: RCB3 Unit 1: PTS limits apply (RT14) AND RCS pressure and temperature are left of the NDTT/LTOP limit lines, on EOP Figure 3 (Note 12) Unit 2: Uncontrolled RCS cooldown (50°F step change which is below 500°F from NOT) AND RCS pressure and temperature are to the left of line B (200 degrees MTS), Standard Attachment 1, P-T Limits (Note 12) Note 12: Once PTS limits are first invoked, if RCS temperature and pressure are not brought within the limits within 15 minutes, this threshold is met and an immediate declaration is warranted. This threshold is met immediately upon exceeding the limits after this initial 15 minute period until PTS limits no longer apply. Definition(s): None Basis:

This condition indicates an extreme challenge to the integrity of the RCS pressure boundary due to pressurized thermal shock - a transient that causes rapid RCS cooldown while the RCS is in Mode 3 or higher (i.e., hot and pressurized).

to 0CAN031801 Page 132 of 240 Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. OP-1202.012 Repetitive Task 14 Control RCS Pressure 2. OP-1202.013 EOP Figures, Figure 3 RCS Pressure vs Temperature Limits

3. OP-1202.011 HPI Cooldown

4. Calculation No: 90-E-0116-01 ANO- EOP Setpoint Basis Document OP Setpoint P.2, RCS Pressure-Temperature 5. OP-2202.010 Standard Attachments, Attachment 1, P-T Limits 6. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.B to 0CAN031801 Page 133 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: B - Inadequate Heat Removal Degradation Threat: Loss Threshold: None

to 0CAN031801 Page 134 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: B - Inadequate Heat Removal Degradation Threat: Potential Loss Threshold: RCB4 RCS heat removal cannot be established using steam generators AND HPI[Once Through] cooling initiated Definition(s): None Basis: This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the RCS Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using this threshold is not warranted.

In combination with Potential Loss FCB4, meeting this threshold results in a Site Area Emergency.

This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and raise RCS pressure to the point where mass will be lost from the system.

There is no RCS barrier Loss threshold associated with Inadequate Heat Removal.

Reference(s): 1. OP-1202.004 Overheating 2. OP-1202.013 Figure 4, Core Exit Thermocouple for Inadequate Core Cooling

3. OP-2202.006 Loss of Feedwater 4. OP-2202.009 Functional Recovery, Safety Function Status Check 5 5. NEI 99-01 Inadequate Heat Removal RCS Potential Loss 2.B to 0CAN031801 Page 135 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: C - CTMT Radiation/ RCS Activity Degradation Threat: Loss Threshold: RCB5 Containment High Range Radiation Monitor RE-8060/8061[2RE-8925-1/8925-2] > 40[50] R/hr Definition(s): None Basis: NRC Information Notice 97-045, Supplement 1, identifies the potential for erratic indications from the high range radiation monitors (HRRMs) as a result of thermally induced currents (TIC) which may cause the HRRM to read falsely high (for approximately 15 minutes) on a rapid temperature rise, and fail low intermittently on a rapid temperature fall. Because of this phenomenon, any trends or alarms on the HRRM's should be validated by comparison to the containment low range/area radiation monitors and Air Monitoring Systems trends before actions are taken.

The containment radiation monitor reading (42.8[50.4] R/hr rounded to 40[50] R/hr for readability) corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold FCB5 since it indicates a loss of the RCS Barrier only.

There is no Potential Loss threshold associated with CTMT Radiation/RCS Activity.

Reference(s): 1. EP-CALC-ANO-1702 Containment High Range Radiation Monitor EAL Values 2. NEI 99-01 CMT Radiation / RCS Activity RCS Loss 3.A to 0CAN031801 Page 136 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: B - CTMT Radiation/ RCS Activity Degradation Threat: Potential Loss Threshold: None

to 0CAN031801 Page 137 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: D - CTMT Integrity or Bypass Degradation Threat: Loss Threshold: None

to 0CAN031801 Page 138 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: D - CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold: None

to 0CAN031801 Page 139 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: E - Emergency Director Judgment Degradation Threat: Loss Threshold: RCB6 Any condition in the opinion of the Emergency Director that indicates loss of the RCS barrier Definition(s): None Basis: This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is lost. Reference(s): 1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A to 0CAN031801 Page 140 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: E - Emergency Director Judgment Degradation Threat: Potential Loss Threshold: RCB7 Any condition in the opinion of the Emergency Director that indicates potential loss of the RCS barrier Definition(s): None Basis:

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Reference(s): 1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6.A

to 0CAN031801 Page 141 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: A - RCS or S/G Tube Leakage Degradation Threat: Loss Threshold: CNB1 A S/G that is leaking > 50[44] gpm (excluding normal reductions in RCS inventory) or that is RUPTURED is also FAULTED outside of containment Definition(s): FAULTED - The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

RUPTURED - The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection (automatic or manual).

Basis: This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outside of containment. The condition of the SG, whether leaking or RUPTURED, is determined in accordance with the thresholds for RCS Barrier Potential Loss RCB2 and Loss RCB1, respectively. This condition represents a bypass of the containment barrier. FAULTED is a defined term within the NEI 99-01 methodology; this determination is not necessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steam generator is dropping uncontrollably (part of the FAULTED definition) and the FAULTED steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FAULTED for emergency classification purposes. The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification. Steam releases of this size are readily observable with normal Control Room indications. The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU4 for the fuel clad barrier (i.e., RCS activity values) and IC SU5 for the RCS barrier (i.e., RCS leak rate values). This threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown the plant. These types of conditions will result in a significant and sustained release of radioactive steam to the environment (and are thus similar to a FAULTED condition). The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss of containment. to 0CAN031801 Page 142 of 240 Attachment 1 - Emergency Action Level Technical Bases Steam releases associated with the expected operation of a SG power operated relief valve or safety relief valve do not meet the intent of this threshold. Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown. Steam releases associated with the unexpected operation of a valve (e.g., a stuck-open safety valve) do meet this threshold. Following a SG tube leak or rupture, there may be minor radiological releases through a secondary-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 but should be evaluated using the Recognition Category A ICs. The ECLs resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below.

Affected SG is FAULTED Outside of Containment? P-to-S Leak Rate Yes No Less than or equal to 25 gpm No classification No classification Greater than 25 gpm Unusual Event per SU5.1Unusual Event per SU5.1Greater than 50[44] gpm (RCS Barrier Potential Loss) Site Area Emergency per FS1.1 Alert per FA1.1 Requires an automatic or manual ESAS[ESFAS] actuation (RCS Barrier Loss) Site Area Emergency per FS1.1 Alert per FA1.1 There is no Potential Loss threshold associated with RCS or S/G Tube Leakage. Reference(s): 1. NEI 99-01 RCS or SG Tube Leakage Containment Loss 1.A to 0CAN031801 Page 143 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: A - RCS or S/G Tube Leakage Degradation Threat: Potential Loss Threshold: None

to 0CAN031801 Page 144 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: B - Inadequate Heat Removal Degradation Threat: Loss Threshold: None

to 0CAN031801 Page 145 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: B - Inadequate Heat Removal Degradation Threat: Potential Loss Threshold: CNB2 CETs > 1200°F AND Restoration procedures not effective within 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Definition(s): IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: The restoration procedure is considered "effective" if core exit thermocouple readings are dropping and/or if reactor vessel level is rising. Whether or not the procedure(s) will be effective should be apparent within 15 minutes. The Emergency Director should escalate the emergency classification level as soon as it is determined that the procedure(s) will not be effective.

This condition represents an IMMINENT core melt sequence which, if not corrected, could lead to vessel failure and an increased potential for containment failure. For this condition to occur, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. If implementation of a procedure(s) to restore adequate core cooling is not effective (successful) within 15 minutes, it is assumed that the event trajectory will likely lead to core melting and a subsequent challenge of the Containment Barrier.

Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation in a significant fraction of core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence.

Reference(s): 1. NEI 99-01 Inadequate Heat Removal Containment Potential Loss 2.A to 0CAN031801 Page 146 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: C - CTMT Radiation/RCS Activity Degradation Threat: Loss Threshold: None

to 0CAN031801 Page 147 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: C - CTMT Radiation/RCS Activity Degradation Threat: Potential Loss Threshold: CNB3 Containment High Range Radiation Monitor RE-8060/8061[2RE-8925-1/8925-2] > 10,000[12,000] R/hr Definition(s): None Basis:

The containment radiation monitor reading (10,300[12,100] R/hr rounded to 10,000[12,000] R/hr for readability) corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds. NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the ECL to a General Emergency.

There is no Loss threshold associated with CTMT Radiation/RCS Activity.

Reference(s): 1. EP-CALC-ANO-1702 Containment High Range Radiation Monitor EAL Values 2. NEI 99-01 CTMT Radiation / RCS Activity Containment Potential Loss 3.A to 0CAN031801 Page 148 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: D - CTMT Integrity or Bypass Degradation Threat: Loss Threshold: CNB4 Containment isolation is required AND EITHER: Containment integrity has been lost based on Emergency Director judgment UNISOLABLE pathway from Containment to the environment exists Definition(s):

UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally. Basis: Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold CNB1.

These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. Users are reminded that there may be accident and release conditions that simultaneously meet both bulleted thresholds.

First Threshold - Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage (or sometimes referred 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 integrity condition may (or may not) be accompanied by a noticeable drop in containment pressure. Recognizing the inherent difficulties in determining a containment leak rate during accident conditions, it is expected that the Emergency Director will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g., containment pressure, readings on radiation monitors outside containment, operating status of containment pressure control equipment, etc.).

to 0CAN031801 Page 149 of 240 Attachment 1 - Emergency Action Level Technical Bases Refer to the middle piping run of Figure 1. Two simplified examples are provided. One is leakage from a penetration and the other is leakage from an in-service system valve. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure.

Another example would be a loss or potential loss of the RCS barrier, and the simultaneous occurrence of two FAULTED locations on a steam generator where one fault is located inside containment (e.g., on a steam or feedwater line) and the other outside of containment. In this case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.

Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category A ICs.

Second Threshold - Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment. As used here, the term "environment" includes the atmosphere of a room or area, outside the containment, 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 variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure. Refer to the top piping run of Figure 1. In this simplified example, the inboard and outboard isolation valves remained open after a containment isolation was required (i.e., containment isolation was not successful). There is now an UNISOLABLE pathway from the containment to the environment.

The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.

Leakage between two interfacing liquid systems, by itself, does not meet this threshold.

Refer to the bottom piping run of Figure 1. In this simplified example, leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building. The radioactivity would be detected by the Process Monitor. If there is no leakage from the closed water cooling system to the Auxiliary Building, then no threshold has been met. If the pump developed a leak that allowed steam/water to enter the Auxiliary Building, then second threshold would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause the first threshold to be met as well.

to 0CAN031801 Page 150 of 240 Attachment 1 - Emergency Action Level Technical Bases Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable containment leakage through various penetrations or system components. Minor releases may also occur if a containment isolation valve(s) fails to close but the containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category A ICs.

Reference(s): 1. NEI 99-01 CTMT Integrity or Bypass Containment Loss 4.A

to 0CAN031801 Page 151 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: D - CTMT Integrity or Bypass Degradation Threat: Loss Threshold: CNB5 Indications of RCS leakage outside of Containment Definition(s): None Basis: To ensure proper escalation of the emergency classification, the RCS leakage outside of containment must be related to the mass loss that is causing the RCS Loss RCB1 and/or Potential Loss RCB2 threshold to be met.

The status of the containment barrier during an event involving steam generator tube leakage is assessed using Containment Loss Threshold CNB1. Containment sump, temperature, pressure and/or radiation levels will rise if reactor coolant mass is leaking into the containment. If these parameters have not risen, then the reactor coolant mass may be leaking outside of containment (i.e., a containment bypass sequence).

Rises in sump, temperature, pressure, flow and/or radiation level readings outside of the containment may indicate that the RCS mass is being lost outside of containment. Unexpected elevated readings and alarms on radiation monitors with detectors outside containment should be corroborated with other available indications to confirm that the source is 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 rise 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 1. In this simplified example, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure and cause threshold CNB4 to be met as well.

Reference(s): 1. NEI 99-01 CTMT Integrity or Bypass Containment Loss 4.B to 0CAN031801 Page 152 of 240 Attachment 1 - Emergency Action Level Technical Bases Figure 1: Containment Integrity or Bypass Examples

Open valveOpen valveOpen valve Open valve Open valve Open valvePenetrationDamperDamper Interface leakage RCP Seal Cooling Inside Reactor Building Auxiliary Building1st Threshold - Airborne 1st Threshold - Airborne release from penetration2nd Threshold - Airborne release from pathway2nd Threshold - RCS leakage outside ABAirborne MonitorArea MonitorProcess MonitorEffluent MonitorClosed CoolingPump to 0CAN031801 Page 153 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: D - CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold: CNB6 Containment pressure > 73.7 psia Definition(s): None Basis:

If containment pressure exceeds the design pressure, there exists a potential to lose the Containment Barrier. To reach this level, there must be an inadequate core cooling condition for an extended period of time; therefore, the RCS and Fuel Clad barriers would already be lost.

Thus, this threshold is a discriminator between a Site Area Emergency and General Emergency since there is now a potential to lose the third barrier. Reference(s): 1. 1SAR 1.4.43 Criterion 50 - Containment Design Basis 2. 2SAR Table 6.2-7 Principle Containment Design Parameters

3. NEI 99-01 CTMT Integrity or Bypass Containment Potential Loss 4.A to 0CAN031801 Page 154 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: D - CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold: CNB7 Containment hydrogen concentration > 3% Definition(s): None Basis: The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a potential loss of the Containment Barrier.

Reference(s): 1. Unit 1 SAMG Figure III-1B 2. Unit 2 SAMG Phase 1 Instructions, Containment Flowchart

3. NEI 99-01 CTMT Integrity or Bypass Containment Potential Loss 4.B to 0CAN031801 Page 155 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: D - CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold: CNB8 Containment pressure > 44.7 psia[23.3 psia] with < one full train of containment heat removal systems (Note 9) operating per design for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 9: One full train of containment heat removal systems consists of one train of RB [Containment] Spray and one train of RB [Containment] Cooling System. Definition(s): None Basis:

This threshold describes a condition where containment pressure is greater than the setpoint at 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-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. This threshold represents a potential loss of containment in that containment heat removal/depressurization systems (e.g., containment sprays but not including containment venting strategies) are either lost or performing in a degraded manner. Reference(s): 1. 1SAR 6.2 Reactor Building Spray System 2. 1SAR 6.3 Reactor Building Cooling System 3. OP-2202.003 Loss of Coolant Accident

4. OP-2202.010 Standard Attachments, Attachment 22

5. 2SAR 6.2.2 Containment Heat Removal Systems

6. 2SAR 7.3.1.1.11.2 Containment Spray System 7. NEI 99-01 CTMT Integrity or Bypass Containment Potential Loss 4.C to 0CAN031801 Page 156 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: E - Emergency Director Judgment Degradation Threat: Loss Threshold: CNB9 Any condition in the opinion of the Emergency Director that indicates loss of the Containment barrier Definition(s): None Basis:

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is lost.

Reference(s):

1. NEI 99-01 Emergency Director Judgment Containment Loss 6.A to 0CAN031801 Page 157 of 240 Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: E - Emergency Director Judgment Degradation Threat: Potential Loss Threshold: CNB10 Any condition in the opinion of the Emergency Director that indicates potential loss of the Containment barrier Definition(s): None Basis:

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Reference(s): 1. NEI 99-01 Emergency Director Judgment Containment Potential Loss 6.A

to 0CAN031801 Page 158 of 240 Attachment 1 - Emergency Action Level Technical Bases Category H - Hazards and Other Conditions Affecting Plant Safety EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

Hazards are non-plant, system-related events that can directly or indirectly affect plant operation, reactor plant safety or personnel safety. 1. Security Unauthorized entry attempts into the PROTECTED AREA, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant. 2. Seismic Event Natural events such as earthquakes have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety. 3. Natural or Technological Hazard Other natural and non-naturally occurring events that can cause damage to plant facilities include tornados, FLOODING, hazardous material releases and events restricting site access warranting classification. 4. Fire FIRES can pose significant hazards to personnel and reactor safety. Appropriate for classification are FIRES within the plant PROTECTED AREA or which may affect operability of equipment needed for safe shutdown. 5. Hazardous Gas Toxic, corrosive, asphyxiant or flammable gas leaks can affect normal plant operations or preclude access to plant areas required to safely shutdown the plant. 6. Control Room Evacuation Events that are indicative of loss of Control Room habitability. If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities. 7. Emergency Director Judgment The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification.

While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the Emergency Director the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Director judgment. to 0CAN031801 Page 159 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL: HU1.1 Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by ANO Security Shift Supervision OR Notification of a credible security threat directed at the site OR A validated notification from the NRC providing information of an aircraft threat Mode Applicability:

All Definition(s): HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station.

HOSTILE ACTION - An act toward ANO or its personnel that includes the use of violent force 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 other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on ANO. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA).

OWNER CONTROLLED AREA - For the purposes of classification this is the Security Owner Controlled Area (SOCA). The SOCA is demarcated as a Vehicle Barrier System (VBS) and a detection fence on the outside and a delay fence on the inside of the passive and active barriers. The SOCA is the area inside the SOCA VBS up to the PROTECTED AREA fence line. Access to this area is controlled by the SOCA Personnel Access Control Point. PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

to 0CAN031801 Page 160 of 240 Attachment 1 - Emergency Action Level Technical Bases PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the 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 result in potential offsite exposures.

SECURITY CONDITION - Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a HOSTILE ACTION. Basis: This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR 73.71 or 10 CFR 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1 and HS1.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].

The first threshold references the Security Shift Supervision because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR 2.39 information.

The second threshold addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with OP-1203.048 Security Event .

to 0CAN031801 Page 161 of 240 Attachment 1 - Emergency Action Level Technical Bases The third threshold addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with 11-S 1 Security Contingency Events (ref. 2).

Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Security Plan for ANO (ref. 1).

Escalation of the emergency classification level would be via IC HA1. Reference(s): 1. ANO Security Plan 2. OP-1203.048 Security Event 3. NEI 99-01 HU1 to 0CAN031801 Page 162 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL: HA1.1 Alert A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by ANO Security Shift Supervision OR A validated notification from NRC of an aircraft attack threat within 30 min. of the site Mode Applicability: All Definition(s): HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION - An act toward ANO or its personnel that includes the use of violent force 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 other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on ANO. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA). HOSTILE FORCE - One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction. OWNER CONTROLLED AREA - For the purposes of classification this is the Security Owner Controlled Area (SOCA). The SOCA is demarcated as a Vehicle Barrier System (VBS) and a detection fence on the outside and a delay fence on the inside of the passive and active barriers. The SOCA is the area inside the SOCA VBS up to the PROTECTED AREA fence line. Access to this area is controlled by the SOCA Personnel Access Control Point. PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. to 0CAN031801 Page 163 of 240 Attachment 1 - Emergency Action Level Technical Bases Basis: This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact. Timely and accurate communications between Security Shift Supervision and the 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, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].

As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of Offsite Response Organizations (OROs), allowing them to be better prepared should it be necessary to consider further actions.

This EAL 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. Examples include 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 the requirements of 10 CFR 73.71 or 10 CFR 50.72.

The first threshold is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. The second threshold addresses the threat from the impact of an aircraft on the plant, and the anticipated 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 a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with OP-1203.048 Security Event (ref. 2).

The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD through the NRC.

In some cases, it may not be readily apparent if an aircraft impact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency. to 0CAN031801 Page 164 of 240 Attachment 1 - Emergency Action Level Technical Bases Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Security Plan for ANO (ref. 1).

Escalation of the emergency classification level would be via IC HS1. Reference(s): 1. ANO Security Plan 2. OP-1203.048 Security Event 3. NEI 99-01 HA1 to 0CAN031801 Page 165 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the PROTECTED AREA EAL: HS1.1 Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by ANO Security Shift Supervision Mode Applicability: All Definition(s):

HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station.

HOSTILE ACTION - An act toward ANO or its personnel that includes the use of violent force 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 other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on ANO. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA). HOSTILE FORCE - One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

OWNER CONTROLLED AREA - For the purposes of classification this is the Security Owner Controlled Area (SOCA). The SOCA is demarcated as a Vehicle Barrier System (VBS) and a detection fence on the outside and a delay fence on the inside of the passive and active barriers. The SOCA is the area inside the SOCA VBS up to the PROTECTED AREA fence line. Access to this area is controlled by the SOCA Personnel Access Control Point. PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. to 0CAN031801 Page 166 of 240 Attachment 1 - Emergency Action Level Technical Bases 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 plant equipment.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 1, 2).

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program]. As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Site Area Emergency declaration will mobilize Offsite Response Organization (ORO) resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions. This EAL 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. Examples include 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 the requirements of 10 CFR 73.71 or 10 CFR 50.72.

Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Security Plan for ANO (ref. 1). Reference(s): 1. ANO Security Plan 2. OP-1203.048 Security Event 3. NEI 99-01 HS1 to 0CAN031801 Page 167 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 2 - Seismic Event Initiating Condition: Seismic event greater than OBE levels EAL: HU2.1 Unusual Event Seismic event > OBE as indicated by annunciation of the 0.10 g acceleration alarm Mode Applicability: All Definition(s): None Basis: This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE). An earthquake greater than an OBE but less than a Design Basis Earthquake (DBE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the 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 and inspections, and fully understand any impacts, this event represents a potential degradation of the 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 a seismic event (e.g., lateral accelerations in excess of 0.1g). The Shift Manager or Emergency Director may seek external verification if deemed appropriate (e.g., a call to the USGS, check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9. Two strong motion triaxial accelerometers, ACS-8001 and ACS-8003, located at the base slab provide alarms to the Unit 1 control room via the seismic network control center, C529-NCC.

One alarm from C529-NCC is triggered when a setpoint of 0.01g has been exceeded. This alarm indicates that an earthquake has occurred and the seismic monitoring system is recording seismic data. Another alarm from C529-NCC is triggered when the pre-determined value of 0.1g, indicating the OBE has been exceeded (ref. 2, 3). to 0CAN031801 Page 168 of 240 Attachment 1 - Emergency Action Level Technical Bases To avoid inappropriate emergency classification resulting from spurious actuation of the seismic instrumentation or felt motion not attributable to seismic activity, an offsite agency (USGS, National Earthquake Information Center (NEIC)) can confirm that an earthquake has occurred in the area of the plant. Such confirmation should not, however, preclude a timely emergency declaration based on receipt of the OBE alarm. If requested, provide the analyst with the following ANO coordinates: 35º 18' 36" north latitude, 93º 13' 53" west longitude (ref. 4). Alternatively, near real-time seismic activity can be accessed via the NEIC website: Reference(s): 1. 1SAR 2.2.1 Location 2. 1SAR 2.7.2 Site Seismic Evaluation 3. 1SAR 2.7.6 Time-History Accelerograph

4. OP-1203.025 Natural Emergencies

5. OP-2203.008 Natural Emergencies

6. NEI 99-01 HU2 to 0CAN031801 Page 169 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.1 Unusual Event A tornado strike within the PROTECTED AREA Mode Applicability: All Definition(s): PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

This EAL addresses a tornado striking (touching down) within the PROTECTED AREA.

Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, S or C. If damage is confirmed visually or by other in-plant indications, the event may be escalated to an Alert under EAL CA6.1 or SA9.1.

A tornado striking (touching down) within the PROTECTED AREA warrants declaration of an Unusual Event regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base of a thunderstorm.

Reference(s): 1. OP-1203.025 Natural Emergencies 2. OP-2203.008 Natural Emergencies

3. NEI 99-01 HU3 to 0CAN031801 Page 170 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.2 Unusual Event Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component required by Technical Specifications for the current operating mode Mode Applicability: All Definition(s): FLOODING - A condition where water is entering a room or area faster than installed equipment 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/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the 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 result in potential offsite exposures.

Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

This EAL addresses FLOODING of a building room or area that results in operators isolating power 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 isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode. to 0CAN031801 Page 171 of 240 Attachment 1 - Emergency Action Level Technical Bases Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, S or C.

Refer to EAL CA6.1 or SA9.1 for internal FLOODING affecting more than one SAFETY SYSTEM train.

Reference(s): 1. OP-1203.025 Natural Emergencies 2. OP-2203.008 Natural Emergencies

3. NEI 99-01 HU3 to 0CAN031801 Page 172 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.3 Unusual Event Movement of personnel within the PROTECTED AREA is IMPEDED due to an event external to the PROTECTED AREA involving hazardous materials (e.g., an offsite chemical spill or toxic gas release) Mode Applicability: All Definition(s): IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures 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 (i.e., the security fence) and to which access is controlled.

Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

This EAL addresses a hazardous materials event originating at a location outside the PROTECTED AREA and of sufficient magnitude to IMPEDE the movement of personnel within the PROTECTED AREA. Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, S or C.

Reference(s): 1. NEI 99-01 HU3

to 0CAN031801 Page 173 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.4 Unusual Event A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing 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 or accidents. Mode Applicability: All Definition(s):

FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.

Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

This EAL addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal 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 blocking the access road.

This EAL is not intended to apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the FLOODING around the Cooper Station during the Midwest floods of 1993, or the FLOODING around Ft. Calhoun Station in 2011. Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, S or C.

Reference(s): 1. NEI 99-01 HU3

to 0CAN031801 Page 174 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.1 Unusual Event A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (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 alarm AND The FIRE is located within any Table 1[2]H-1 area Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table 1H-1 Unit 1 Fire Areas Reactor Building All elevations Auxiliary Building All elevations including: Penthouse/MSIV Room Exceptions: Boric Acid Mix Tank Room (Chem Add Area), 404' (157-B), EDG Exhaust Fan area on 386' (1-E and 2-E)

Turbine Building All elevations including: Pipechase under ICW Coolers, CRD Pump Pit/T-28 Room/Area under ICW Pumps Outside Areas Manholes adjacent to Startup #2 XFMR (MH-03/MH-04) Manholes adjacent to Intake Structure (MH-05/MH-06) Intake Structure (354' and 366') Diesel Fuel Vault Diesel Fuel Vault Pump Manholes (MH-09 and MH-10) to 0CAN031801 Page 175 of 240 Attachment 1 - Emergency Action Level Technical Bases Table 2H-1 Unit 2 Fire Areas Reactor Building All elevations Auxiliary Building All elevations including: Aux Extension Turbine Building All elevations Outside Areas Intake Structure (354' and 366') Concrete Manhole East, NE of intake (2MH-01) Concrete Manhole East of Turbine Building next to train bay (2MH-03) Diesel Fuel Vault Diesel Fuel Vault Pump Manholes (MH-09 and MH-10) Mode Applicability: All

Definition(s): FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc. to 0CAN031801 Page 176 of 240 Attachment 1 - Emergency Action Level Technical Bases 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 at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SA9.1.

The 15 minute requirement begins with a credible notification that a FIRE is occurring, or receipt of multiple VALID fire detection system alarms or field validation of a single fire alarm. The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field.

Table 1[2]H-1 Fire Areas are those areas that contain equipment necessary for safe operation and shutdown of the plant (ref. 1, 2).

Reference(s): 1. OP-1203.049 Fires in Areas Affecting Safe Shutdown 2. OP- 2203.049 Fires in Areas Affecting Safe Shutdown

3. NEI 99-01 HU4 to 0CAN031801 Page 177 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.2 Unusual Event Receipt of a single fire alarm (i.e., no other indications of a FIRE) AND The fire alarm is indicating a FIRE within any Table 1[2]H-1 area AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table 1H-1 Unit 1 Fire Areas Reactor Building All elevations Auxiliary Building All elevations including: Penthouse/MSIV Room Exceptions: Boric Acid Mix Tank Room (Chem Add Area), 404' (157-B), EDG Exhaust Fan area on 386' (1-E and 2-E) Turbine Building All elevations including: Pipechase under ICW Coolers, CRD Pump Pit/T-28 Room/Area under ICW Pumps Outside Areas Manholes adjacent to Startup #2 XFMR (MH-03/MH-04) Manholes adjacent to Intake Structure (MH-05/MH-06) Intake Structure (354' and 366')

Diesel Fuel Vault Diesel Fuel Vault Pump Manholes (MH-09 and MH-10) to 0CAN031801 Page 178 of 240 Attachment 1 - Emergency Action Level Technical Bases Table 2H-1 Unit 2 Fire Areas Reactor Building All elevations Auxiliary Building All elevations including: Aux Extension Turbine Building All elevations Outside Areas Intake Structure (354' and 366') Concrete Manhole East, NE of intake (2MH-01) Concrete Manhole East of Turbine Building next to train bay (2MH-03) Diesel Fuel Vault Diesel Fuel Vault Pump Manholes (MH-09 and MH-10) Mode Applicability: All Definition(s): FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.

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 take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed. to 0CAN031801 Page 179 of 240 Attachment 1 - Emergency Action Level Technical Bases A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.

If an actual FIRE is verified by a report from the field, then HU4.1 is immediately applicable, 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 spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted.

Basis-Related Fire Protection Requirements Criterion 3 of 10 CFR 50, Appendix A, states, in part:

"Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions."

In this respect, noncombustible and heat resistant materials are used wherever practical throughout the unit, particularly in locations such as the containment and Control Room. Fire detection and fighting systems of appropriate capacity and capability are provided and designed to minimize the adverse effects of fires on SSCs important to safety. Firefighting systems are designed to assure that the rupture or inadvertent operation of a fire system does not significantly impair the safety capability of these structures, systems, and components.

In addition, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train is employed. As used in HU4.2, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SA9.1.

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 it is not spurious, or by reports from the field. Actual field reports must be made within the 30-minute time limit or a classification must be made. If a fire is verified to be occurring by field report, classification shall be made based on EAL HU4.1, with the 15-minute requirement beginning with the verification of the fire by field report. Table 1[2]H-1 Fire Areas are those areas that contain equipment necessary for safe operation and shutdown of the plant (ref. 1, 2).

to 0CAN031801 Page 180 of 240 Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. OP-1203.049 Fires in Areas Affecting Safe Shutdown 2. OP- 2203.049 Fires in Areas Affecting Safe Shutdown

3. NEI 99-01 HU4 to 0CAN031801 Page 181 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.3 Unusual Event A FIRE within the PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: All Definition(s): FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled.

Basis:

This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.

In addition to a FIRE addressed by EAL HU4.1 or HU4.2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SA8.1.

Reference(s):

1. NEI 99-01 HU4 to 0CAN031801 Page 182 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.4 Unusual Event A FIRE within the PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Mode Applicability: All Definition(s):

FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.

PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. If a FIRE within the plant 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. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SA9.1. Reference(s): 1. NEI 99-01 HU4 to 0CAN031801 Page 183 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 5 - Hazardous Gas Initiating Condition: Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown EAL: HA5.1 Alert Release of a toxic, corrosive, asphyxiant or flammable gas into any Table 1[2]H-2 room or area AND Entry 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, then no emergency classification is warranted. Table 1H-2 Unit 1 Safe Operation & Shutdown Rooms/Areas Room/Area Mode A-4 Switchgear Room 3, 4 Upper North Electrical Penetration Room 3, 4 Lower South Electrical Equipment Room 3, 4 Table 2H-2 Unit 2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Aux Building 317' Emergency Core Cooling Rooms 3, 4 Aux Building 317' Tendon Gallery Access 3, 4 Aux Building 335' Charging Pumps / MCC 2B-52 3, 4 Aux Building 354' MCC 2B-62 Area 3, 4 Emergency Diesel Generator Corridor 3, 4 Lower South Piping Penetration Room 3, 4 Aux Building 386' Containment Hatch 3, 4 to 0CAN031801 Page 184 of 240 Attachment 1 - Emergency Action Level Technical Bases Mode Applicability:

3 - Hot Standby, 4 - Hot Shutdown Definition(s): IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures 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). Basis:

This IC addresses an event involving a release of a hazardous gas that precludes or IMPEDES access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown. This condition represents an actual or potential substantial degradation of the level of safety of the plant.

An Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the gaseous release. The emergency classification is not contingent upon whether entry is actually necessary at the time of the release.

Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the Emergency Director's judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly IMPEDE procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as IMPEDED if extraordinary measures 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).

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 affected room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous 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 require entry into the affected room until Mode 4. The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing). The access control measures are of a conservative or precautionary nature, and would not actually prevent or IMPEDE a required action. to 0CAN031801 Page 185 of 240 Attachment 1 - Emergency Action Level Technical Bases If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event. An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels.

Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death. This EAL does not apply to firefighting activities that generate smoke and that automatically or manually activate a fire suppression system in an area.

The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1). Escalation of the emergency classification level would be via Recognition Category A, C or F ICs.

EAL HA5.1 mode applicability has been limited to the mode limitations of Table 1[2]H-2 (Modes 3 and 4 only). Reference(s): 1. Attachment 2 Safe Operation & Shutdown Areas Tables 1[2]A-3 & 1[2]H-2 Bases 2. NEI 99-01 HA5 to 0CAN031801 Page 186 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations EAL: HA6.1 Alert An event has resulted in plant control being transferred from the Control Room to alternate locations Mode Applicability: All Definition(s): None Basis:

This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety.

Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown 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, will present challenges to plant operators and other on-shift personnel. Activation of the ERO and emergency response facilities will assist in responding to these challenges.

Transfer of plant control begins when the last licensed operator leaves the Control Room. Escalation of the emergency classification level would be via IC HS6.

Reference(s):

1. OP-1203.002 Alternate Shutdown 2. OP-2203.014 Alternate Shutdown

3. NEI 99-01 HA6 to 0CAN031801 Page 187 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Inability to control a key safety function from outside the Control Room EAL: HS6.1 Site Area Emergency An event has resulted in plant control being transferred from the Control Room to alternate locations AND Control of any of the following key safety functions is not re-established within 15 min. (Note 1): Reactivity (Modes 1, 2 and 3 only) Core cooling RCS heat removal Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling

Definition(s): None Basis:

This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time. The determination of whether or not "control" is established at the remote safe shutdown location(s) is based on Emergency Director judgment. The Emergency Director is expected to make a reasonable, informed judgment within 15 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s).

to 0CAN031801 Page 188 of 240 Attachment 1 - Emergency Action Level Technical Bases Transfer of plant control and the time period to establish control begins when the last licensed operator leaves the Control Room. Escalation of the emergency classification level would be via IC FG1 or CG1 Reference(s):

1. OP-1203.002 Alternate Shutdown 2. OP-2203.014 Alternate Shutdown

3. NEI 99-01 HS6 to 0CAN031801 Page 189 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a UE EAL: HU7.1 Unusual Event Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which indicate a potential degradation of the level of 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 expected unless further degradation of SAFETY SYSTEMS occurs. Mode Applicability:

All Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the 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 result in potential offsite exposures.

Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an UNUSUAL EVENT. Reference(s): 1. NEI 99-01 HU7 to 0CAN031801 Page 190 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of an ALERT EAL: HA7.1 Alert Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. Mode Applicability: All Definition(s):

HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station.

HOSTILE ACTION - An act toward ANO or its personnel that includes the use of violent force 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 other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on ANO. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA). OWNER CONTROLLED AREA - For the purposes of classification this is the Security Owner Controlled Area (SOCA). The SOCA is demarcated as a Vehicle Barrier System (VBS) and a detection fence on the outside and a delay fence on the inside of the passive and active barriers. The SOCA is the area inside the SOCA VBS up to the PROTECTED AREA fence line. Access to this area is controlled by the SOCA Personnel Access Control Point.

PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. to 0CAN031801 Page 191 of 240 Attachment 1 - Emergency Action Level Technical Bases Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an ALERT.

Reference(s): 1. NEI 99-01 HA7

to 0CAN031801 Page 192 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a SITE AREA EMERGENCY EAL: HS7.1 Site Area Emergency Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY. Mode Applicability: All Definition(s): HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station.

HOSTILE ACTION - An act toward ANO or its personnel that includes the use of violent force 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 other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on ANO. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA).

OWNER CONTROLLED AREA - For the purposes of classification this is the Security Owner Controlled Area (SOCA). The SOCA is demarcated as a Vehicle Barrier System (VBS) and a detection fence on the outside and a delay fence on the inside of the passive and active barriers. The SOCA is the area inside the SOCA VBS up to the PROTECTED AREA fence line. Access to this area is controlled by the SOCA Personnel Access Control Point.

PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

to 0CAN031801 Page 193 of 240 Attachment 1 - Emergency Action Level Technical Bases PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. SITE BOUNDARY - That boundary defined by a 1046 meter (0.65 mile) radius around the plant. Basis:

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a SITE AREA EMERGENCY.

Reference(s): 1. NEI 99-01 HS7

to 0CAN031801 Page 194 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a GENERAL EMERGENCY EAL: HG7.1 General Emergency Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area. Mode Applicability: All Definition(s):

HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station.

HOSTILE ACTION - An act toward ANO or its personnel that includes the use of violent force 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 other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on ANO. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA). IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

OWNER CONTROLLED AREA - For the purposes of classification this is the Security Owner Controlled Area (SOCA). The SOCA is demarcated as a Vehicle Barrier System (VBS) and a detection fence on the outside and a delay fence on the inside of the passive and active barriers. The SOCA is the area inside the SOCA VBS up to the PROTECTED AREA fence line. Access to this area is controlled by the SOCA Personnel Access Control Point.

PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. to 0CAN031801 Page 195 of 240 Attachment 1 - Emergency Action Level Technical Bases PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a GENERAL EMERGENCY.

Reference(s): 1. NEI 99-01 HG7

to 0CAN031801 Page 196 of 240 Attachment 1 - Emergency Action Level Technical Bases Category S - System Malfunction EAL Group: Hot Conditions (RCS temperature > 200°F); EALs in this category are applicable only in one or more hot operating modes.

Numerous system-related equipment failure events that warrant emergency classification have been 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 Vital AC Power Loss of vital electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite sources for vital 4.16 KV buses. 2. Loss of Vital DC Power Loss of emergency electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of vital plant 125V DC power sources. 3. Loss of Control Room Indications Certain events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification. Losses of indicators are in this subcategory. 4. RCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant rise from these base-line levels (2% - 5% clad failures) is indicative of fuel failures and is covered under the Fission Product Barrier Degradation category. However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits. These fission products will be circulated with the reactor coolant and can be detected by coolant sampling. 5. RCS Leakage The reactor vessel provides a volume for the coolant that covers the reactor core. The reactor pressure vessel and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail. Excessive RCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and containment integrity. to 0CAN031801 Page 197 of 240 Attachment 1 - Emergency Action Level Technical Bases 6. RPS Failure This 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 the RPS to complete a reactor trip comprise a specific set of analyzed events referred to as Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, ATWS is intended to mean any trip failure event that does not achieve reactor shutdown. If RPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could cause a threat to fuel clad, RCS and containment integrity. 7. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification. 8. Containment Failure Failure of containment isolation capability (under conditions in which the containment is not currently challenged) warrants emergency classification. Failure of containment pressure control capability also warrants emergency classification. 9. Hazardous Event Affecting Safety Systems Various natural and technological events that result in degraded plant safety system performance or significant VISIBLE DAMAGE warrant emergency classification under this subcategory.

to 0CAN031801 Page 198 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of Vital AC Power Initiating Condition: Loss of all offsite AC power capability to vital buses for 15 minutes or longer EAL: SU1.1 Unusual Event Loss of all offsite AC power capability, Table 1[2]S-1, to vital 4.16 KV buses A3[2A3] and A4[2A4] for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table 1S-1 Unit 1 AC Power Sources Offsite Startup Transformer No. 1 Startup Transformer No. 2 Unit Auxiliary Transformer (from 22 KV switchyard) Onsite Unit Auxiliary Transformer (main generator via main transformer) DG1 DG2 AAC Gen to 0CAN031801 Page 199 of 240 Attachment 1 - Emergency Action Level Technical Bases Table 2S-1 Unit 2 AC Power Sources Offsite Startup Transformer No. 3 Startup Transformer No. 2 Unit Auxiliary Transformer (backfed from main transformer) Onsite Unit Auxiliary Transformer (main generator via main transformer) 2DG1 2DG2 AAC Gen Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s): None

Basis: This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC vital buses. This condition represents 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 vital buses, whether or not the buses are powered from it.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power. Escalation of the emergency classification level would be via IC SA1.

Reference(s): 1. 1SAR Figure 8-1 Station Single Line Diagram 2. OP-1202.007 Degraded Power 3. OP-1202.008 Blackout

4. OP-2104.037 Alternate AC Diesel Generator Operations to 0CAN031801 Page 200 of 240 Attachment 1 - Emergency Action Level Technical Bases 5. 2SAR Figure 8.3-1 Station Single Line Diagram 6. OP-2202.007 Loss of Off-Site Power 7. OP-2202.008 Station Blackout 8. OP-2107.006 Backfeed of Unit Auxiliary Transformer

9. NEI 99-01 SU1 to 0CAN031801 Page 201 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of Vital AC Power Initiating Condition: Loss of all but one AC power source to vital buses for 15 minutes or longer EAL: SA1.1 Alert AC power capability, Table 1[2]S-1, to vital 4.16 KV buses A3[2A3] and A4[2A4] reduced to a single power source for 15 min. (Note 1) AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table 1S-1 Unit 1 AC Power Sources Offsite Startup Transformer No. 1 Startup Transformer No. 2 Unit Auxiliary Transformer (from 22 KV switchyard) Onsite Unit Auxiliary Transformer (main generator via main transformer) DG1 DG2 AAC Gen to 0CAN031801 Page 202 of 240 Attachment 1 - Emergency Action Level Technical Bases Table 2S-1 Unit 2 AC Power Sources Offsite Startup Transformer No. 3 Startup Transformer No. 2 Unit Auxiliary Transformer (backfed from main transformer) Onsite Unit Auxiliary Transformer (main generator via main transformer) 2DG1 2DG2 AAC Gen Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the 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 result in potential offsite exposures. Basis: This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC SU1. to 0CAN031801 Page 203 of 240 Attachment 1 - Emergency Action Level Technical Bases An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to a vital bus. Some examples of this condition are presented below. A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator). A loss of all offsite power and loss of all vital power sources (e.g., onsite diesel generators) with a single train of vital buses being back-fed from the unit main generator. A loss of vital power sources (e.g., onsite diesel generators) with a single train of vital buses being back-fed from an offsite power source.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. Escalation of the emergency classification level would be via IC SS1.

This EAL is the hot condition equivalent of the cold condition EAL CU2.1.

Reference(s): 1. 1SAR Figure 8-1 Station Single Line Diagram 2. OP-1202.007 Degraded Power

3. OP-1202.008 Blackout

4. OP-2104.037 Alternate AC Diesel Generator Operations 5. 2SAR Figure 8.3-1 Station Single Line Diagram 6. OP-2202.007 Loss of Off-Site Power

7. OP-2202.008 Station Blackout

8. OP-2107.006 Backfeed of Unit Auxiliary Transformer

9. NEI 99-01 SA1 to 0CAN031801 Page 204 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of Vital AC Power Initiating Condition: Loss of all offsite power and all onsite AC power to vital buses for 15 minutes or longer EAL: SS1.1 Site Area Emergency Loss of all offsite and all onsite AC power to vital 4.16 KV buses A3[2A3] and A4[2A4] for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown

Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2). Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the 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 result in potential offsite exposures. Basis:

Although the AAC may be considered available, it will not prevent declaration of this EAL unless it is powering a vital bus within the 15 minute time period of the EAL.

This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. Mitigative strategies using non-safety related power sources (FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or emergency operating procedures, or beyond design basis to 0CAN031801 Page 205 of 240 Attachment 1 - Emergency Action Level Technical Bases accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures 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 IC AG1, FG1 or SG1.

This EAL is the hot condition equivalent of the cold condition EAL CA2.1.

Reference(s): 1. 1SAR Figure 8-1 Station Single Line Diagram 2. OP-1202.007 Degraded Power

3. OP-1202.008 Blackout

4. OP-2104.037 Alternate AC Diesel Generator Operations 5. 2SAR Figure 8.3-1 Station Single Line Diagram 6. OP-2202.007 Loss of Off-Site Power

7. OP-2202.008 Station Blackout

8. OP-2107.006 Backfeed of Unit Auxiliary Transformer

9. NEI 99-01 SS1 to 0CAN031801 Page 206 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of Vital AC Power Initiating Condition: Prolonged loss of all offsite and all onsite AC power to vital buses EAL: SG1.1 General Emergency Loss of all offsite and all onsite AC power to vital 4.16 KV buses A3[2A3] and A4[2A4] AND EITHER: Restoration of at least one vital 4.16 KV bus in < 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely (Note 1) CETs > 1200°F Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the 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 result in potential offsite exposures. Basis: This IC addresses a prolonged loss of all power sources to AC vital buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. Mitigative strategies using non-safety related power sources (FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or to 0CAN031801 Page 207 of 240 Attachment 1 - Emergency Action Level Technical Bases emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. A prolonged loss of these buses will lead to a loss of one or more fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these conditions.

The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1. 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 projected that power cannot be restored to at least one AC vital bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is a greater likelihood of challenges to multiple fission product barriers. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is the site-specific SBO coping analysis time (ref. 4, 5).

The estimate for restoring at least one vital bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.

The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.

Reference(s): 1. OP-1202.005 Inadequate Core Cooling 2. OP-2202.009 Functional Recovery

3. OP-2202.011 Lower Mode Functional Recovery

4. Unit 1 Calculation 85-E-0072-02 Time from Loss of All AC Power to Loss of Subcooling 5. Unit 2 Calculation 85-E-0072-01 Time from Loss of All AC Power to Loss of Subcooling 6. 1SAR Figure 8-1 Station Single Line Diagram

7. OP-1202.007 Degraded Power

8. OP-1202.008 Blackout

9. OP-2104.037 Alternate AC Diesel Generator Operations 10. 2SAR Figure 8.3-1 Station Single Line Diagram 11. OP-2202.007 Loss of Off-Site Power

12. OP-2202.008 Station Blackout

13. OP-2107.006 Backfeed of Unit Auxiliary Transformer

14. NEI 99-01 SG1 to 0CAN031801 Page 208 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of Vital AC Power Initiating Condition: Loss of all vital AC and vital DC power sources for 15 minutes or longer EAL: SG1.2 General Emergency Loss of all offsite and all onsite AC power to vital 4.16 KV buses A3[2A3] and A4[2A4] for 15 min. (Note 1) AND Indicated voltage is < 105 VDC on D01[2D01] and D02[2D02] vital 125 VDC buses for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the 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 result in potential offsite exposures. Basis: Unit 1 batteries D06 and D07 and Unit 2 batteries 2D11 and 2D12 contain 58 cells each with a minimum cell voltage of 1.81 V or 105 VDC (ref. 9, 10). This IC addresses a concurrent and prolonged loss of both vital AC and Vital DC power. A loss of all vital AC power compromises the performance of all SAFETY SYSTEMS requiring electric to 0CAN031801 Page 209 of 240 Attachment 1 - Emergency Action Level Technical Bases power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. Mitigative strategies using non-safety related power sources (FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. A loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both vital AC and vital DC power will lead to multiple challenges to fission 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 are met. Reference(s): 1. 1SAR Figure 8-1 Station Single Line Diagram 2. OP-1202.007 Degraded Power 3. OP-1202.008 Blackout 4. OP-2104.037 Alternate AC Diesel Generator Operations

5. 2SAR Figure 8.3-1 Station Single Line Diagram

6. OP-2202.007 Loss of Off-Site Power

7. OP-2202.008 Station Blackout 8. OP-2107.006 Backfeed of Unit Auxiliary Transformer 9. 1SAR 8.3.2.1.1 Batteries

10. 2SAR 8.3.2.1.1 Batteries

11. OP-1203.036 Loss of 125V DC

12. OP-2203.037 Loss of 125V DC

13. 2SAR Figure 8.3-6 Low Voltage Safety System Power Supplies 14. NEI 99-01 SG8 to 0CAN031801 Page 210 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 2 - Loss of Vital DC Power Initiating Condition: Loss of all vital DC power for 15 minutes or longer EAL: SS2.1 Site Area Emergency Indicated voltage is < 105 VDC on D01[2D01] and D02[2D02] vital 125 VDC buses for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the 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 result in potential offsite exposures. Basis: Unit 1 batteries D06 and D07 and Unit 2 batteries 2D11 and 2D12 contain 58 cells each with a minimum cell voltage of 1.81 V or 105 VDC (ref. 2, 3). This IC addresses a loss of vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure 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 IC AG1, FG1 or SG1. This EAL is the hot condition equivalent of the cold condition EAL CU4.1. to 0CAN031801 Page 211 of 240 Attachment 1 - Emergency Action Level Technical Bases Reference(s): 1. 1SAR Figure 8-1 Station Single Line Diagram 2. 1SAR 8.3.2.1.1 Batteries

3. 2SAR 8.3.2.1.1 Batteries

4. OP-1203.036 Loss of 125V DC 5. OP-2203.037 Loss of 125V DC 6. 2SAR Figure 8.3-6 Low Voltage Safety System Power Supplies

7. NEI 99-01 SS8 to 0CAN031801 Page 212 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer EAL: SU3.1 Unusual Event An UNPLANNED event results in the inability to monitor one or more Table 1[2]S-2 parameters from within the Control Room for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table 1[2]S-2 Safety System Parameters Reactor power RCS level RCS pressure CET temperature Level in at least one S/G EFW flow to at least one S/G Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the 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 result in potential offsite exposures. to 0CAN031801 Page 213 of 240 Attachment 1 - Emergency Action Level Technical Bases UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis:

This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.

As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level would be via EAL SA3.1. Reference(s): 1. 1SAR 7.5 Safety-Related Display Instrumentation 2. 2SAR 7.5 Safety-Related Display Instrumentation 3. NEI 99-01 SU2 to 0CAN031801 Page 214 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress EAL: SA3.1 Alert An UNPLANNED event results in the inability to monitor one or more Table 1[2]S-2 parameters from within the Control Room for 15 min. (Note 1) AND Any significant transient is in progress, Table 1[2]S-3 Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table 1[2]S-2 Safety System Parameters Reactor power RCS level RCS pressure CET temperature Level in at least one S/G EFW flow to at least one S/G Table 1[2]S-3 Significant Transients Reactor trip Runback > 25% thermal power Electrical load rejection > 25% electrical load Safety injection actuation to 0CAN031801 Page 215 of 240 Attachment 1 - Emergency Action Level Technical Bases Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the 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 result in potential offsite exposures. UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis: This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant. As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be to 0CAN031801 Page 216 of 240 Attachment 1 - Emergency Action Level Technical Bases more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level would be via IC FS1 or AS1.

Reference(s): 1. 1SAR 7.1.3 Engineered Safeguards Actuation System 2. 2SAR 7.3 Engineered Safety Features Systems

3. NEI 99-01 SA2 to 0CAN031801 Page 217 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 4 - RCS Activity Initiating Condition: RCS activity greater than Technical Specification allowable limits EAL: SU4.1 Unusual Event Failed Fuel Iodine radiation monitor RI-1237S[2RITS-4806B] > 9.0 E5 cpm Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s): None Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.

Escalation of the emergency classification level would be via IC FA1 or the Recognition Category A ICs.

Unit 1 RE-1237S, Failed Fuel Monitor, is in the letdown system to monitor the letdown line for evidence of fuel damage.

Unit 2 specific activity monitor 2RITS-4806B monitors the Letdown fluid for the presence of Iodine-131.

A monitor reading corresponding to the instantaneous dose equivalent I-131 value of 60 uCi/gm is determined by multiplying by 30 the monitor reading listed in the table in OP-1203.019[OP-2203.020] that represents a projected 2.0 uCi/gm I-131 RCS activity(ref. 2, 5).

This yields values of 3.1E6 cpm for Unit 1 and 3.9E6 cpm for Unit 2. The top of scale of the monitor is 1E6. The EAL value is set at 9.0 E5 cpm for both units which is 90% of the top of the scale. to 0CAN031801 Page 218 of 240 Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. 1SAR Table 11-7 2. OP-1203-019 High Activity in Reactor Coolant

3. Unit 1 Technical Specifications LCO 3.4.12 RCS Specific Activity 4. 2SAR 9.3.5 Failed Fuel Detection System

5. OP-2203.020 High Activity in RCS 6. OP- 2203.012L ANNUNCIATOR 2K12 CORRECTIVE ACTION, A-1 7. Unit 2 Technical Specifications LCO 3.4.8 Reactor Coolant System Specific Activity 8. NEI 99-01 SU3 to 0CAN031801 Page 219 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 4 - RCS Activity Initiating Condition: RCS activity greater than Technical Specification allowable limits EAL: SU4.2 Unusual Event RCS sample activity > 1.0 µCi/gm dose equivalent I-131 for > 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> (Note 1) OR RCS sample activity > 60 µCi/gm dose equivalent I-131 OR RCS sample activity > 2200[3100] µCi/gm dose equivalent Xe-133 for > 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s): None Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.

Escalation of the emergency classification level would be via IC FA1 or the Recognition Category A ICs.

Reference(s): 1. Unit 1 Technical Specifications LCO 3.4.12 RCS Specific Activity 2. Unit 2 Technical Specifications LCO 3.4.8 Reactor Coolant System Specific Activity 3. NEI 99-01 SU3 to 0CAN031801 Page 220 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 5 - RCS Leakage Initiating Condition: RCS leakage for 15 minutes or longer EAL: SU5.1 Unusual Event RCS unidentified or pressure boundary leakage > 10 gpm for 15 min. (Note 1) OR RCS identified leakage > 25 gpm for 15 min. (Note 1) OR Reactor coolant leakage to a location outside containment > 25 gpm for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s): UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally. Basis: Failure to isolate the leak (from the Control Room or locally) within 15 minutes, or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification. Steam generator tube leakage is identified RCS leakage. This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant. The first and second EAL conditions are focused on a loss of mass from the RCS due to "unidentified leakage", "pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the plant Technical Specifications). The third condition addresses an RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions thus apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage) or a location outside of containment. to 0CAN031801 Page 221 of 240 Attachment 1 - Emergency Action Level Technical Bases The leak rate values for each condition were selected because they are usually observable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). The first condition uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.

The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. An emergency classification would be required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated).

The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.

Escalation of the emergency classification level would be via ICs of Recognition Category A or F.

Reference(s):

1. Unit 1 and Unit 2 Technical Specifications Section 1.1 Definitions 2. NEI 99-01 SU4 to 0CAN031801 Page 222 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor EAL: SU6.1 Unusual Event An automatic trip did not shut down the reactor as indicated by reactor power > 5% after any RPS setpoint is exceeded AND A subsequent automatic trip or manual trip action taken at the reactor control console (C03 [2C03/2C14]) (manual reactor trip pushbuttons or DROPS[DSS]) is successful in shutting down the reactor as indicated by reactor power 5% (Note 8) Note 8: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Mode Applicability: 1 - Power Operation Definition(s):

IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.

In the event that the operator identifies a reactor trip is IMMINENT and initiates a successful manual reactor trip before the automatic trip setpoint is reached, no declaration is required. The successful manual trip of the reactor before it reaches its automatic trip setpoint or reactor trip signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss. to 0CAN031801 Page 223 of 240 Attachment 1 - Emergency Action Level Technical Bases Following the failure of an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.

If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip using a different switch). Depending upon several factors, the initial or subsequent effort to manually trip the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to 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, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip).

This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles."

The plant response to the failure of an automatic or manual reactor trip 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, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event.

Should a reactor trip signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.

If the signal generated as a result of plant work causes a plant transient that results in a condition that should have included an automatic reactor trip and the RPS fails to automatically shutdown the reactor, then this IC and associated EALs are applicable, and should be evaluated. If the signal generated as a result of plant work does not cause a plant transient and the trip failure is determined through other means (e.g., assessment of test results), then this IC and associated EALs are not applicable and no classification is warranted. to 0CAN031801 Page 224 of 240 Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. Unit 1 Technical Specifications Table 3.3.1-1 Reactor Protection System Instrumentation 2. Unit 2 Technical Specifications Table 3.3-1 Reactor Protective Instrumentation

3. Unit 1 and Unit 2 Technical Specifications Table 1.1-1 Modes

4. OP-1202.001 Reactor Trip

5. OP-2202.001 Standard Post Trip Actions 6. NEI 99-01 SU5 to 0CAN031801 Page 225 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor EAL: SU6.2 Unusual Event A manual trip did not shut down the reactor as indicated by reactor power > 5% after any manual trip action was initiated AND A subsequent automatic trip or manual trip action taken at the reactor control console (C03 [2C03/2C14]) (manual reactor trip pushbuttons or DROPS[DSS]) is successful in shutting down the reactor as indicated by reactor power 5% (Note 8) Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Mode Applicability: 1 - Power Operation Definition(s):

None Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.

This EAL addresses a failure of a manually initiated trip in the absence of having exceeded an automatic RPS trip setpoint and a subsequent automatic or manual trip is successful in shutting down the reactor.

Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.

to 0CAN031801 Page 226 of 240 Attachment 1 - Emergency Action Level Technical Bases If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip using a different switch). Depending upon several factors, the initial or subsequent effort to manually the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to 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, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip).

This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles."

The plant response to the failure of an automatic or manual reactor trip 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, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event. Should a reactor trip signal be generated as a result of plant work (e.g., RPS setpoint testing),

the following classification guidance should be applied.

If the signal generated as a result of plant work causes a plant transient that results in a condition that should have included an automatic reactor trip and the RPS fails to automatically shutdown the reactor, then this IC and associated EALs are applicable, and should be evaluated. If the signal generated as a result of plant work does not cause a plant transient and the trip failure is determined through other means (e.g., assessment of test results), then this IC and associated EALs are not applicable and no classification is warranted. Reference(s):

1. Unit 1 Technical Specifications Table 3.3.1-1 Reactor Protection System Instrumentation 2. Unit 2 Technical Specifications Table 3.3-1 Reactor Protective Instrumentation

3. Unit 1 and Unit 2 Technical Specifications Table 1.1-1 Modes

4. OP-1202.001 Reactor Trip

5. OP-2202.001 Standard Post Trip Actions 6. NEI 99-01 SU5 to 0CAN031801 Page 227 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor EAL: SA6.1 Alert An automatic or manual trip fails to shut down the reactor as indicated by reactor power > 5% AND Manual trip actions taken at the reactor control console (C03[2C03/2C14]) (manual reactor trip pushbuttons or DROPS[DSS]) are not successful in shutting down the reactor as indicated by reactor power > 5% (Note 8) Note 8: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Mode Applicability: 1 - Power Operation Definition(s): None Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS.

A manual action at the reactor control console is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip).

This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control console (e.g., locally opening breakers). Actions taken at back panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control console." to 0CAN031801 Page 228 of 240 Attachment 1 - Emergency Action Level Technical Bases The plant response to the failure of an automatic or manual reactor trip 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, other concurrent plant conditions, etc. If the failure to shut down the reactor is prolonged enough to cause a challenge to the core cooling or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS6. Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent the plant conditions needed to meet either IC SS6 or FS1, an Alert declaration is appropriate for this event.

It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration. Reference(s): 1. Unit 1 Technical Specifications Table 3.3.1-1 Reactor Protection System Instrumentation 2. Unit 2 Technical Specifications Table 3.3-1 Reactor Protective Instrumentation 3. Unit 1 and Unit 2 Technical Specifications Table 1.1-1 Modes 4. OP-1202.001 Reactor Trip

5. OP-2202.001 Standard Post Trip Actions

6. NEI 99-01 SA5 to 0CAN031801 Page 229 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Inability to shut down the reactor causing a challenge to core cooling or RCS heat removal EAL: SS6.1 Site Area Emergency An automatic or manual trip fails to shut down the reactor as indicated by reactor power > 5% AND All actions to shut down the reactor are not successful as indicated by reactor power > 5% AND EITHER: CETs > 1200°F RCS heat removal cannot be established using steam generators and HPI[Once Through] cooling initiated. Mode Applicability: 1 - Power Operation

Definition(s): None Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.

In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shutdown the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor.

Escalation of the emergency classification level would be via IC AG1 or FG1. to 0CAN031801 Page 230 of 240 Attachment 1 - Emergency Action Level Technical Bases Reference(s):

1. Unit 1 Technical Specifications Table 3.3.1-1 Reactor Protection System Instrumentation 2. Unit 2 Technical Specifications Table 3.3-1 Reactor Protective Instrumentation

3. Unit 1 and Unit 2 Technical Specifications Table 1.1-1 Modes

4. OP-1202.001 Reactor Trip

5. OP-2202.001 Standard Post Trip Actions 6. OP-1202.004 Overheating 7. OP-2202.006 Loss of Feedwater

8. OP-1202.013 Figure 1, Saturation and Adequate SCM

9. Calculation 90-E-0116-07 Unit 1 EOP Setpoint Document, Setpoint B.19

10. OP-2202.009 Functional Recovery 11. Calculation 90-E-0116-01 Unit 2 EOP Setpoint Document 12. NEI 99-01 SS5 to 0CAN031801 Page 231 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 7 - Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL: SU7.1 Unusual Event Loss of all Table 1[2]S-4 onsite communication methods OR Loss of all Table 1[2]S-4 State and local agency communication methods OR Loss of all Table 1[2]S-4 NRC communication methods Table 1[2]S-4 Communication Methods System Onsite State / Local NRC Station radio system X ANO plant phone system X Gaitronics X Telephone Systems: Commercial Microwave Satellite VOIP X X INFORM Notification System X Emergency Notification System (ENS) X Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown to 0CAN031801 Page 232 of 240 Attachment 1 - Emergency Action Level Technical Bases Definition(s):

None Basis: This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to State and local agencies and the NRC.

This IC should be assessed only when extraordinary means are being utilized to make communications 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 to offsite locations, etc.).

The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations.

The second EAL condition addresses a total loss of the communications methods used to notify all State and local agencies of an emergency declaration. The State and local agencies referred to here are the Arkansas Department of Health, Arkansas Department of Emergency Management, Pope, Yell, Johnson, and Logan County agencies.

The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

This EAL is the hot condition equivalent of the cold condition EAL CU5.1.

Reference(s): 1. OP-1903.062 Communications System Operating Procedure 2. NEI 99-01 SU6 to 0CAN031801 Page 233 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 8 - Containment Failure Initiating Condition: Failure to isolate containment or loss of containment pressure control EAL: SU8.1 Unusual Event Any penetration is not closed within 15 min. of an ESAS[CIAS] actuation signal OR Containment pressure > 44.7 psia[23.3 psia] with < one full train of containment heat removal systems (Note 9) operating per design for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 9: One full train of containment heat removal systems consists of one train of RB [Containment] Spray and one train of RB [Containment] Cooling System. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown

Definition(s): None Basis:

A penetration is closed for this EAL if either side of the penetration has a closed valve or a check valve is intact (for penetrations that only have one automatic valve and a check valve).

This EAL addresses a failure of one or more containment penetrations to automatically isolate (close) when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems. Absent challenges to another fission product barrier, either condition represents potential degradation of the level of safety of the plant.

For the first condition, the containment isolation signal must be generated as the result on an off-normal/accident condition (e.g., a safety injection or high containment pressure); a failure resulting from testing or maintenance does not warrant classification. The determination of containment and penetration status - isolated or not isolated - should be made in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible. to 0CAN031801 Page 234 of 240 Attachment 1 - Emergency Action Level Technical Bases The second condition addresses a condition where containment pressure is greater than the setpoint at 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-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. The inability to start the required equipment indicates that containment heat removal/depressurization systems (e.g., containment sprays) 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 a concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers.

Reference(s): 1. OP-1202.010 ESAS 2. 1SAR 6.2 Reactor Building Spray System

3. 1SAR 6.3 Reactor Building Cooling System

4. OP-2202.003 Loss of Coolant Accident 5. OP-2202.010 Standard Attachments, Attachment 22 6. 2SAR 6.2.2 Containment Heat Removal Systems

7. 2SAR 7.3.1.1.11.2 Containment Spray System

8. NEI 99-01 SU7 to 0CAN031801 Page 235 of 240 Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 9 - Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode EAL: SA9.1 Alert The occurrence of any Table 1[2]S-5 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER: Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 10, 11) Note 10: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted. Note 11: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted. Table 1[2]S-5 Hazardous Events Seismic event (earthquake) Internal or external FLOODING event High winds or tornado strike FIRE EXPLOSION Other events with similar hazard characteristics as determined by the Shift Manager to 0CAN031801 Page 236 of 240 Attachment 1 - Emergency Action Level Technical Bases Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s): EXPLOSION - A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines 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 drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred 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 installed equipment 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/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10 CFR 50.2).

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the 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 result in potential offsite exposures.

VISIBLE DAMAGE - Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train. Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. to 0CAN031801 Page 237 of 240 Attachment 1 - Emergency Action Level Technical Bases Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.

Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make a determination of VISIBLE DAMAGE based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

Escalation of the emergency classification level would be via IC FS1 or AS1.

This EAL is the hot condition equivalent of the cold condition EAL CA6.1.

Reference(s): 1. EP FAQ 2016-002 2. NEI 99-01 SA9 to 0CAN031801 Page 238 of 240 Attachment 2 - Safe Operation & Shutdown Areas Tables 1[2]A-3 & 1[2]H-2 Bases Background NEI 99-01 Revision 6 ICs AA3 and HA5 prescribe declaration of an Alert based on impeded access 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 Developers Notes 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/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Do not include rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations). In addition, the list should specify the plant mode(s) during which entry would be required for each room or area. The list should not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections). Further, as specified in IC HA5:

The list need not include the Control Room if adequate engineered safety/design features are 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 air intakes at different and separate locations, inner and outer atmospheric boundaries, or the capability to acquire and maintain positive pressure within the Control Room envelope.

to 0CAN031801 Page 239 of 240 Attachment 2 - Safe Operation & Shutdown Areas Tables 1[2]A-3 & 1[2]H-2 Bases ANO Table 1[2]A-3 and 1[2]H-2 Bases A review of station operating procedures identified the following mode dependent in-plant actions and associated areas that are required for normal plant operation, cooldown or shutdown:

Unit 1 AREA MODES PURPOSE REFERENCE A-4 Switchgear Room 3, 4 Core flood tank valves, decay heat removal (DHR) OP-1102.010 OP-1104.004 Upper North Electrical Penetration Room 3, 4 DHR alignment OP-1104.004 Lower South Electrical Equipment Room 3, 4 DHR alignment OP-1104.004 Unit 2 AREA MODES PURPOSE REFERENCE Aux Building 317' Emergency Core Cooling Rooms 3, 4 Shutdown Cooling (SDC) venting and alignment OP-2104.004 Aux Building 317' Tendon Gallery Access 3, 4 SDC alignment OP-2104.004 Aux Building 335' Charging Pumps / Motor Control Center (MCC) 2B-52 3, 4 Charging low pressure operation, T-Hot injection valves, and SDC alignment OP-2102.010 OP-2104.004 Auxiliary Building 354' MCC 2B-62 Area 3, 4 SDC alignment and T-Hot injection valves at MCC 2B-62 OP-2102.010 OP-2104.004 Emergency Diesel Generator Corridor 3, 4 Close Safety Injection Tank (SIT) valves and SDC / Low Temperature Overpressure (LTOP) valve alignment at MCC 2B-51 OP-2102.010 Lower South Piping Penetration Room 3, 4 SDC alignment OP-2104.004 Aux Building 386' Containment Hatch 3, 4 Close SIT valves at MCC 2B-61 OP-2102.010 Mode 3 is included above for DHR- and SDC-related activities because the procedures begin alignment in Mode 3; however, these actions could be delayed until Mode 4, if necessary. In order to ensure adequate guidance to emergency response personnel, the above areas are added to the EAL in order to provide prompt operator guidance for EAL declaration. to 0CAN031801 Page 240 of 240 Attachment 2 - Safe Operation & Shutdown Areas Tables 1[2]A-3 & 1[2]H-2 Bases Both ANO-1 and ANO-2 Control Room ventilation systems have adequate engineered safety/design features in place to preclude a Control Room evacuation due to the release of a hazardous gas. Therefore the Control Room is not included in this assessment or in Tables 1[2]H-2. Table 1[2]A-3 & 1[2]H-2 Results Table 1[2]A-3 & 1[2]H-2 Safe Operation & Shutdown Rooms/Areas Unit 1 Room/Area Mode Applicability A-4 Switchgear Room 3, 4 Upper North Electrical Penetration Room 3, 4 Lower South Electrical Equipment Room 3, 4 Unit 2 Room/Area Mode Applicability Aux Building 317' Emergency Core Cooling Rooms 3, 4 Aux Building 317' Tendon Gallery Access 3, 4 Aux Building 335' Charging Pumps / MCC 2B-52 3, 4 Auxiliary Building 354' MCC 2B-62 Area 3, 4 Emergency Diesel Generator Corridor 3, 4 Lower South Piping Penetration Room 3, 4 Aux Building 386' Containment Hatch 3, 4

Enclosure 4 to 0CAN031801 NEI 99-01, Rev. 6, Deviations and Differences, ANO Units 1 and 2 to 0CAN031801 Page 1 of 120 Table of Contents Section Page Introduction ------------------------------------------------------------------------------------------------------------ 2 Comparison Matrix Format ---------------------------------------------------------------------------------------- 2

EAL Wording ---------------------------------------------------------------------------------------------------------- 2 EAL Emphasis Techniques ---------------------------------------------------------------------------------------- 2

Global Differences --------------------------------------------------------------------------------------------------- 3 Differences and Deviations ---------------------------------------------------------------------------------------- 5 Category A - Abnormal Rad Levels / Rad Effluent -------------------------------------------------------- 17

Category C - Cold Shutdown / Refueling System Malfunction ----------------------------------------- 38 Category D - Permanently Defueled Station Malfunction ----------------------------------------------- 61 Category E -Independent Spent Fuel Storage Installation (ISFSI) ----------------------------------- 62

Category F - Fission Product Barrier Degradation -------------------------------------------------------- 63 Category H - Hazards and Other Conditions Affecting Plant Safety --------------------------------- 76

Category S - System Malfunction ----------------------------------------------------------------------------- 95

Table 1 - ANO EAL Categories/Subcategories -------------------------------------------------------------- 8

Table 2 - NEI / ANO EAL Identification Cross-Reference ------------------------------------------------- 9 Table 3 - Summary of Deviations ----------------------------------------------------------------------------- 14

to 0CAN031801 Page 2 of 120 Introduction This document provides a line-by-line comparison of the Initiating Conditions (ICs), Mode Applicability, and Emergency Action Levels (EALs) in NEI 99-01, Rev. 6, "Final, Development of Emergency Action Levels for Non-Passive Reactors," ADAMS Accession Number ML12326A805, and Arkansas Nuclear One (ANO) Initiating Conditions (ICs), Mode Applicability, and EALs. This document provides a means of assessing ANO differences and deviations from the NRC endorsed guidance given in NEI 99-01. Discussion of ANO EAL bases and lists of source document references are given in the EAL Technical Bases Document. It is, therefore, advisable to reference the EAL Technical Bases Document for background information while using this document.

Comparison Matrix Format The ICs and EALs discussed in this document are grouped according to NEI 99-01 Recognition Categories. Within each Recognition Category, the ICs and EALs are listed in tabular format according to the order in which they are given in NEI 99-01. Generally, each row of the comparison matrix provides the following information:

NEI EAL/IC identifier NEI EAL/IC wording ANO EAL/IC identifier ANO EAL/IC wording Description of any differences or deviations EAL Wording In Section 4.1, NEI recommends the following: "The guidance in NEI 99-01 is not intended to be applied to plants "as-is"; however, developers should attempt to keep their site-specific schemes as close to the generic guidance as possible. The goal is to meet the intent of the generic Initiating Conditions (ICs) and Emergency Action Levels (EALs) within the context of site-specific characteristics - locale, plant design, operating features, terminology, etc. Meeting this goal will result in a shorter and less cumbersome NRC review and approval process, closer alignment with the schemes of other nuclear power plant sites and better positioning to adopt future industry-wide scheme enhancements." To assist the Emergency Director (ED), the ANO EALs have been written in a clear and concise style (to the extent that the differences from the NEI EAL wording could be reasonably documented and justified). This supports timely and accurate classification in the tense atmosphere of an emergency event. The EAL differences introduced to reduce reading burden comprise almost all of the differences justified in this document. EAL Emphasis Techniques Due to the width of the table columns and table formatting constraints in this document, line breaks and indentation may differ slightly from the appearance of comparable wording in the source documents. NEI 99-01 is the source document for the NEI EALs; the ANO EAL Technical Bases Document for the ANO EALs. to 0CAN031801 Page 3 of 120 Development of the ANO IC/EAL wording has attempted to minimize inconsistencies and apply sound human factors principles. As a result, differences occur between NEI and ANO ICs/EALs for these reasons alone. When such difference may infer a technical difference in the associated NEI IC/EAL, the difference is identified and a justification provided.

The print and paragraph formatting conventions summarized below guide presentation of the ANO EALs in accordance with the EAL writing criteria. Space restrictions in the EAL table of this document sometimes override this criteria in cases when following the criteria would introduce undesirable complications in the EAL layout. Upper case-bold print is used for the logic terms AND, OR, and EITHER. Bold font is used for certain logic terms, negative terms (not, cannot, etc.), any, all. Upper case print is reserved for defined terms, acronyms, system abbreviations, logic terms (and, or, etc. when not used as a conjunction), annunciator window engravings. Three or more items in a list are normally introduced with "Any of the following..." or "All of the following..." Items of the list begin with bullets when a priority or sequence is not inferred. The use of AND/OR logic within the same EAL has been avoided when possible. When such logic cannot be avoided, indentation and separation of subordinate contingent phrases is employed. Global Differences The differences listed below generally apply throughout the set of EALs and are not repeated in the Justification sections of this document. The global differences do not decrease the effectiveness of the intent of NEI 99-01.

1. The NEI phrase "Notification of Unusual Event" has been changed to "Unusual Event" or abbreviated "UE" to reduce EAL-user reading burden. 2. NEI 99-01 IC Example EALs are implemented in separate plant EALs to improve clarity and readability. For example, NEI lists all IC HU3 Example EALs under one IC. The corresponding ANO EALs appear as unique EALs (e.g., HU3.1 through HU3.4). 3. Mode applicability identifiers (numbers/letter) modify the NEI 99-01 mode applicability names as follows: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling, DEF - Defueled. NEI 99-01defines Defueled as follows: "Reactor Vessel contains no irradiated fuel (full core off-load during refueling or extended outage)." 4. NEI 99-01 uses the terms greater than, less than, greater than or equal to, etc., in the wording of some example EALs. For consistency and to reduce EAL-user reading burden, ANO has adopted use of boolean symbols in place of the NEI 99-01 text modifiers within the EAL wording. to 0CAN031801 Page 4 of 120 5. "min." is the standard abbreviation for "minutes" and is used to reduce EAL user reading burden. 6. Wherever the generic bracketed PWR term "reactor vessel/RCS" is provided, ANO uses the term "RCS" as the site-specific nomenclature. 7. IC/EAL identification: NEI 99-01 defines the thresholds requiring emergency classification (example EALs) and assigns them to ICs which, in turn, are grouped in "Recognition Categories." ANO endeavors to optimize the NEI EAL organization and identification scheme to enhance usability of the plant-specific EAL set. To this end, the ANO IC/EAL scheme includes the following features: a. Division of the NEI EAL set into three groups: o EALs applicable under all plant operating modes - This group would be reviewed by the EAL-user any time emergency classification is considered. o EALs applicable only under hot operating modes - This group would only be reviewed 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 be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling, or Defueled mode. The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition. This approach significantly minimizes the total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency. b. Within each of the above three groups, EALs are assigned to categories and subcategories. Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. Subcategories are used as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The ANO EAL categories/subcategories and their relationship to NEI Recognition Categories are listed in Table 1. to 0CAN031801 Page 5 of 120 c. Unique identification of each EAL - Four characters comprise the EAL identifier as illustrated in Figure 1. Figure 1 - EAL Identifier EAL Identifier XXX.X The first character is a letter associated with the category in which the EAL is located. The second character is a letter associated with the emergency classification level (G for General Emergency, S for Site Area Emergency, A for Alert, and U for Notification of Unusual Event). The third character is a number associated with one or more subcategories within a given category. Subcategories are sequentially numbered beginning with the number "1". If a category does not have a subcategory, this character is assigned the number "1". The fourth character is a number preceded by a period for each EAL within a subcategory. EALs are sequentially numbered within the emergency classification level of a subcategory beginning with the number "1". The EAL identifier is designed to fulfill the following objectives: o Uniqueness - The EAL identifier ensures that there can be no confusion over which EAL is driving the need for emergency classification. o Speed in locating the EAL of concern - When the EALs are displayed in a matrix format, knowledge of the EAL identifier alone can lead the EAL-user to the location of the EAL within the classification matrix. The identifier conveys the category, subcategory and classification level. This assists Emergency Response Organization (ERO) responders (who may not be in the same facility as the ED) to find the EAL of concern in a timely manner without the need for a word description of the classification threshold. o Possible classification upgrade - The category/subcategory/identifier scheme helps the EAL-user find higher emergency classification EALs that may become active if plant conditions worsen. Table 2 lists the ANO ICs and EALs that correspond to the NEI ICs/Example EALs when the above EAL/IC organization and identification scheme is implemented. Differences and Deviations In accordance NRC Regulatory Issue Summary (RIS) 2003-18 "Use of Nuclear Energy Institute (NEI) 99-01, Methodology for Development of Emergency Action Levels" Supplements 1 and 2, a difference is an EAL change in which the basis scheme guidance differs in wording but agrees in meaning and intent, such that classification of an event would be the same, whether using the basis scheme guidance or the ANO EAL. A deviation is an EAL change in which the basis Category (A, H, E, S, F, C) Emergency classification (G, S, A, U) Sequential number within subcategory/classification Subcategory number (1 if no subcategory) to 0CAN031801 Page 6 of 120 scheme guidance differs in wording and is altered in meaning or intent, such that classification of the event could be different between the basis scheme guidance and the ANO proposed EAL.

Administrative changes that do not actually change the textual content are neither differences nor deviations. Likewise, any format change that does not alter the wording of the IC or EAL is considered neither a difference nor a deviation.

The following are examples of differences: Choosing the applicable EAL based upon plant type (i.e., BWR vs. PWR). Using a numbering scheme other than that provided in NEI 99-01 that does not change the intent of the overall scheme. Where the NEI 99-01 guidance specifically provides an option to not include an EAL, if equipment for the EAL does not exist at ANO (e.g., automatic real-time dose assessment capability). Pulling information from the bases section up to the actual EAL that does not change the intent of the EAL. Choosing to state ALL Operating Modes are applicable instead of stating N/A, or listing each mode individually under the Abnormal Rad Level/Radiological Effluent and Hazard and Other Conditions Affecting Plant Safety sections. Using synonymous wording (e.g., greater than or equal to vs. at or above, less than or equal vs. at or below, greater than or less than vs. above or below, etc.) Adding ANO equipment/instrument identification and/or noun names to EALs. Combining like ICs that are exactly the same, but have different operating modes as long as the intent of each IC is maintained and the overall progression of the EAL scheme is not affected. Any change to the IC and/or EAL, and/or basis wording, as stated in NEI 99-01, that does not alter the intent of the IC and/or EAL, i.e., the IC and/or EAL continues to: o Classify at the correct classification level. o Logically integrate with other EALs in the EAL scheme. o Ensure that the resulting EAL scheme is complete (i.e., classifies all potential emergency conditions). The following are examples of deviations: Use of altered mode applicability. Altering key words or time limits. to 0CAN031801 Page 7 of 120 Changing words of physical reference (protected area, safety-related equipment, etc.). Eliminating an IC. This includes the removal of an IC from the Fission Product Barrier Degradation category as this impacts the logic of Fission Product Barrier ICs. Changing a Fission Product Barrier from a Loss to a Potential Loss or vice-versa. Not using NEI 99-01definitions as the intent is for all NEI 99-01 users to have a standard set of defined terms as defined in NEI 99-01. Differences due to plant types are permissible (BWR or PWR). Verbatim compliance to the wording in NEI 99-01 is not necessary as long as the intent of the defined word is maintained. Use of the wording provided in NEI 99-01 is encouraged since the intent is for all users to have a standard set of defined terms as defined in NEI 99-01. Any change to the IC and/or EAL, and/or basis wording as stated in NEI 99-01 that does alter the intent of the IC and/or EAL, i.e., the IC and/or EAL: o Does not classify at the classification level consistent with NEI 99-01. o Is not logically integrated with other EALs in the EAL scheme. o Results in an incomplete EAL scheme (i.e., does not classify all potential emergency conditions).

The "Difference/Deviation Justification" columns in the remaining sections of this document identify each difference between the NEI 99-01 IC/EAL wording and the ANO IC/EAL wording. An explanation that justifies the reason for each difference is then provided. If the difference is determined to be a deviation, a statement is made to that affect and explanation is given that states why classification may be different from the NEI 99-01 IC/EAL and the reason for its acceptability. In all cases, however, the differences and deviations do not decrease the effectiveness of the intent of NEI 99-01. A summary list of ANO EAL deviations from NEI 99-01 is given in Table 3.

to 0CAN031801 Page 8 of 120 Table 1 - ANO EAL Categories/Subcategories ANO EALs NEI Recognition Category Category Subcategory Group: Any Operating Mode: A - Abnormal Rad Levels/Rad Effluent 1 - Radiological Effluent 2 - Irradiated Fuel Event 3 - Area Radiation Levels Abnormal Rad Levels/Radiological Effluent ICs/EALs H - Hazards and Other Conditions Affecting Plant Safety E - ISFSI 1 - Security 2 - Seismic Event 3 - Natural or Technological Hazard 4 - Fire 5 - Hazardous Gas 6 - Control Room Evacuation 7 - Emergency Director Judgment 1 - Confinement Boundary Hazards and Other Conditions Affecting Plant Safety ICs/EALs ISFSI ICs/EALs Group: Hot Conditions: S - System Malfunction 1 - Loss of Vital AC Power 2 - Loss of Vital DC Power 3 - Loss of Control Room Indications 4 - RCS Activity 5 - RCS Leakage 6 - RPS Failure 7 - Loss of Communications 8 - Containment Failure 9 - Hazardous Event Affecting Safety Systems System Malfunction ICs/EALs F - Fission Product Barrier None Fission Product Barrier ICs/EALs Group: Cold Conditions: C - Cold Shutdown/Refueling System Malfunction 1 - RCS Level 2 - Loss of Vital AC Power 3 - RCS Temperature 4 - Loss of Vital DC Power 5 - Loss of Communications 6 - Hazardous Event Affecting Safety Systems Cold Shutdown./ Refueling System Malfunction ICs/EALs to 0CAN031801 Page 9 of 120 Table 2 - NEI / ANO EAL Identification Cross-Reference NEI ANO IC Example EAL Category and Subcategory EAL AU1 1 A - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent AU1.1 AU1 2 A - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent AU1.1 AU1 3 A - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent AU1.2 AU2 1 A - Abnormal Rad Levels / Rad Effluent, 2 - Irradiated Fuel Event AU2.1 AA1 1 A - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent AA1.1 AA1 2 A - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent AA1.2 AA1 3 A - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent AA1.3 AA1 4 A - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent AA1.4 AA2 1 A - Abnormal Rad Levels / Rad Effluent, 2 - Irradiated Fuel Event AA2.1 AA2 2 A - Abnormal Rad Levels / Rad Effluent, 2 - Irradiated Fuel Event AA2.2 AA2 3 A - Abnormal Rad Levels / Rad Effluent, 2 - Irradiated Fuel Event AA2.3 AA3 1 A - Abnormal Rad Levels / Rad Effluent, 3 - Area Radiation Levels AA3.1 AA3 2 A - Abnormal Rad Levels / Rad Effluent, 3 - Area Radiation Levels AA3.2 AS1 1 A - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent AS1.1 AS1 2 A - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent AS1.2 AS1 3 A - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent AS1.3 to 0CAN031801 Page 10 of 120 NEI ANO IC Example EAL Category and Subcategory EAL AS2 1 A - Abnormal Rad Levels / Rad Effluent, 2 - Irradiated Fuel Event AS2.1 AG1 1 A - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent AG1.1 AG1 2 A - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent AG1.2 AG1 3 A - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent AG1.3 AG2 1 A - Abnormal Rad Levels / Rad Effluent, 2 - Irradiated Fuel Event AG2.1 CU1 1 C - Cold SD/ Refueling System Malfunction, 1 - RCS Level CU1.1 CU1 2 C - Cold SD/ Refueling System Malfunction, 1 - RCS Level CU1.2 CU2 1 C - Cold SD/ Refueling System Malfunction, 2 - Loss of Vital AC Power CU2.1 CU3 1 C - Cold SD/ Refueling System Malfunction, 3 - RCS Temperature CU3.1 CU3 2 C - Cold SD/ Refueling System Malfunction, 3 - RCS Temperature CU3.2 CU4 1 C - Cold SD/ Refueling System Malfunction, 4 - Loss of Vital DC Power CU4.1 CU5 1, 2, 3 C - Cold SD/ Refueling System Malfunction, 5 - Loss of Communications CU5.1 CA1 1 C - Cold SD/ Refueling System Malfunction, 1 - RCS Level CA1.1 CA1 2 C - Cold SD/ Refueling System Malfunction, 1 - RCS Level CA1.2 CA2 1 C - Cold SD/ Refueling System Malfunction, 1 - Loss of Vital AC Power CA2.1 CA3 1, 2 C - Cold SD/ Refueling System Malfunction, 3 - RCS Temperature CA3.1 CA6 1 C - Cold SD/ Refueling System Malfunction, 6 - Hazardous Event Affecting Safety Systems HA4.1 CS1 1 C - Cold SD/ Refueling System Malfunction, 1 - RCS Level CS1.1 to 0CAN031801 Page 11 of 120 NEI ANO IC Example EAL Category and Subcategory EAL CS1 2 N/A N/A CS1 3 C - Cold SD/ Refueling System Malfunction, 1 - RCS Level CS1.2 CG1 1 C - Cold SD/ Refueling System Malfunction, 1 - RCS Level CG1.1 CG1 2 C - Cold SD/ Refueling System Malfunction, 1 - RCS Level CG1.2 E-HU1 1 E - ISFSI EU1.1 FA1 1 F - Fission Product Barrier Degradation FA1.1 FS1 1 F - Fission Product Barrier Degradation FS1.1 FG1 1 F - Fission Product Barrier Degradation FG1.1 HU1 1, 2, 3 H - Hazards and Other Conditions Affecting Plant Safety, 1 - Security HU1.1 HU2 1 H - Hazards and Other Conditions Affecting Plant Safety, 2 - Seismic Event HU2.1 HU3 1 H - Hazards and Other Conditions Affecting Plant Safety, 3 - Natural or Technological Hazard HU3.1 HU3 2 H - Hazards and Other Conditions Affecting Plant Safety, 3 - Natural or Technological Hazard HU3.2 HU3 3 H - Hazards and Other Conditions Affecting Plant Safety, 3 - Natural or Technological Hazard HU3.3 HU3 4 H - Hazards and Other Conditions Affecting Plant Safety, 3 - Natural or Technological Hazard HU3.4 HU3 5 H - Hazards and Other Conditions Affecting Plant Safety, 3 - Natural or Technological Hazard N/A HU4 1 H - Hazards and Other Conditions Affecting Plant Safety, 4 - Fire or Explosion HU4.1 HU4 2 H - Hazards and Other Conditions Affecting Plant Safety, 4 - Fire or Explosion HU4.2 HU4 3 H - Hazards and Other Conditions Affecting Plant Safety, 4 - Fire or Explosion HU4.3 to 0CAN031801 Page 12 of 120 NEI ANO IC Example EAL Category and Subcategory EAL HU4 4 H - Hazards and Other Conditions Affecting Plant Safety, 4 - Fire or Explosion HU4.4 HU7 1 H - Hazards and Other Conditions Affecting Plant Safety, 7 - Judgment HU7.1 HA1 1, 2 H - Hazards and Other Conditions Affecting Plant Safety, 1 - Security HA1.1 HA5 1 H - Hazards and Other Conditions Affecting Plant Safety, 5 - Hazardous Gases HA5.1 HA6 1 H - Hazards and Other Conditions Affecting Plant Safety, 6 - Control Room Evacuation HA6.1 HA7 1 H - Hazards and Other Conditions Affecting Plant Safety, 7 - Judgment HA7.1 HS1 1 H - Hazards and Other Conditions Affecting Plant Safety, 1 - Security HS1.1 HS6 1 H - Hazards and Other Conditions Affecting Plant Safety, 6 - Control Room Evacuation HS6.1 HS7 1 H - Hazards and Other Conditions Affecting Plant Safety, 7 - Judgment HS7.1 HG1 1 N/A N/A HG7 2 H - Hazards and Other Conditions Affecting Plant Safety, 7 - Judgment HG7.1 SU1 1 S - System Malfunction, 1 - Loss of Emergency AC Power SU1.1 SU2 1 S - System Malfunction, 3 - Loss of Control Room Indications SU3.1 SU3 1 S - System Malfunction, 4 - RCS Activity SU4.1 SU3 2 S - System Malfunction, 4 - RCS Activity SU4.2 SU4 1, 2, 3 S - System Malfunction, 5 - RCS Leakage SU5.1 SU5 1 S - System Malfunction, 6 - RPS Failure SU6.1 SU5 2 S - System Malfunction, 6 - RPS Failure SU6.2 to 0CAN031801 Page 13 of 120 NEI ANO IC Example EAL Category and Subcategory EAL SU6 1, 2, 3 S - System Malfunction, 7 -Loss of Communications SU7.1 SU7 1, 2 S - System Malfunction, 8 -Containment Failure SU8.1 SA1 1 S - System Malfunction, 1 - Loss of Emergency AC Power SA1.1 SA2 1 S - System Malfunction, 3 - Loss of Control Room Indications SA3.1 SA5 1 S - System Malfunction, 6 - RPS Failure SA6.1 SA9 1 S - Hazardous Event Affecting Safety Systems SA9.1 SS1 1 S - System Malfunction, 1 - Loss of Emergency AC Power SS1.1 SS5 1 S - System Malfunction, 6 - RPS Failure SS6.1 SS8 1 S - System Malfunction, 2 - Loss of Vital DC Power SS2.1 SG1 1 S - System Malfunction, 1 - Loss of Emergency AC Power SG1.1 SG8 2 S - System Malfunction, 1 - Loss of Emergency AC Power SG1.2 to 0CAN031801 Page 14 of 120 Table 3 - Summary of Deviations NEI ANO EAL Description IC Example EAL HG1 1 N/A IC HG1 and associated example EAL not implemented in the ANO scheme. There are several other ICs that are redundant with this IC, and are better suited to ensure timely and effective emergency declarations. In addition, the development of new spent fuel pool level EALs, as a result of NRC Order EA 051, clarified the intended emergency classification level for spent fuel pool level events. This deviation is justified because: 1. Hostile Action in the Protected Area is bounded by ICs HS1 and HS7. Hostile Action resulting in a loss of physical control is bound by EAL HG7, as well as any event that may lead to radiological releases to the public in excess of Environmental Protection Agency (EPA) Protective Action Guides (PAGs). a. If, for whatever reason, the Control Room must be evacuated, and control of safety functions (e.g., reactivity control, core cooling, and RCS heat removal) cannot be reestablished, then IC HS6 would apply, as well as IC HS7 if desired by the EAL decision-maker. b. Also, as stated above, any event (including Hostile Action) that could reasonably be expected to have a release exceeding EPA PAGs would be bound by IC HG7. c. From a Hostile Action perspective, ICs HS1, HS7 and HG7 are appropriate, and therefore, make this part of HG1 redundant and unnecessary. d. From a loss of physical control perspective, ICs HS6, HS7 and HG7 are appropriate, and therefore, make this part of HG1 redundant and unnecessary. 2. Any event which causes a loss of spent fuel pool level will be bounded by ICs AA2, AS2 and AG2, regardless of whether it was based upon a Hostile Action or not, thus making this part of HG1 redundant and unnecessary. a. An event that leads to a radiological release will be bounded by ICs AU1, AA1, AS1 and AG1. Events that lead to radiological releases in excess of EPA PAGs will be bounded by EALs AG1 and HG7, thus making this part of HG1 redundant and unnecessary. ICs AA2, AS2, AG2, AS1, AG1, HS1, HS6, HS7 and HG7 have been implemented consistent with NEI 99-01, Revision 6, and thus HG1 is adequately bounded as described above. Therefore, this is an acceptable deviation from the generic NEI 99-01, Revision 6, guidance and is consistent with NRC approved EP FAQ 2015-013. to 0CAN031801 Page 15 of 120 NEI ANO EAL Description IC Example EAL HS6 1 HS6.1 Deleted defueled mode applicability. Control of the cited safety functions are not critical for a defueled reactor as there is no energy source in the reactor vessel or RCS. The Mode applicability for the reactivity control safety function has been limited to Modes 1, 2, and 3 (hot operating conditions). In the cold operating modes adequate shutdown margin exists under all conditions. Therefore, this is an acceptable deviation from the generic NEI 99-01, Revision 6, guidance and is consistent with NRC approved EP FAQ 2015-014. CA6 SA9 1 1 CA6.1 SA9.1 The proposed ANO CA6.1 and SA9.1 wording is intended to ensure that an Alert should be declared only when actual or potential performance issues with SAFETY SYSTEMS have occurred as a result of a hazardous event. The occurrence of a hazardous event will result in an Unusual Event classification at a minimum. In order to warrant escalation to the Alert classification, the hazardous event should cause indications of degraded performance to one train of a SAFETY SYSTEM with either indications of degraded performance on the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second SAFETY SYSTEM train, such that the operability or reliability of the second train is a concern. In addition, escalation to the Alert classification should not occur if the damage from the hazardous event is limited to a SAFETY SYSTEM that was inoperable, or out of service, prior to the event occurring. As such, the proposed EALs will reduce the potential of declaring an Alert when events are in progress that do not involve an actual or potential substantial degradation of the level of safety of the plant, i.e., does not cause significant concern with shutting down or cooling down the plant. EALs CA6.1 and SA9.1 do not directly escalate to a Site Area Emergency or a General Emergency due to a hazardous event. The Fission Product Barrier and/or Abnormal Radiation Levels/Radiological Effluent recognition categories would provide an escalation path to a Site Area Emergency or a General Emergency. The EALs and the Basis sections have been revised to ensure potential escalations from an Unusual Event to an Alert, due to a hazardous event, is appropriate as the concern with these EALs is: (1) a hazardous event has occurred, (2) one SAFETY SYSTEM train is having performance issues as a result of the hazardous event, and (3) either the second SAFETY SYSTEM train is having performance issues or the VISIBLE DAMAGE is enough to be concerned that the second SAFETY SYSTEM train may have operability or reliability issues. The definition for VISIBLE DAMAGE has been revised to reflect the fact that the EALs are based upon SAFETY SYSTEM trains rather than individual components or structures. to 0CAN031801 Page 16 of 120 NEI ANO EAL Description IC Example EAL CA6 SA9 1 1 CA6.1 SA9.1 (continued) Note 10 has been added to CA6.1 and SA9.1 as it meets the intent of the EALs, is consistent with other EALs (e.g., EAL HA5.1 which was previously endorsed by the NRC), and ensures that declared emergencies are based upon unplanned events with the potential to pose a radiological risk to the public. Note 11 has been added to CA6.1 and SA9.1 to help reinforce and succinctly capture the more detailed information from the revised basis section related to when conditions would require the declaration of an Alert. CA6.1 and SA9.1 are consistent with NRC FAQ 2016-002 addressing degraded performance or visible damage to more than one safety system train caused by the specified events. Based on the above information, this revised wording is an acceptable deviation from the generic NEI 99-01, Revision 6, guidance and is consistent with NRC-approved EP FAQ 2016-002. to 0CAN031801 Page 17 of 120 Category A Abnormal Rad Levels / Radiological Effluent NEI IC# NEI IC Wording and Mode Applicability ANO IC#(s) ANO IC Wording and Mode Applicability Difference/Deviation Justification AU1 Release of gaseous or liquid radioactivity greater than 2 times the (site-specific effluent release controlling document) limits for 60 minutes or longer. MODE: All AU1 Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer MODE: All The ANO ODCM is the site-specific effluent release controlling document. NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Reading on ANY effluent radiation monitor greater than 2 times the (site-specific effluent release controlling document) limits for 60 minutes or longer: (site-specific monitor list and threshold values corresponding to 2 times the controlling document limits) AU1.1 Reading on any Table 1[2]A-1 effluent radiation monitor > column "UE" for 60 min. (Notes 1, 2, 3) Example EALs #1 and #2 have been combined into a single EAL to simplify presentation. The NEI phrase "-effluent radiation monitor greater than 2 times the (site-specific effluent release controlling document)" and "effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit " have been replaced with "-any Table 1[2]A-1 effluent radiation monitor > column "UE". UE thresholds for all ANO continuously monitored gaseous and liquid release pathways are listed in Tables 1[2]A-1 to consolidate the information in a single location and, thereby, simplify identification of the thresholds by the EAL user. The values shown in Table 1[2]A-1 column "UE", consistent with the NEI bases, represent two times the ODCM release limits for gaseous and liquid releases. 2 Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer. to 0CAN031801 Page 18 of 120 NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 3 Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the (site-specific effluent release controlling document) limits for 60 minutes or longer. AU1.2 Sample analysis for a gaseous or liquid release indicates a concentration or release rate

> 2 x ODCM limits for 60 min. (Notes 1, 2) The ANO ODCM is the site-specific effluent release controlling document. Notes The Emergency Director should declare the Unusual Event promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded. If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 60 minutes. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. N/A Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock.

The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording.

None to 0CAN031801 Page 19 of 120 Table 1A-1 Unit 1 Effluent Monitor Classification Thresholds (2 min. avg reading) Release Point Monitor GE SAE Alert UE Gaseous Containment Purge RX-9820 (SPING 1)4.15E+01 Ci/cc 4.15E+00 µCi/cc 4.15E-01 µCi/cc 1.21E-03 µCi/cc Radwaste Area RX-9825 (SPING 2)2.67E+01 Ci/cc 2.67E+00 µCi/cc 2.67E-01 µCi/cc 4.94E-04 µCi/cc Fuel Handling Area RX-9830 (SPING 3)6.20E+02 Ci/cc 6.20E+01 µCi/cc 6.20E+00 µCi/cc 5.44E-04 µCi/cc Emergency Penetration Room RX-9835 (SPING 4)6.55E+02 Ci/cc 6.55E+01 µCi/cc 6.55E+00 µCi/cc 1.21E-02 µCi/cc Liquid Liquid Radwaste RE-4642 ---- ---- ---- 2.46E+05 cpm to 0CAN031801 Page 20 of 120 Table 2A-1 Unit 2 Effluent Monitor Classification Thresholds (2 min. avg reading) Release Point Monitor GE SAE Alert UE Gaseous Containment Purge 2RX-9820 (SPING 5)1.88E+01 Ci/cc 1.88E+00 µCi/cc 1.88E-01 µCi/cc 5.48E-04 µCi/cc Radwaste Area 2RX-9825 (SPING 6)2.35E+01 Ci/cc 2.35E+00 µCi/cc 2.35E-01 µCi/cc 4.35E-04 µCi/cc Fuel Handling Area 2RX-9830 (SPING 7)6.86E+02 Ci/cc 6.86E+01 µCi/cc 6.86E+00 µCi/cc 6.04E-04 µCi/cc Emergency Penetration Room 2RX-9835 (SPING 8)5.88E+02 Ci/cc 5.88E+01 µCi/cc 5.88E+00 µCi/cc 1.09E-02 µCi/cc Liquid BMS Liquid Discharge 2RE-2330---- ---- ---- 2.45E+04 cpm Regenerative Waste Discharge 2RE-4423---- ---- ---- 2.45E+05 cpm to 0CAN031801 Page 21 of 120 NEI IC# NEI IC Wording and Mode Applicability ANO IC#(s) ANO IC Wording and Mode Applicability Difference/Deviation Justification AU2 UNPLANNED loss of water level above irradiated fuel. MODE: All AU2 UNPLANNED loss of water level above irradiated fuel. MODE: All None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following: (site-specific level indications). AND b. UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors. (site-specific list of area radiation monitors) AU2.1 UNPLANNED water level drop in the REFUELING PATHWAY as indicated by low water level alarm, visual observation, or BWST[RWT] level drop due to makeup demands AND UNPLANNED rise in corresponding area radiation levels as indicated by any of the following radiation monitors: Unit 1 o RE-8009 Spent Fuel Area o RE-8017 Fuel Handling Area Unit 2 o 2RE-8914 Spent Fuel Area o 2RE-8915 Spent Fuel Area o 2RE-8916 Spent Fuel Area o 2RE-8912 Containment Incore Instrumentation Site-specific level indications incorporated. Site-specific area radiation monitors incorporated. to 0CAN031801 Page 22 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification AA1 Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE. MODE: All AA1 Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE. MODE: All None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer: (site-specific monitor list and threshold values) AA1.1 Reading on any Table 1[2]A-1 effluent radiation monitor > column "ALERT" for 15 min. (Notes 1, 2, 3, 4) The ANO radiation monitors that detect radioactivity effluent release to the environment are listed in Table 1[2]A-1. UE, Alert, SAE and GE thresholds for all ANO continuously monitored gaseous and liquid release pathways are listed in Table 1[2]A-1 to consolidate the information in a single location and, thereby, simplify identification of the thresholds by the EAL-user. 2 Dose assessment using actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond (site-specific dose receptor point). AA1.2 Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4) The site boundary is the site-specific receptor point. 3 Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond (site-specific dose receptor point) for one hour of exposure. AA1.3 Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2) The site boundary is the site-specific receptor point. to 0CAN031801 Page 23 of 120 NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 4 Field survey results indicate EITHER of the following at or beyond (site-specific dose receptor point): Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer. Analyses of field survey samples indicate thyroid CDE greater than 50 mrem for one hour of inhalation. AA1.4 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: Closed window dose rates > 10 mR/hr expected to continue for 60 min. Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation. (Notes 1, 2) The site boundary is the site-specific receptor point. Notes The Emergency Director should declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes. N/A Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. to 0CAN031801 Page 24 of 120 NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification Notes (cont.) If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. N/A Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. Note 4 The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. None

Incorporated site-specific EAL numbers associated with generic EAL#1. to 0CAN031801 Page 25 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification AA2 Significant lowering of water level above, or damage to, irradiated fuel. MODE: All AA2 Significant lowering of water level above, or damage to, irradiated fuel. MODE: All None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Uncovery of irradiated fuel in the REFUELING PATHWAY. AA2.1 IMMINENT uncovery of irradiated fuel in the REFUELING PATHWAY. Added the defined term "IMMINENT." Determination of irradiated fuel uncovery in the refueling pathway will always be an anticipatory determination as no direct indication is available to determine when the irradiated fuel has become uncovered. 2 Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY of the following radiation monitors: (site-specific listing of radiation monitors, and the associated readings, setpoints and/or alarms) AA2.2 Damage to irradiated fuel resulting in a release of radioactivity AND High alarm on any Table 1[2]A-2 radiation monitor. The EAL is clarified to refer to "mechanical" damage to irradiated fuel in alignment with the basis description. Site-specific list of radiation monitors incorporated. Radiation monitor high alarms specified. 3 Lowering of spent fuel pool level to (site-specific Level 2 value). [See Developer Notes] AA2.3 Lowering of spent fuel pool level to 387.0 ft. [389.5 ft.] (Alarm 2) on LIT-2020-3(4) [2LIT-2020-1(2)] For ANO, Level 2, which corresponds to ~10 ft. above the top of the fuel racks in the SFP, is an indicated level of: Unit 1: 387.0 ft. (Alarm 2) on LIT-2020-3(4)

Unit 2: 389.5 ft. (Alarm 2) on 2LIT-2020-1(2) to 0CAN031801 Page 26 of 120 Table 1A-2 Unit 1 Fuel Damage Radiation Monitors RE-8009 Spent Fuel Area RE-8017 Fuel Handling RE-8060 Containment High Range Radiation Monitors RE-8061 Containment High Range Radiation Monitors RX-9820 (SPING 1) Containment Purge RX-9825 (SPING 2) Radwaste Area RX-9830 (SPING 3) Fuel Handling Area Table 2A-2 Unit 2 Fuel Damage Radiation Monitors 2RE-8905 Containment Equipment Hatch Area 2RE-8909 Containment Personnel Access Area 2RE-8912 Containment Incore Inst. 2RE-8914 Spent Fuel Area 2RE-8915 Spent Fuel Area 2RE-8916 Spent Fuel Area 2RE-8925-1 Containment High Range Radiation Monitors 2RE-8925-2 Containment High Range Radiation Monitors 2RX-9820 (SPING 5) Containment Purge 2RX-9825 (SPING 6) Radwaste Area 2RX-9830 (SPING 7) Fuel Handling Area to 0CAN031801 Page 27 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification AA3 Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown MODE: All AA3 Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown. MODE: All (AA3.2 Modes 3 and 4 only) EAL AA3.2 mode applicability has been limited to the applicable mode of Table 1[2]A-3 (Modes 3 and 4). NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Dose rate greater than 15 mR/hr in ANY of the following areas: Control Room Central Alarm Station (other site-specific areas/rooms) AA3.1 Dose rate > 15 mR/hr in EITHER of the following areas: Control Room Central Alarm Station (by survey) No other site-specific areas requiring continuous occupancy exist at ANO. The Control Room envelope (Unit 1 and Unit 2) is monitored for excessive radiation by five detectors. These radiation detectors are RE-8001, 2RE-8001A, 2RE-8001B, 2RE-8750-1A, and 2RE-8750-1B. There are no permanently installed area radiation monitors in CAS that may be used to assess this EAL threshold. Therefore, this threshold is evaluated using local radiation survey for this area. to 0CAN031801 Page 28 of 120 NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 2 An UNPLANNED event results in radiation levels that prohibit or impede access to any of the following plant rooms or areas: (site-specific list of plant rooms or areas with entry-related mode applicability identified) AA3.2 An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table 1[2]A-3 room or area. (Note 5) The site-specific list of plant rooms or areas with entry-related mode applicability are tabularized in Tables 1[2]A-3. The bulleted bases item "the action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections)" was removed from the list of exceptions to classification in the basis information. These actions are a consideration when the site-specific list was developed. Rooms requiring entry for these types of actions are already excluded from the list when it was developed. Note If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. N/A Note 5 If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. None to 0CAN031801 Page 29 of 120 Table 1A-3 Unit 1 Safe Operation & Shutdown Rooms/Areas Room/Area Mode A-4 Switchgear Room 3, 4 Upper North Electrical Penetration Room 3, 4 Lower South Electrical Equipment Room 3, 4 Table 2A-3 Unit 2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Aux Building 317' Emergency Core Cooling Rooms 3, 4 Aux Building 317' Tendon Gallery Access 3, 4 Aux Building 335' Charging Pumps / MCC 2B-52 3, 4 Aux Building 354' MCC 2B-62 Area 3, 4 Emergency Diesel Generator Corridor 3, 4 Lower South Piping Penetration Room 3, 4 Aux Building 386' Containment Hatch 3, 4 to 0CAN031801 Page 30 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification AS1 Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE. MODE: All MODE: All AS1 Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE. MODE: All None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer: (site-specific monitor list and threshold values) AS1.1 Reading on any Table 1[2]A-1 effluent radiation monitor > column "SAE" for 15 min. (Notes 1, 2, 3, 4) The ANO radiation monitors that detect radioactivity effluent release to the environment are listed in Tables 1[2]A-1. UE, Alert, SAE and GE thresholds for all ANO continuously monitored gaseous and liquid release pathways are listed in Table 1[2]A-1 to consolidate the information in a single location and, thereby, simplify identification of the thresholds by the EAL-user. 2 Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond (site-specific dose receptor point) AS1.2 Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY. (Note 4) The site boundary is the site-specific receptor point. to 0CAN031801 Page 31 of 120 NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 3 Field survey results indicate EITHER of the following at or beyond (site-specific dose receptor point): Closed window dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer. Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation. AS1.3 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: Closed window dose rates > 100 mR/hr expected to continue for 60 min. Analyses of field survey samples indicate thyroid CDE > 500 mrem for 60 min. of inhalation. (Notes 1, 2) The site boundary is the site-specific receptor point. Notes The Emergency Director should declare the Site Area Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes. Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. to 0CAN031801 Page 32 of 120 NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification Notes (cont.) If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Note 4 The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. None Incorporated site-specific EAL numbers associated with generic EAL#1. to 0CAN031801 Page 33 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification AS2 Spent fuel pool level at (site-specific Level 3 description). MODE: All AS2 Spent fuel pool level at the top of the fuel racks. MODE: All Top of the fuel racks is the site-specific Level 3 description. NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Lowering of spent fuel pool level to (site-specific Level 3 value) AS2.1 Lowering of spent fuel pool level to 377.0 ft.[379.5 ft.] (Alarm 3) on LIT-2020-3(4)[2LIT-2020-1(2)]. For ANO, Level 3, which corresponds to the top of the fuel racks in the SFP, is an indicated level of: Unit 1: 377.0 ft. (Alarm 3) on LIT-2020-3(4) Unit 2: 379.5 ft. (Alarm 3) on 2LIT-2020-1(2) to 0CAN031801 Page 34 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification AG1 Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE. MODE: All AG1 Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE MODE: All None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer: (site-specific monitor list and threshold values) AG1.1 Reading on any Table A-1 effluent radiation monitor > column "GE" for 15 min. (Notes 1, 2, 3, 4) The ANO radiation monitors that detect radioactivity effluent release to the environment are listed in Tables 1[2]A-1. UE, Alert, SAE and GE thresholds for all ANO continuously monitored gaseous and liquid release pathways are listed in Tables 1[2]A-1 to consolidate the information in a single location and, thereby, simplify identification of the thresholds by the EAL-user. 2 Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond (site-specific dose receptor point). AG1.2 Dose assessment using actual meteorology indicates doses > 1000 mrem TEDE or 5000 mrem thyroid CDE at or beyond the SITE BOUNDARY. (Note 4) The site boundary is the site-specific receptor point. to 0CAN031801 Page 35 of 120 NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 3 Field survey results indicate EITHER of the following at or beyond (site-specific dose receptor point): Closed window dose rates greater than 1,000 mR/hr expected to continue for 60 minutes or longer. Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrem for one hour of inhalation. AG1.3 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: Closed window dose rates > 1000 mR/hr expected to continue for 60 min. Analyses of field survey samples indicate thyroid CDE > 5000 mrem for 60 min. of inhalation. (Notes 1, 2) The site boundary is the site-specific receptor point. to 0CAN031801 Page 36 of 120 NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification Notes The Emergency Director should declare the Site Area Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes for classification purposes. Note 4 The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock.

The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording.

None

Incorporated site-specific EAL numbers associated with generic EAL#1. to 0CAN031801 Page 37 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification AG2 Spent fuel pool level cannot be restored to at least (site-specific Level 3 description) for 60 minutes or longer. MODE: All AG2 Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer. MODE: All Top of the fuel racks is the site-specific Level 3 description. NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Spent fuel pool level cannot be restored to at least (site-specific Level 3 value) for 60 minutes or longer. AG2.1 Spent fuel pool level cannot be restored to at least 377.0 ft.[379.5 ft.] (Alarm 3) on LIT-2020-3(4) [2LIT-2020-1(2)] for 60 min. (Note 1) For ANO, Level 3, which corresponds to the top of the fuel racks in the SFP, is an indicated level of: Unit 1: 377.0 ft. (Alarm 3) on LIT-2020-3(4)

Unit 2: 379.5 ft. (Alarm 3) on 2LIT-2020-1(2) Note The Emergency Director should declare the General Emergency promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded. N/A Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock.

to 0CAN031801 Page 38 of 120 Category C Cold Shutdown / Refueling System Malfunction NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification CU1 UNPLANNED loss of (reactor vessel/RCS [PWR] or RPV

[BWR]) inventory for 15 minutes or longer. MODE: Cold Shutdown, Refueling CU1 UNPLANNED loss of RCS inventory. MODE: 5 - Cold Shutdown, 6 - Refueling Deleted the words "-for 15 minutes or longer" as the 15 minute criteria only applies to EAL #1. NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 UNPLANNED loss of reactor coolant results in (reactor vessel/RCS [PWR] or RPV [BWR]) level less than a required lower limit for 15 minutes or longer. CU1.1 UNPLANNED loss of reactor coolant results in RCS water level less than a required lower limit for 15 min. (Note 1) None 2 a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level cannot be monitored. AND b. UNPLANNED increase in (site-specific sump and/or tank) levels. CU1.2 RCS level cannot be monitored AND EITHER UNPLANNED rise in any Table 1[2]C-1 sump/tank level due to loss of RCS inventory Visual observation of UNISOLABLE RCS leakage Added the words "-due to loss of RCS inventory" to be consistent with the IC wording. Replaced the term "increase" with the word "rise" consistent with allowed usage. Site-specific applicable sumps and tanks are listed in Table 1[2]C-1 to improve the readability of the EAL. Added bulleted criteria "Visual observation" to include direct observation of significant RCS unisolable leakage. to 0CAN031801 Page 39 of 120 NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification Note The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. N/A Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock. Table 1C-1 Unit 1 Sumps / Tanks Reactor Building Sump Reactor Drain Tank Aux. Building Equipment Drain Tank Aux. Building Sump Quench Tank Table 2C-1 Unit 2 Sumps / Tanks CNTMT Sump Reactor Drain Tank LRW Waste Tank (2T-20) Holdup Tank Aux. Building Sump Quench Tank to 0CAN031801 Page 40 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification CU2 Loss of all but one AC power source to emergency buses for 15 minutes or longer. MODE: Cold Shutdown, Refueling, Defueled CU2 Loss of all but one AC power source to vital buses for 15 minutes or longer. MODE: 5 - Cold Shutdown, 6 - Refueling, DEF - Defueled "vital buses" is the ANO-specific terminology for "emergency buses". NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 a. AC power capability to (site-specific emergency buses) is reduced to a single power source for 15 minutes or longer. AND b. Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS. CU2.1 AC power capability, Table 1[2]C-3, to vital 4.16 KV buses A3[2A3] and A4[2A4] reduced to a single power source for 15 min. (Note 1) AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS 4.16KV buses A3[2A3] and A4[2A4] are the emergency (vital) buses. Site-specific AC power sources are tabularized in Table 1[2]C-3. Note The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. N/A Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock. to 0CAN031801 Page 41 of 120 Table 1C-3 Unit 1 AC Power Sources Offsite Startup Transformer No. 1 Startup Transformer No. 2 Unit Auxiliary Transformer (from 22 KV switchyard) Onsite DG1 DG2 AAC Gen Table 2C-3 Unit 2 AC Power Sources Offsite Startup Transformer No. 3 Startup Transformer No. 2 Unit Auxiliary Transformer (backfed from main transformer) Onsite 2DG1 2DG2 AAC Gen to 0CAN031801 Page 42 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification CU3 Unplanned increase in RCS temperature. MODE: Cold Shutdown, Refueling CU3 Unplanned rise in RCS temperature MODE: 5 - Cold Shutdown, 6 - Refueling Replaced the term "increase" with the word "rise" consistent with allowed usage. NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 UNPLANNED increase in RCS temperature to greater than (site-specific Technical Specification cold shutdown temperature limit) CU3.1 UNPLANNED rise in RCS temperature to > 200°F due to loss of decay heat removal capability Replaced the term "increase" with the word "rise" consistent with allowed usage. 200°F is the site-specific Tech. Spec. cold shutdown temperature limit. Added "due to loss of decay heat removal capability" to reinforce the generic bases that states "EAL #1 involves a loss of decay heat removal capability" 2 Loss of ALL RCS temperature and (reactor vessel/RCS [PWR] or RPV

[BWR]) level indication for 15 minutes or longer. CU3.2 Loss of all RCS temperature and RCS level indication for 15 min. (Note 1) None Note The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded N/A Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock. to 0CAN031801 Page 43 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification CU4 Loss of Vital DC power for 15 minutes or longer. MODE: Cold Shutdown, Refueling CU4 Loss of Vital DC power for 15 minutes or longer. MODE: 5 - Cold Shutdown, 6 - Refueling None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Indicated voltage is less than (site-specific bus voltage value) on required Vital DC buses for 15 minutes or longer. CU4.1 Indicated voltage is < 105 VDC on required vital 125 VDC buses for 15 min. (Note 1) 105 VDC is the site-specific minimum vital DC bus voltage. Safety-related DC bus operability requirements are specified in Technical Specifications. Note The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. N/A Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock. to 0CAN031801 Page 44 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification CU5 Loss of all onsite or offsite communications capabilities. MODE: Cold Shutdown, Refueling, Defueled CU5 Loss of all onsite or offsite communications capabilities. MODE: 5 - Cold Shutdown, 6 - Refueling, DEF - Defueled None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Loss of ALL of the following onsite communication methods: (site specific list of communications methods) CU5.1 Loss of all Table 1[2]C-5 onsite communication methods OR Loss of all Table 1[2]C-5 State and local agency communication methods OR Loss of all Table 1[2]C-5 NRC communication methods Example EALs #1, 2 and 3 have been combined into a single EAL for simplification of presentation. Replaced "ORO" with "State and local agency" for clarification. Table 1[2]C-5 provides a site-specific list of onsite, State and local agency (ORO) and NRC communications methods. 2 Loss of ALL of the following ORO communications methods: (site specific list of communications methods) 3 Loss of ALL of the following NRC communications methods: (site specific list of communications methods) to 0CAN031801 Page 45 of 120 Table 1[2]C-5 Communication Methods System Onsite State/Local NRC Station radio system X ANO plant phone system X Gaitronics X Telephone Systems: Commercial Microwave Satellite VOIP X X INFORM Notification System X Emergency Notification System (ENS) X to 0CAN031801 Page 46 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification CA1 Loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory MODE: Cold Shutdown, Refueling CA1 Significant loss of RCS inventory MODE: 5 - Cold Shutdown, 6 - Refueling Added the word "Significant-" to differentiate the Alert loss of RCS inventory IC from the Unusual Event IC which is "Unplanned loss of RCS inventory." NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory as indicated by level less than (site-specific level). CA1.1 Loss of RCS inventory as indicated by EITHER: RVLMS Levels 1 through 8 [1 through 5] indicate DRY Reactor vessel level 368.5 ft. (LT-1195/LT-1196)[0 in. (L4791/L4792)] (bottom of hot leg) RVLMS Levels 1 through 8 [RVLMS Levels 1 through 5] indicating DRY is the site-specific reactor vessel level corresponding to lowest RVLMS indicated level above the bottom of the RCS Hot Leg. Reactor vessel level 368.5 ft.[0 in.] corresponds to the bottom of RCS Hot Leg. 2 a. (Reactor vessel/RCS [PWR] or RPV [BWR])

level cannot be monitored for 15 minutes or longer AND b. UNPLANNED increase in (site-specific sump and/or tank) levels due to a loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory. CA1.2 RCS level cannot be monitored for 15 min. (Note 1) AND EITHER UNPLANNED rise in any Table 1[2]C-1 Sump / Tank level due to a loss of RCS inventory Visual observation of UNISOLABLE RCS leakage Site-specific applicable sumps and tanks are listed in Table 1[2]C-1 to improve the readability of the EAL. Replaced the term "increase" with the word "rise" consistent with allowed usage. Added bulleted criteria "Visual observation" to include direct observation of significant RCS leakage. to 0CAN031801 Page 47 of 120 NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification Note The Emergency Director should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. N/A Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock.

to 0CAN031801 Page 48 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification CA2 Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer. MODE: Cold Shutdown, Refueling, Defueled CA2 Loss of all offsite and all onsite AC power to vital buses for 15 minutes or longer. MODE: 5 - Cold Shutdown, 6 - Refueling, DEF - Defueled "Vital buses" is the ANO-specific terminology for "emergency buses". NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Loss of ALL offsite and ALL onsite AC Power to (site-specific emergency buses) for 15 minutes or longer. CA2.1 Loss of all offsite and all onsite AC power to vital 4.16 KV buses A3[2A3] and A4[2A4] for 15 min. (Note 1) 4.16KV buses A3[2A3] and A4[2A4] are the site-specific emergency buses. Note The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. N/A Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock. to 0CAN031801 Page 49 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification CA3 Inability to maintain the plant in cold shutdown. MODE: Cold Shutdown, Refueling CA3 Inability to maintain plant in cold shutdown. MODE: 5 - Cold Shutdown, 6 - Refueling None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 UNPLANNED increase in RCS temperature to greater than (site-specific Technical Specification cold shutdown temperature limit) for greater than the duration specified in the following table. CA3.1 UNPLANNED rise in RCS temperature to > 200°F for > Table 1[2]C-4 duration (Note 1) OR Unplanned RCS pressure rise > 10 psig (this EAL does not apply during water-solid plant conditions) Example EALs #1 and #2 have been combined into a single EAL as EAL #2 is the alternative threshold based on a loss of RCS temperature indication. Replaced the term "increase" with the word "rise" consistent with allowed usage 200°F is the site-specific Tech. Spec. cold shutdown temperature limit. Table 1[2]C-3 is the site-specific implementation of the generic RCS Reheat Duration Threshold table. 10 psig is the site-specific pressure rise readable by Control Room indications. 2 UNPLANNED RCS pressure increase greater than (site-specific pressure reading). (This EAL does not apply during water-solid plant conditions. [PWR]) Note The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. N/A Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock. to 0CAN031801 Page 50 of 120 Table: RCS Heat-up Duration Thresholds RCS Status Containment Closure Status Heat-up Duration Intact (but not at reduced inventory [PWR]) Not applicable 60 minutes* Not intact (or at reduced inventory [PWR]) Established 20 minutes* Not Established 0 minutes

• If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable. Table 1[2]C-4: RCS Heat-up Duration Thresholds RCS Status CONTAINMENT CLOSURE Status Heat-up Duration Intact (but not lowered inventory) N/A 60 min.* Not intact OR lowered inventory established 20 min.* not established 0 min.

• If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.

to 0CAN031801 Page 51 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification CA6 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. MODE: Cold Shutdown, Refueling CA6 Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode. MODE: 5 - Cold Shutdown, 6 - Refueling Pluralized safety systems to be consistent with NRC EP FAQ 2016-002 that specifies degraded performance or visible damage in more than one safety system train.

to 0CAN031801 Page 52 of 120 NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 a. The occurrence of ANY of the following hazardous events: Seismic event (earthquake) Internal or external flooding event High winds or tornado strike FIRE EXPLOSION (site-specific hazards) Other events with similar hazard characteristics as determined by the Shift Manager AND b. EITHER of the following: 1. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode. OR 2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode. CA6.1 The occurrence of any Table 1[2]C-6 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER: Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 10, 11) The hazardous events have been tabularized in Table 1[2]C-6. CA6.1 reflects NRC FAQ 2016-002 requiring degraded performance or visible damage to more than one train of a safety system caused by the specified events. This wording is a deviation from NEI 99-01, Revision 6, CA6 generic wording and bases but is deemed acceptable in order to ensure that an Alert is declared only when a hazardous event causes actual or potential performance issues with safety systems. This is consistent with NRC-approved EP FAQ 2016-002. The word "a" is replaced with "the" in the FAQ wording to provide agreement with the FAQ basis information indicating that the criteria is applicable to another train of the same safety system. to 0CAN031801 Page 53 of 120 N/A N/A N/A Note 10: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted Added Note 10 consistent with the recommendation of NRC EP FAQ 2016-002. N/A N/A N/A Note 11: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted. Added Note 11 consistent with the recommendation of NRC EP FAQ 2016-002.

Table 1[2]C-6 Hazardous Events Seismic event (earthquake) Internal or external FLOODING event High winds or tornado strike FIRE EXPLOSION Other events with similar hazard characteristics as determined by the Shift Manager to 0CAN031801 Page 54 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification CS1 Loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory affecting core decay heat removal capability. MODE: Cold Shutdown, Refueling CS1 Loss of RCS inventory affecting core decay heat removal capability MODE: 5 - Cold Shutdown, 6 - Refueling None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 a. CONTAINMENT CLOSURE not established. AND b. (Reactor vessel/RCS [PWR] or RPV [BWR]) level less than (site-specific level). CS1.1 CONTAINMENT CLOSURE not established AND RVLMS Levels 1 through 9 [1 through 6] indicate DRY Unit 1: RVLMS Level 9 is at elevation 367.69 ft. The bottom of the hotleg is at 368 ft. RVLMS Level 9 corresponds to approximately 6" below the bottom of the hotleg. Unit 2: The top of active fuel is 61.4 in. below the bottom of the hotleg. RVLMS Level 6 is 47 in. above the top of active fuel, which is 61.4 in. - 47 in. = 14.4 in.

below the bottom ID of the hot leg. 2 a. CONTAINMENT CLOSURE established. AND b. (Reactor vessel/RCS [PWR] or RPV [BWR])

level less than (site-specific level). N/A N/A Unit 1 cannot measure reactor vessel at or near the top of active fuel or below. The Unit 2 RVLMS does not provide a positive indication of level at or near the top of active fuel, but does provide an indication of core uncovery at Level 7. Consistent with the generic developers guidance: "If the design and operation of water level instrumentation is such that this level value cannot be determined at any time during Cold Shutdown or Refueling modes, then do not include EAL #2 (classification will be accomplished in accordance with EAL #3)." Unit 2 RVLMS levels 7 through 11 are indications of core uncovery. to 0CAN031801 Page 55 of 120 NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 3 a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level cannot be monitored for 30 minutes or longer. AND b. Core uncovery is indicated by ANY of the following: (Site-specific radiation monitor) reading greater than (site-specific value) Erratic source range monitor indication

[PWR] UNPLANNED increase in (site-specific sump and/or tank) levels of sufficient magnitude to indicate core uncovery (Other site-specific indications) CS1.2 [1. RVLMS Levels 1 through 7 indicate DRY] [OR] [2] RCS level cannot be monitored for 30 min. (Note 1) AND Core uncovery is indicated by any of the following: UNPLANNED rise in any Table 1[2]C-1 sump/tank level of sufficient magnitude to indicate core uncovery Containment high range radiation monitor RE-8060/8061 [2RE-8925-1/8925-2] reading > 10 R/hr Erratic Source Range Monitor indication On Unit 2, RVLMS level 7 is an indication of core uncovery when the RVLMS system is in service. An ANO CS1.2 EAL threshold is provided for Unit 2 only where there is indication of core uncovery without applying the 30-minute time period because of the known core uncovery condition. Replaced the term "increase" with the word "rise" consistent with allowed usage. Table 1[2]C-1 provides a tabularized list of site-specific applicable sumps and tanks. Containment High Range Radiation Monitors RE-8060/8061[2RE-8925-1/8925-2] are the site-specific radiation monitors that would be indicative of likely core uncovery in the Refueling mode. The dose rate due to core shine when the top of the core becomes uncovered should result in dose rates > 10 R/hr. Note The Emergency Director should declare the Site Area Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded N/A Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock. to 0CAN031801 Page 56 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification CG1 Loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory affecting fuel clad integrity with containment challenged MODE: Cold Shutdown, Refueling CG1 Loss of RCS inventory affecting fuel clad integrity with containment challenged MODE: 5 - Cold Shutdown, 6 - Refueling None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level less than (site-specific level) for 30 minutes or longer. AND b. ANY indication from the Containment Challenge Table (see below). N/A N/A Unit 1 cannot measure reactor vessel at or near the top of active fuel or below. The Unit 2 RVLMS does not provide a positive indication of level at or near the top of active fuel, but does provide an indication of core uncovery at Level 7. Consistent with the generic developers guidance: "If the design and operation of water level instrumentation is such that this level value cannot be determined at any time during Cold Shutdown or Refueling modes, then do not include EAL #2 (classification will be accomplished in accordance with EAL #3)." Unit 2 RVLMS levels 7 through 11 are indications of core uncovery. to 0CAN031801 Page 57 of 120 NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 2 a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level cannot be monitored for 30 minutes or longer. AND b. Core uncovery is indicated by ANY of the following: (Site-specific radiation monitor) reading greater than (site-specific value) Erratic source range monitor indication [PWR] UNPLANNED increase in (site-specific sump and/or tank) levels of sufficient magnitude to indicate core uncovery (Other site-specific indications) AND c. ANY indication from the Containment Challenge Table (see below). CG1.1 RVLMS Levels 1 through 7 indicate DRY AND Any Containment Challenge indication, Table 1[2]C-2 On Unit 2, RVLMS level 7 is an indication of core uncovery when the RVLMS system is in service. An ANO CG1.1 EAL threshold is provided for Unit 2 only where there is indication of core uncovery without applying the 30-minute time period because of the known core uncovery condition. Entergy EAL CG1.2 is provided for both units. Replaced the term "increase" with the word "rise" consistent with allowed usage. Table 1[2]C-2 provides a tabularized list of containment challenge indications. Containment High Range Radiation Monitors RE-8060/8061 [2RE-8925-1/8925-2] are the site-specific radiation monitors that would be indicative of likely core uncovery in the Refueling mode. The dose rate due to core shine when the top of the core becomes uncovered should result in dose rates > 10 R/hr. 3% hydrogen concentration in the presence of oxygen represents an explosive mixture in containment. to 0CAN031801 Page 58 of 120 NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 2 a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level cannot be monitored for 30 minutes or longer. AND b. Core uncovery is indicated by ANY of the following: (Site-specific radiation monitor) reading greater than (site-specific value) Erratic source range monitor indication [PWR] UNPLANNED increase in (site-specific sump and/or tank) levels of sufficient magnitude to indicate core uncovery (Other site-specific indications) AND c. ANY indication from the Containment Challenge Table (see below). CG1.2 RCS level cannot be monitored for 30 min. (Note 1) AND Core uncovery is indicated by any of the following: UNPLANNED rise in any Table 1[2] C-1 sump/tank level of sufficient magnitude to indicate core uncovery Containment High Range Radiation Monitor RE-8060/8061 [2RE-8925-1/8925-2] reading

> 10 R/hr Erratic Source Range Monitor indication AND Any Containment Challenge indication, Table 1[2]C-2 On Unit 2, RVLMS level 7 is an indication of core uncovery when the RVLMS system is in service. Entergy EAL CG1.1 is provided for Unit 2. Entergy EAL CG1.2 is provided for both units. Replaced the term "increase" with the word "rise" consistent with allowed usage. Table 1[2]C-1 provides a tabularized list of site-specific applicable sumps and tanks. Table 1[2]C-2 provides a tabularized list of containment challenge indications. Containment High Range Radiation Monitors RE-8060/8061[2RE-8925-1/8925-2] are the site-specific radiation monitors that would be indicative of likely core uncovery in the Refueling mode. The dose rate due to core shine when the top of the core becomes uncovered should result in dose rates > 10 R/hr. 3% hydrogen concentration in the presence of oxygen represents an explosive mixture in containment. to 0CAN031801 Page 59 of 120 NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification Note The Emergency Director should declare the General Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.

N/A N/A Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock. Note 6 implements the asterisked note associated with the Containment Closure requirement. to 0CAN031801 Page 60 of 120 Containment Challenge Table CONTAINMENT CLOSURE not established* (Explosive mixture) exists inside containment UNPLANNED increase in containment pressure Secondary containment radiation monitor reading above (site-specific value) [BWR]

• If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required. Table 1[2]C-2 Containment Challenge Indications CONTAINMENT CLOSURE not established (Note 6) Containment hydrogen concentration > 3% UNPLANNED rise in containment pressure to 0CAN031801 Page 61 of 120 Category D Permanently Defueled Station Malfunction NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification PD-AU1 PD-AU2 PD-SU1 PD-HU1 PD-HU2 PD-HU3 PD-AA1 PD-AA2 PD-HA1 PD-HA3 Recognition Category D Permanently Defueled Station N/A N/A NEI Recognition Category PD ICs and EALs are applicable only to permanently defueled stations. ANO is not a defueled station.

to 0CAN031801 Page 62 of 120 Category E Independent Spent Fuel Storage Installation (ISFSI) NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification E-HU1 Damage to a loaded cask confinement BOUNDARY. MODE: All EU1 Damage to a loaded cask confinement BOUNDARY. MODE: All None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Damage to a loaded cask confinement BOUNDARY as indicated by an on-contact radiation reading greater than (2 times the site-specific cask specific technical specification allowable radiation level) on the surface of the spent fuel cask. EU1.1 Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask (VSC-24 VCC or HI-STORM overpack) > any Table 1[2]E-1 value The specified dose rates represent 2 times the site-specific cask technical specification allowable levels per the ISFSI Technical Specifications (licensing document). Table 1[2]E-1 ISFSI Dose Rates VSC-24 VCC HI-STORM 200 mrem/hr on the sides 400 mrem/hr on the top 700 mrem/hr at the air inlet 200 mrem/hr at the air outlet 60 mrem/hr (gamma + neutron) on the top or outlet vent 600 mrem/hr (gamma + neutron on the side of the side of the overpack (excluding inlet and outlet ducts) to 0CAN031801 Page 63 of 120 Category F Fission Product Barrier Degradation NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification FA1 Any Loss or any Potential Loss of either the Fuel Clad or RCS barrier. MODE: Power Operation, Hot Standby, Startup, Hot Shutdown FA1 Any loss or any potential loss of either Fuel Clad or RCS barrier MODE: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Any Loss or any Potential Loss of either the Fuel Clad or RCS barrier. FA1.1 Any loss or any potential loss of either Fuel Clad or RCS (Table 1[2]F-1). Table 1[2]F-1 provides the fission product barrier loss and potential loss thresholds. to 0CAN031801 Page 64 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification FS1 Loss or Potential Loss of any two barriers. MODE: Power Operation, Hot Standby, Startup, Hot Shutdown FS1 Loss or potential loss of any two barriers. MODE: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Loss or Potential Loss of any two barriers. FS1.1 Loss or potential loss of any two barriers (Table 1[2]F-1). Table 1[2]F-1 provides the fission product barrier loss and potential loss thresholds. to 0CAN031801 Page 65 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification FG1 Loss of any two barriers and Loss or Potential Loss of third barrier. MODE: Power Operation, Hot Standby, Startup, Hot Shutdown FG1 Loss of any two barriers and loss or potential loss of the third barrier. MODE: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Loss of any two barriers and Loss or Potential Loss of third barrier. FG1.1 Loss of any two barriers AND Loss or potential loss of the third barrier (Table 1[2]F-1). Table 1[2]F-1 provides the fission product barrier loss and potential loss thresholds. to 0CAN031801 Page 66 of 120 PWR Fuel Clad Fission Product Barrier Degradation Thresholds NEI FPB# NEI Threshold Wording ANO FPB #(s) ANO FPB Wording Difference/Deviation Justification FC Loss 1 RCS or SG Tube Leakage Not Applicable N/A N/A N/A FC Loss 2 Inadequate Heat Removal A. Core exit thermocouple readings greater than (site-specific temperature value). FCB2 CETs > 1200°F. Consistent with the generic developers notes 1200°F CET temperature is used. FC Loss 3 RCS Activity/CMNT Rad A. Containment radiation monitor reading greater than (site-specific value) OR B. (Site-specific indications that reactor coolant activity is greater than 300 Ci/gm dose equivalent I-131). FCB5 Containment high range radiation monitor RE-8060/8061 [2RE-8925-1/8925-2] > 750[700] R/hr. A 750[700] R/hr (768[682] R/hr rounded for readability) reading in the containment is used to indicate a loss of the Fuel Clad barrier and a release of reactor coolant, with elevated activity indicative of fuel damage, into the containment. This value assumes an instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of approximately 300 µCi/gm Dose Equivalent I-131 into the containment atmosphere. FCB6 Coolant activity > 300 µCi/gm dose equivalent I-131. 300 µCi/gm DEI-131 is the site-specific indication for this reactor coolant activity. FC Loss 4 CNMT Integrity or Bypass Not Applicable N/A N/A N/A FC Loss 5 Other Indications A. (site-specific as applicable) N/A N/A No other site-specific Fuel Clad Loss indication has been identified for ANO. FC Loss 6 ED Judgment A. ANY condition in the opinion of the Emergency Director that indicates Loss of the Fuel Clad barrier. FCB7 Any condition in the opinion of the Emergency Director that indicates loss of the fuel clad barrier. None to 0CAN031801 Page 67 of 120 NEI FPB# NEI Threshold Wording ANO FPB #(s) ANO FPB Wording Difference/Deviation Justification FC P-Loss 1 RCS or SG Tube Leakage A. RCS/reactor vessel level less than (site-specific level) FCB1 RVLMS Levels 1 through 9 [1 through 7] indicate DRY. The above core level indication on Unit 1 is used to monitor the approach to and recovery from ICC conditions, but the CETs are used to identify core uncovery, and are the only positive indication of core uncovery. The reactor vessel level indicators installed in Unit 1 extend from the top of the reactor vessel to the fuel alignment plate and indicate that the lowest sensor is greater than 2 feet above the top of active fuel. For Unit 2, RVLMS level 7 is an indication of core uncovery. FC P-Loss 2 Inadequate Heat Removal A. Core exit thermocouple readings greater than (site-specific temperature value) OR B. Inadequate RCS heat removal capability via steam generators as indicated by (site-specific indications). FCB3 CETs > 700°F. Consistent with the generic developers notes 700°F CET temperature is used. FCB4 RCS heat removal cannot be established using steam generators AND HPI[Once Through] cooling initiated. The initiation of HPI[Once Through] cooling is a readily identifiable procedurally driven action that is taken when steam generators are not functioning as an effective RCS heat removal source. FC P-Loss 3 RCS Activity/CMNT Rad Not Applicable N/A N/A N/A FC P-Loss 4 CNMT Integrity or Bypass Not Applicable N/A N/A N/A FC P-Loss 5 Other Indications A. (site-specific as applicable) N/A N/A No other site-specific Fuel Clad Potential Loss indication has been identified for ANO. to 0CAN031801 Page 68 of 120 NEI FPB# NEI Threshold Wording ANO FPB #(s) ANO FPB Wording Difference/Deviation Justification FC P-Loss 6 Emergency Director Judgment A. Any condition in the opinion of the Emergency Director that indicates Potential Loss of the Fuel Clad barrier. FCB8 Any condition in the opinion of the Emergency Director that indicates potential loss of the Fuel Clad barrier None to 0CAN031801 Page 69 of 120 PWR RCS Fission Product Barrier Degradation Thresholds NEI FPB# NEI IC Wording ANO FPB #(s) ANO FPB Wording Difference/Deviation Justification RCS Loss 1 RCS or SG Tube Leakage A. An automatic or manual ECCS (SI) actuation is required by EITHER of the following: 1. UNISOLABLE RCS leakage OR 2. SG tube RUPTURE. RCB1 An automatic or manual ESAS [ESFAS] actuation required by EITHER: UNISOLABLE RCS leakage SG tube RUPTURE ESAS[ESFAS] is the site-specific nomenclature for ECCS (SI). Added "Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification" to the basis. This provides agreement with the definition of "Unisolable" and ensures isolation attempts, both locally and remotely, are achieved in a timely manner. RCS Loss 2 Inadequate Heat Removal Not Applicable N/A N/A N/A RCS Loss 3 RCS Activity/CMNT Rad A. Containment radiation monitor reading greater than (site-specific value). RCB5 Containment high range radiation monitor RE-8060/8061

[2RE-8925-1/8925-2] > 40[50] R/hr. A 40[50] R/hr (42.8[50.4] R/hr rounded for readability) reading in containment is used to indicate a loss of the RCS barrier and a release of reactor coolant, at the T.S. coolant activity limit, into the containment. This value assumes an instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated T.S. coolant activity into the containment atmosphere. Site-specific information added to the basis on the phenomenon of thermally induced current and its potential effects on the radiation monitor indication. RCS Loss 4 CNMT Integrity or Bypass Not Applicable N/A N/A N/A RCS Loss 5 Other Indications A. (site-specific as applicable). N/A N/A No other site-specific RCS Loss indication has been identified for ANO. to 0CAN031801 Page 70 of 120 NEI FPB# NEI IC Wording ANO FPB #(s) ANO FPB Wording Difference/Deviation Justification RCS Loss 6 Emergency Director Judgment A. ANY condition in the opinion of the Emergency Director that indicates Loss of the RCS barrier. RCB6 Any condition in the opinion of the Emergency Director that indicates loss of the RCS barrier. None RCS P-Loss 1 RCS or SG Tube Leakage A. Operation of a standby charging (makeup) pump is required by EITHER of the following: 1. UNISOLABLE RCS leakage OR 2. SG tube leakage. OR RCB2 UNISOLABLE RCS leakage or S/G tube leakage > 50[44] gpm excluding normal reductions in RCS inventory (e.g., letdown, RCP seal leakoff) ANO uses the alternative indication of the capacity of one pump for the threshold leakage value as described in the NEI guidance. Added "Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification" to the basis. This provides agreement with the definition of "Unisolable" and ensures isolation attempts, both locally and remotely, are achieved in a timely manner. to 0CAN031801 Page 71 of 120 NEI FPB# NEI IC Wording ANO FPB #(s) ANO FPB Wording Difference/Deviation Justification B. RCS cooldown rate greater than (site-specific pressurized thermal shock criteria/limits defined by site-specific indications). RCB3 UNISOLABLE RCS leakage or S/G tube leakage > 50[44] gpm excluding normal reductions in RCS inventory (e.g., letdown, RCP seal leakoff) Unit 1: PTS limits apply (RT14) AND RCS pressure and temperature are left of the NDTT/LTOP limit lines on EOP Figure 3 (Note 12) Unit 2: Uncontrolled RCS cooldown (50°F step change which is below 500°F from NOT) AND RCS pressure and temperature are to the left of line B (200 degrees MTS), Standard Attachment 1, P-T Limits (Note 12) Unit 1 and Unit 2 specific PTS criteria is specified. Note 12 - "Once PTS limits are first invoked, if RCS temperature and pressure are not brought within the limits within 15 minutes, this threshold is met and an immediate declaration is warranted. This threshold is met immediately upon exceeding the limits after this initial 15 minute period until PTS limits no longer apply" is added to allow RCS parameters to be brought within the limits in a timely manner when they are first outside the limits. RCS parameters must be maintained within afterward. RCS P-Loss 2 Inadequate Heat Removal A. Inadequate RCS heat removal capability via steam generators as indicated by (site-specific indications). RCB4 RCS heat removal cannot be established AND HPI[Once Through] cooling initiated. The initiation of HPI[Once Through] cooling is a readily identifiable procedurally driven action that is taken when steam generators are not functioning as an effective RCS heat removal source. RCS P-Loss 3 CS Activity/CMNT Rad Not Applicable N/A N/A N/A RCS P-Loss 4 CNMT Integrity or Bypass Not Applicable N/A N/A N/A to 0CAN031801 Page 72 of 120 NEI FPB# NEI IC Wording ANO FPB #(s) ANO FPB Wording Difference/Deviation Justification RCS P-Loss 5 Other Indications A. (site-specific as applicable) N/A N/A No other site-specific RCS Potential Loss indication has been identified for ANO. RCS P-Loss 6 Emergency Director Judgment A. ANY condition in the opinion of the Emergency Director that indicates Potential Loss of the RCS barrier. RCB7 Any condition in the opinion of the Emergency Director that indicates potential loss of the RCS barrier None to 0CAN031801 Page 73 of 120 PWR Containment Fission Product Barrier Degradation Thresholds NEI FPB# NEI IC Wording ANO FPB #(s) ANO FPB Wording Difference/Deviation Justification CNMT Loss 1 RCS or SG Tube Leakage A. A leaking or RUPTURED SG is FAULTED outside of containment. CNB1 A S/G that is leaking > 50[44] gpm (excluding normal reductions in RCS inventory) or that is RUPTURED is also FAULTED outside of containment The threshold is reworded to clarify that the steam generator leakage is that level of leakage associated with the related the NEI RCS barrier potential loss threshold 1.A (ANO RCB2). Revised the table on page 2 of the basis information to reflect pump capacity values to align with the change made to the EAL. CNMT Loss 2 Inadequate Heat Removal Not Applicable N/A N/A N/A CNMT Loss 3 RCS Activity/CMNT Rad Not applicable N/A N/A N/A CNMT Loss 4 CNMT Integrity or Bypass A. Containment isolation is required AND EITHER of the following: 1. Containment integrity has been lost based on Emergency Director judgment. OR 2. UNISOLABLE pathway from the containment to the environment exists. OR B. Indications of RCS leakage outside of containment. CNB4 Containment isolation is required AND EITHER: Containment integrity has been lost based on Emergency Director judgment UNISOLABLE pathway from Containment to the environment exists Added "Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification" to the basis. This provides agreement with the definition of "Unisolable" and ensures isolation attempts, both locally and remotely, are achieved in a timely manner. CNB5 Indications of RCS leakage outside of containment None to 0CAN031801 Page 74 of 120 NEI FPB# NEI IC Wording ANO FPB #(s) ANO FPB Wording Difference/Deviation Justification CNMT Loss 5 Other Indications A. (site-specific as applicable) N/A N/A No other site-specific Containment Loss indication has been identified for ANO. CNMT Loss 6 Emergency Director Judgment ANY condition in the opinion of the Emergency Director that indicates Loss of the Containment barrier. CNB9 Any condition in the opinion of the Emergency Director that indicates loss of the containment barrier. None CNMT P-Loss 1 RCS or SG Tube Leakage Not Applicable N/A N/A N/A CNMT P-Loss 2 Inadequate Heat Removal A. 1. (Site-specific criteria for entry into core cooling restoration procedure) AND 2. Restoration procedure not effective within 15 minutes. CNB2 CETs > 1200°F AND Restoration procedures not effective within 15 min. (Note 1) Consistent with the generic developers notes 1200°F CET temperature is used. Added Note 1 consistent with other thresholds with a timing component. CNMT P-Loss 3 RCS Activity/CMNT Rad A. Containment radiation monitor reading greater than (site-specific value). CNB3 Containment high range radiation monitor RE-8060/8061 [2RE-8925-1/8925-2] reading

> 10,000[12,000] R/hr 10,000[12,000] R/hr (10,300[12,100] R/hr rounded for readability) reading in the containment is used to indicate a potential loss of the containment barrier and a release of reactor coolant, with significant activity indicative of 20% fuel damage, into the containment. This value assumes an instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration associated with 20% clad damage into the containment atmosphere. to 0CAN031801 Page 75 of 120 NEI FPB# NEI IC Wording ANO FPB #(s) ANO FPB Wording Difference/Deviation Justification CNMT P-Loss 4 CNMT Integrity or Bypass A. Containment pressure greater than (site-specific value) OR B. Explosive mixture exists inside containment OR C. 1. Containment pressure greater than (site-specific pressure setpoint) AND 2. Less than one full train of (site-specific system or equipment) is operating per design for 15 minutes or longer. CNB6 Containment pressure > 73.7 psia. 73.7 psia is the site specific containment design pressure. CNB7 Containment hydrogen concentration > 3%. 3% hydrogen concentration in the presence of oxygen represents an explosive mixture in containment. CNB8 Containment pressure > 44.7 psia [23.3 psia] with < one full train of containment heat removal systems (Note 9) operating per design for 15 min. (Note 1) The Containment pressure setpoint is the pressure at which the Containment Spray System should actuate and begin performing its function. Added Note 1 consistent with other thresholds with a timing component. Added Note 9 to clarify what constitutes a full train of containment heat removal systems. CNMT P-Loss 5 Other Indications A. (site-specific as applicable) N/A N/A No other site-specific Containment Potential Loss indication has been identified for ANO. CNMT P-Loss 6 Emergency Director Judgment A. ANY condition in the opinion of the Emergency Director that indicates Potential Loss of the Containment barrier. CNB10 Any condition in the opinion of the Emergency Director that indicates potential loss of the containment barrier. None to 0CAN031801 Page 76 of 120 Category H Hazards and Other Conditions Affecting Plant Safety NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification HU1 Confirmed SECURITY CONDITION or threat MODE: All HU1 Confirmed SECURITY CONDITION or threat. MODE: All None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the (site-specific security shift supervision). HU1.1 A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by ANO Security Shift Supervision OR Notification of a credible security threat directed at the site OR A validated notification from the NRC providing information of an aircraft threat. Example EALs #1, 2 and 3 have been combined into a single EAL for ease of presentation and use. 2 Notification of a credible security threat directed at the site. 3 A validated notification from the NRC providing information of an aircraft threat. to 0CAN031801 Page 77 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification HU2 Seismic event greater than OBE levels. MODE: All HU2 Seismic event greater than OBE levels. MODE: All None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Seismic event greater than Operating Basis Earthquake (OBE) as indicated by: (site-specific indication that a seismic event met or exceeded OBE limits) HU2.1 Seismic event > OBE as indicated by annunciation of the 0.10 g acceleration alarm The ANO OBE alarm is located on the seismic network control center, C529-NCC in the Unit 1 Control Room. to 0CAN031801 Page 78 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification HU3 Hazardous event. MODE: All HU3 Hazardous event. MODE: All None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 A tornado strike within the PROTECTED AREA. HU3.1 A tornado strike within the PROTECTED AREA. None 2 Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode. HU3.2 Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component required by Technical Specifications for the current operating mode. Replaced the word "needed" with "-required by Technical Specifications-" consistent with the generic bases. 3 Movement of personnel within the PROTECTED AREA is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release). HU3.3 Movement of personnel within the PROTECTED AREA is IMPEDED due to an event external to the PROTECTED AREA involving hazardous materials (e.g., an offsite chemical spill or toxic gas release). Changed the term "offsite" to "external to the PROTECTED AREA" to address events located external to the PROTECTED AREA but not considered offsite. 4 A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles. HU4.1 A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles. (Note 7) Added reference to Note 7. 5 (Site-specific list of natural or technological hazard events) N/A N/A No other site-specific hazard has been identified for ANO. Note EAL #3 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents. N/A Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents. This note, designated Note #7, is intended to apply to generic example EAL #4, not #3 as specified in the generic guidance. to 0CAN031801 Page 79 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification HU4 FIRE potentially degrading the level of safety of the plant. MODE: All HU4 FIRE potentially degrading the level of safety of the plant. MODE: All None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 a. A FIRE is NOT extinguished within 15-minutes of ANY of the following FIRE detection indications: Report from the field (i.e., visual observation) Receipt of multiple (more than 1) fire alarms or indications Field verification of a single fire alarm AND b. The FIRE is located within ANY of the following plant rooms or areas: (site-specific list of plant rooms or areas) HU4.1 A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (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 alarm AND The FIRE is located within any Table 1[2]H-1 area. Table 1[2]H-1 provides a tabularized list of site-specific fire areas. to 0CAN031801 Page 80 of 120 NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 2 a. Receipt of a single fire alarm (i.e., no other indications of a FIRE). AND b. The FIRE is located within ANY of the following plant rooms or areas: (site-specific list of plant rooms or areas) AND c. The existence of a FIRE is not verified within 30-minutes of alarm receipt. HU4.2 Receipt of a single fire alarm (i.e., no other indications of a FIRE) AND The fire alarm is indicating a FIRE within any Table 1[2]H-1 area AND The existence of a FIRE is not verified within 30 min. of alarm receipt. (Note 1) Table 1[2]H-1 provides a tabularized list of site-specific fire areas. In addition, the wording in the Basis section supporting fire detection and response design is modified to remove specific reference to 10 CFR 50, Appendix R. The Basis statement relating to Criterion 3 of 10 CFR 50, Appendix A is maintained, along with generic statements that are applicable without regard to Appendix R. Both ANO units are NFPA 805 plants; therefore, the requirements of Appendix R are no longer applicable. This wording is a difference from NEI 99-01, Revision 6, HU4 generic wording and bases. This difference is acceptable in order eliminate reference to an inappropriate licensing basis while maintaining sufficient information to address fire prevention and fire system functions 3 A FIRE within the plant or ISFSI [for plants with an ISFSI outside the plant Protected Area] PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication. HU4.3 A FIRE within the PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication. (Note 1) ANO has an ISFSI located inside the plant Protected Area. 4 A FIRE within the plant or ISFSI [for plants with an ISFSI outside the plant Protected Area] PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish. HU4.4 A FIRE within the PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish ANO has an ISFSI located inside the plant Protected Area. to 0CAN031801 Page 81 of 120 NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification Note Note: The Emergency Director should declare the Unusual Event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. N/A Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock. Table 1H-1 Unit 1 Fire Areas Reactor Building All elevations Auxiliary Building All elevations including: Penthouse/MSIV Room Exceptions: Boric Acid Mix Tank Room (Chem Add Area), 404' (157-B), EDG Exhaust Fan area on 386' (1-E and 2-E) Turbine Building All elevations including: Pipechase under ICW Coolers, CRD Pump Pit/T-28 Room/Area under ICW Pumps Outside Areas Manholes adjacent to Startup #2 XFMR (MH-03/MH-04) Manholes adjacent to Intake Structure (MH-05/MH-06) Intake Structure (354' and 366')

Diesel Fuel Vault Diesel Fuel Vault Pump Manholes (MH-09 and MH-10) to 0CAN031801 Page 82 of 120 Table 2H-1 Unit 2 Fire Areas Reactor Building All elevations Auxiliary Building All elevations including: Aux Extension Turbine Building All elevations Outside Areas Intake Structure (354' and 366') Concrete Manhole East, NE of intake (2MH-01)

Concrete Manhole East of Turbine Building next to train bay (2MH-03) Diesel Fuel Vault Diesel Fuel Vault Pump Manholes (MH-09 and MH-10) to 0CAN031801 Page 83 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification HU7 Other conditions exist which in the judgment of the Emergency Director warrant declaration of a (NO)UE. MODE: All HU7 Other conditions exist that in the judgment of the Emergency Director warrant declaration of a UE. MODE: All None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of 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 expected unless further degradation of safety systems occurs. HU7.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of 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 expected unless further degradation of SAFETY SYSTEMS occurs. None to 0CAN031801 Page 84 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification HA1 HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes. MODE: All HA1 HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes. MODE: All None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the (site-specific security shift supervision). HA1.1 A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by ANO Security Shift Supervision OR A validated notification from NRC of an aircraft attack threat within 30 min. of the site. Example EALs #1 and #2 have been combined into a single EAL for ease of use. 2 A validated notification from NRC of an aircraft attack threat within 30 minutes of the site. to 0CAN031801 Page 85 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification HA5 Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown. MODE: All HA5 Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown. MODE: 3 - Hot Standby, 4 - Hot Shutdown The mode applicability has been limited to the modes restrictions of Table H-2, Modes 3 and 4 only. NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 a. Release of a toxic, corrosive, asphyxiant or flammable gas into any of the following plant rooms or areas: (site-specific list of plant rooms or areas with entry-related mode applicability identified) AND b. Entry into the room or area is prohibited or impeded. HA5.1 Release of a toxic, corrosive, asphyxiant or flammable gas into any Table 1[2]H-2 room or area AND Entry into the room or area is prohibited or IMPEDED. (Note 5) The site-specific list of plant rooms or areas with entry-related mode applicability are tabularized in Table 1[2]H-2. The bulleted bases item "the action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections)" was removed from the list of exceptions to classification in the basis information. These actions are a consideration when the site-specific list was developed. Rooms requiring entry for these types of actions are already excluded from the list when it was developed. Note Note: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. N/A Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. None to 0CAN031801 Page 86 of 120 Table 1H-2 Unit 1 Safe Operation & Shutdown Rooms/Areas Room/Area Mode A-4 Switchgear Room 3, 4 Upper North Electrical Penetration Room 3, 4 Lower South Electrical Equipment Room 3, 4 Table 2H-2 Unit 2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Aux Building 317' Emergency Core Cooling Rooms 3, 4 Aux Building 317' Tendon Gallery Access 3, 4 Aux Building 335' Charging Pumps / MCC 2B-52 3, 4 Aux Building 354' MCC 2B-62 Area 3, 4 Emergency Diesel Generator Corridor 3, 4 Lower South Piping Penetration Room 3, 4 Aux Building 386' Containment Hatch 3, 4 to 0CAN031801 Page 87 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification HA6 Control Room evacuation resulting in transfer of plant control to alternate locations. MODE: All HA6 Control Room evacuation resulting in transfer of plant control to alternate locations. MODE: All None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 An event has resulted in plant control being transferred from the Control Room to (site-specific remote shutdown panels and local control stations). HA6.1 An event has resulted in plant control being transferred from the Control Room to alternate locations. Shutdown activities at ANO are transferred from the Control Room to multiple locations within the plant. to 0CAN031801 Page 88 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification HA7 Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert. MODE: All HA7 Other conditions exist that in the judgment of the Emergency Director warrant declaration of an Alert. MODE: All None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. HA7.1 Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. None to 0CAN031801 Page 89 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification HS1 HOSTILE ACTION within the PROTECTED AREA. MODE: All HS1 HOSTILE ACTION within the PROTECTED AREA. MODE: All None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the (site-specific security shift supervision). HS1.1 A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by ANO Security Shift Supervision. None to 0CAN031801 Page 90 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification HS6 Inability to control a key safety function from outside the Control Room. MODE: All HS6 Inability to control a key safety function from outside the Control Room. MODE: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling Deleted defueled mode applicability. Control of the cited safety functions is not critical for a defueled reactor as there is no energy source in the reactor vessel or RCS. This is an acceptable deviation from the generic NEI 99-01 Revision 6 guidance. NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 a. An event has resulted in plant control being transferred from the Control Room to (site-specific remote shutdown panels and local control stations). AND b. Control of ANY of the following key safety functions is not reestablished within (site-specific number of minutes). Reactivity control Core cooling [PWR] / RPV water level [BWR] RCS heat removal HS6.1 An event has resulted in plant control being transferred from the Control Room to alternate locations AND Control of any of the following key safety functions is not reestablished within 15 min. (Note 1): Reactivity (Modes 1, 2 and 3 only) Core cooling RCS heat removal Shutdown activities at ANO are transferred from the Control Room to multiple locations within the plant. The Mode applicability for the reactivity control safety function has been limited to Modes 1, 2, and 3 (hot operating conditions).

In the cold operating modes adequate shutdown margin exists under all conditions. EP FAQ 2015-014.

This is an acceptable deviation from the generic NEI 99-01 Revision 6 guidance.

to 0CAN031801 Page 91 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification HS7 Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency. MODE: All HS7 Other conditions exist that in the judgment of the Emergency Director warrant declaration of a Site Area Emergency. MODE: All None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary. HS7.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY. None to 0CAN031801 Page 92 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification HG1 HOSTILE ACTION resulting in loss of physical control of the facility. MODE: All N/A N/A IC HG1 and associated example EAL are not implemented in the ANO scheme. There are several other ICs that are redundant with this IC, and are better suited to ensure timely and effective emergency declarations. In addition, the development of new spent fuel pool level EALs, as a result of NRC Order EA 051, clarified the intended emergency classification level for spent fuel pool level events. This is an acceptable deviation from the generic NEI 99-01 Revision 6 guidance. to 0CAN031801 Page 93 of 120 NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 a. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the (site-specific security shift supervision). AND b. EITHER of the following has occurred: 1. ANY of the following safety functions cannot be controlled or maintained. Reactivity control Core cooling [PWR]/RPV water level [BWR] RCS heat removal OR 2. Damage to spent fuel has occurred or is IMMINENT. N/A N/A IC HG1 and associated example EAL are not implemented in the ANO scheme. There are several other ICs that are redundant with this IC, and are better suited to ensure timely and effective emergency declarations. In addition, the development of new spent fuel pool level EALs, as a result of NRC Order EA 051, clarified the intended emergency classification level for spent fuel pool level events. This deviation is justified because: 1. Hostile Action in the Protected Area is bounded by ICs HS1 and HS7. Hostile Action resulting in a loss of physical control is bound by EAL HG7, as well as any event that may lead to radiological releases to the public in excess of Environmental Protection Agency (EPA) Protective Action Guides (PAGs). a. If, for whatever reason, the Control Room must be evacuated, and control of safety functions (e.g., reactivity control, core cooling, and RCS heat removal) cannot be reestablished, then IC HS6 would apply, as well as IC HS7 if desired by the EAL decision-maker. b. Also, as stated above, any event (including Hostile Action) that could reasonably be expected to have a release exceeding EPA PAGs would be bound by IC HG7. c. From a Hostile Action perspective, ICs HS1, HS7 and HG7 are appropriate, and therefore, make this part of HG1 redundant and unnecessary. d. From a loss of physical control perspective, ICs HS6, HS7 and HG7 are appropriate, and therefore, make this part of HG1 redundant and unnecessary. 2. Any event which causes a loss of spent fuel pool level will be bounded by ICs AA2, AS2 and AG2, regardless of whether it was based upon a Hostile Action or not, thus making this part of HG1 redundant and unnecessary. a. An event that leads to a radiological release will be bounded by ICs AU1, AA1, AS1 and AG1. Events that lead to radiological releases in excess of EPA PAGs will be bounded by EALs AG1 and HG7, thus making this part of HG1 redundant and unnecessary. ICs AA2, AS2, AG2, AS1, AG1, HS1, HS6, HS7 and HG7 have been implemented consistent with NEI 99-01, Revision 6, and thus HG1 is adequately bounded as described above. EP FAQ 2015-013 This is an acceptable deviation from the generic NEI 99-01 Revision 6 guidance. to 0CAN031801 Page 94 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification HG7 Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency. MODE: All HG7 Other conditions exist that in the judgment of the Emergency Director warrant declaration of a General Emergency. MODE: All None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area. HG7.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area. None to 0CAN031801 Page 95 of 120 Category S System Malfunction NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification SU1 Loss of all offsite AC power capability to emergency buses for 15 minutes or longer. MODE: Power Operation, Startup, Hot Standby, Hot Shutdown SU1 Loss of all offsite AC power capability to vital buses for 15 minutes or longer. MODE: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown "vital buses" is the ANO-specific terminology for "emergency buses". NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Loss of ALL offsite AC power capability to (site-specific emergency buses) for 15 minutes or longer. SU1.1 Loss of all offsite AC power capability, Table 1[2]S-1, to vital 4.16 KV buses A3[2A3] and A4[2A4] for 15 min. (Note 1) 4.16 KV buses A3[2A3] and A4[2A4] are the site-specific emergency buses. Site-specific AC power sources are tabularized in Table 1[2]S-1. Note The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. N/A Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock.

to 0CAN031801 Page 96 of 120 Table 1S-1 Unit 1 AC Power Sources Offsite Startup Transformer No. 1 Startup Transformer No. 2 Unit Auxiliary Transformer (from 22 KV switchyard) Onsite Unit Auxiliary Transformer (main generator via main transformer) DG1 DG2 AAC Gen Table 2S-1 Unit 2 AC Power Sources Offsite Startup Transformer No. 3 Startup Transformer No. 2 Unit Auxiliary Transformer (backfed from main transformer) Onsite Unit Auxiliary Transformer (main generator via main transformer) 2DG1 2DG2 AAC Gen to 0CAN031801 Page 97 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification SU2 UNPLANNED loss of Control Room indications for 15 minutes or longer. MODE: Power Operation, Startup, Hot Standby, Hot Shutdown SU3 UNPLANNED loss of Control Room indications for 15 minutes or longer. MODE: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer. SU3.1 An UNPLANNED event results in the inability to monitor one or more Table 1[2]S-2 parameters from within the Control Room for 15 min. (Note 1) The site-specific Safety System Parameter list is tabulated in Table 1[2]S-2. Added the words "to at least one S/G" to Auxiliary or emergency feedwater flow. This is consistent with Level in at least on S/G. Note The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. N/A Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock.

to 0CAN031801 Page 98 of 120 [BWR parameter list] [PWR parameter list] Reactor Power Reactor Power RPV Water Level RCS Level RPV Pressure RCS Pressure Primary Containment Pressure In-Core/Core Exit Temperature Suppression Pool Level Levels in at least (site-specific number) steam generators Suppression Pool Temperature Steam Generator Auxiliary or Emergency Feed Water Flow Table 1[2]S-2 Safety System Parameters Reactor power RCS level RCS pressure CET temperature Level in at least one S/G EFW flow to at least one S/G to 0CAN031801 Page 99 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification SU3 Reactor coolant activity greater than Technical Specification allowable limits MODE: Power Operation, Startup, Hot Standby, Hot Shutdown SU4 RCS activity greater than Technical Specification allowable limits. MODE: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Changed "Reactor coolant" to read "RCS" for consistency with normally used terminology. NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 (Site-specific radiation monitor) reading greater than (site-specific value). SU4.1 Failed Fuel Iodine radiation monitor RI-1237S[2RITS-4806B] reading > 9.0 E5 cpm Unit 1 RE-1237S, Failed Fuel Monitor, is in the letdown system to monitor the letdown line for evidence of fuel damage. Unit 2 specific activity monitor 2RITS-4806B monitors the letdown fluid for the presence of Iodine-131. 2 Sample analysis indicates that a reactor coolant activity value is greater than an allowable limit specified in Technical Specifications. SU4.2 RCS sample activity > 1.0 uCi/gm dose equivalent I-131 for > 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> (Note 1)OR RCS sample activity > 60 uCi/gm dose equivalent I-131 OR RCS sample activity > 2200[3100] uCi/gm dose equivalent Xe-133 for > 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> (Note 1) Changed "Reactor coolant" to read "RCS" for consistency with normally used terminology. ANO Unit 1 T.S. LCO 3.4.12 and Unit 2 T.S. LCO 3.4.8 provides the TS allowable coolant activity limits that are duplicated in the EAL. to 0CAN031801 Page 100 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification SU4 RCS leakage for 15 minutes or longer. MODE: Power Operation, Startup, Hot Standby, Hot Shutdown SU5 RCS leakage for 15 minutes or longer. MODE: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 RCS unidentified or pressure boundary leakage greater than (site-specific value) for 15 minutes or longer. SU5.1 RCS unidentified or pressure boundary leakage > 10 gpm for 15 min. (Note 1)OR RCS identified leakage > 25 gpm for 15 min. (Note 1) OR Reactor coolant leakage to a location outside containment > 25 gpm for 15 min. (Note 1) Example EALs #1, 2 and 3 have been combined into a single EAL for usability. Changed the term 'RCS' to 'Reactor' coolant for the third threshold. For ANO, leakage outside containment is not termed RCS leakage. Added "Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification" to the basis. This provides agreement with the definition of "Unisolable" and ensures isolation attempts, both locally and remotely, are achieved in a timely manner. 2 RCS identified leakage greater than (site-specific value) for 15 minutes or longer. 3 Leakage from the RCS to a location outside containment greater than 25 gpm for 15 minutes or longer. Note The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. N/A Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock. to 0CAN031801 Page 101 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification SU5 Automatic or manual (trip [PWR] / scram [BWR]) fails to shutdown the reactor. MODE: Power Operation SU6 Automatic or manual trip fails to shut down the reactor. MODE: 1 - Power Operation None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 a. An automatic (trip [PWR] / scram [BWR]) did not shutdown the reactor. AND b. A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor. SU6.1 An automatic trip did not shut down the reactor as indicated by reactor power > 5% after any RPS setpoint is exceeded AND A subsequent automatic trip or manual trip action taken at the reactor control console (C03[2C03/2C14]) (manual reactor trip pushbuttons or DROPS[DSS]) is successful in shutting down the reactor as indicated by reactor power 5% (Note 8) As specified in the generic developers guidance "Developers may include site-specific EOP criteria indicative of a successful reactor shutdown in an EAL statement, the Basis or both (e.g., a reactor power level)." Reactor power < 5% is the site-specific indication of a successful reactor trip. Added the words "... as indicated by reactor power > 5% after any RPS setpoint is exceeded" to clarify that it is a failure of the automatic trip when a valid trip signal has been exceeded. Panels C03[2C03/2C14] are the site-specific reactor control consoles with manual trip capability. to 0CAN031801 Page 102 of 120 NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 2 a. A manual trip ([PWR] / scram [BWR]) did not shutdown the reactor. AND b. EITHER of the following: 1. A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor. OR 2 A subsequent automatic (trip [PWR] / scram [BWR]) is successful in shutting down the reactor. SU6.2 A manual trip did not shut down the reactor as indicated by reactor power > 5% after any manual trip action was initiated AND A subsequent automatic trip or manual trip action taken at the reactor control console (C03[2C03/2C14]) (manual reactor trip pushbuttons or DROPS[DSS]) is successful in shutting down the reactor as indicated by reactor power 5% (Note 8) As specified in the generic developers guidance "Developers may include site-specific EOP criteria indicative of a successful reactor shutdown in an EAL statement, the Basis or both (e.g., a reactor power level)." Reactor power < 5% is the site-specific indication of a successful reactor trip. Added the words "... as indicated by reactor power > 5% after any manual trip action was initiated" to clarify that it is a failure of any manual trip when an actual manual trip signal has been inserted. Combined conditions b.1 and b.2 into a single statement to simplify the presentation. Panels C03[2C03/2C14] are the site-specific reactor control consoles with manual trip capability. Notes Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. N/A Note 8: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. None to 0CAN031801 Page 103 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification SU6 Loss of all onsite or offsite communications capabilities. MODE: Power Operation, Startup, Hot Standby, Hot Shutdown SU7 Loss of all onsite or offsite communications capabilities. MODE: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Loss of ALL of the following onsite communication methods: (site-specific list of communications methods) SU7.1 Loss of all Table 1[2]S-4 onsite communication methods OR Loss of all Table 1[2]S-4 State and local agency communication methods OR Loss of all Table 1[2]S-4 NRC communication methods. Example EALs #1, 2 and 3 have been combined into a single EAL for simplification of presentation. Replaced "ORO" with "State and local agency" for clarification. Table 1[2]S-4 provides a site-specific list of onsite, State and local agency (ORO) and NRC communications methods. 2 Loss of ALL of the following ORO communications methods: (site-specific list of communications methods) 3 Loss of ALL of the following NRC communications methods: (site-specific list of communications methods) to 0CAN031801 Page 104 of 120 Table 1[2]S-4 Communication Methods System Onsite State/Local NRC Station radio system X ANO plant phone system X Gaitronics X Telephone Systems: Commercial Microwave Satellite VOIP X X INFORM Notification System X Emergency Notification System (ENS) X to 0CAN031801 Page 105 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification SU7 Failure to isolate containment or loss of containment pressure control. [PWR] MODE: Power Operation, Startup, Hot Standby, Hot Shutdown SU8 Failure to isolate containment or loss of containment pressure control. MODE: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 a. Failure of containment to isolate when required by an actuation signal. AND b. ALL required penetrations are not closed within 15 minutes of the actuation signal. SU8.1 Any penetration is not closed within 15 min. of an ESAS[CIAS] actuation signal OR Containment pressure > 44.7 psia [23.3 psia] with < one full train of containment heat removal systems (Note 9) operating per design for 15 min. (Note 1) Reworded EAL to better describe the intent. Penetrations cannot close, but they can be isolated by closure of one or more isolation valves associated with that penetration. The revised wording maintains the generic example EAL intent while more clearly describing failure to isolate threshold. ESAS[CIAS] is the site-specific system that initiates containment isolation signals. The Containment pressure setpoint is the pressure at which the Containment Spray System should actuate and begin performing its function. Added statement to the NEI basis that penetrations are considered closed when one of the two associated valves are closed or a check valve is intact. This change provides clarification and meets the intent of the EAL of ensuring that a functional boundary is maintained between the containment and the outside environment. 2 a. Containment pressure greater than (site-specific pressure). AND b. Less than one full train of (site-specific system or equipment) is operating per design for 15 minutes or longer. to 0CAN031801 Page 106 of 120 NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification N/A N/A N/A Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock. N/A N/A N/A Note 9: One full train of containment heat removal systems consists of one train of RB[Containment] Spray and one train of RB[Containment] Cooling System. Added new Note 9 to clarify what constitutes one full train of containment heat removal systems. to 0CAN031801 Page 107 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification SA1 Loss of all but one AC power source to emergency buses for 15 minutes or longer. MODE: Power Operation, Startup SA1 Loss of all but one AC power source to vital buses for 15 minutes or longer. MODE: 1 - Power Operation "vital buses" is the ANO-specific terminology for "emergency buses". NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 a. AC power capability to (site-specific emergency buses) is reduced to a single power source for 15 minutes or longer. AND b. Any additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS. SA1.1 AC power capability, Table 1[2]S-1, to vital 4.16 KV buses A3[2A3] and A4[2A4] reduced to a single power source for 15 min. (Note 1) AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS. 4.16KV buses A3[2A3] and A4[2A4] are the site-specific emergency buses. Site-specific AC power sources are listed in Table 1[2]S-1. Note The Emergency Director should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. N/A Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock. to 0CAN031801 Page 108 of 120 Table 1S-1 Unit 1 AC Power Sources Offsite Startup Transformer No. 1 Startup Transformer No. 2 Unit Auxiliary Transformer (from 22 KV switchyard) Onsite Unit Auxiliary Transformer (main generator via main transformer) DG1 DG2 AAC Gen Table 2S-1 Unit 2 AC Power Sources Offsite Startup Transformer No. 3 Startup Transformer No. 2 Unit Auxiliary Transformer (backfed from main transformer) Onsite Unit Auxiliary Transformer (main generator via main transformer) 2DG1 2DG2 AAC Gen to 0CAN031801 Page 109 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification SA2 UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress. MODE: Power Operation, Startup, Hot Standby, Hot Shutdown SA3 UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress. MODE: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer. AND ANY of the following transient events in progress. Automatic or manual runback greater than 25% thermal reactor power Electrical load rejection greater than 25% full electrical load Reactor scram [BWR] / trip [PWR] ECCS (SI) actuation Thermal power oscillations greater than 10% [BWR] SA3.1 An UNPLANNED event results in the inability to monitor one or more Table 1[2]S-2 parameters from within the Control Room for 15 min. (Note 1) AND Any significant transient is in progress, Table 1[2]S-3 The site-specific Safety System Parameter list to listed in Table 1[2]S-2. The site-specific significant transients list to listed in Table 1[2]S-3. ANO is a PWR and thus does not include thermal power oscillations > 10%. to 0CAN031801 Page 110 of 120 NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification Note The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. N/A Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock. to 0CAN031801 Page 111 of 120 [BWR parameter list] [PWR parameter list] Reactor Power Reactor Power RPV Water Level RCS Level RPV Pressure RCS Pressure Primary Containment Pressure In-Core/Core Exit Temperature Suppression Pool Level Levels in at least (site-specific number) steam generators Suppression Pool Temperature Steam Generator Auxiliary or Emergency Feed Water Flow Table 1[2]S-2 Safety System Parameters Reactor power RCS level RCS pressure CET temperature Level in at least one S/G EFW flow to at least one S/G Table 1[2]S-3 Significant Transients Reactor trip Runback > 25% thermal power Electrical load rejection > 25% electrical load Safety injection actuation to 0CAN031801 Page 112 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification SA5 Automatic or manual (trip [PWR] / scram [BWR]) fails to shutdown the reactor, and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor. MODE: Power Operation SA6 Automatic or manual trip fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor. MODE: 1 - Power Operation None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 a. An automatic or manual (trip [PWR] / scram [BWR]) did not shutdown the reactor. AND b. Manual actions taken at the reactor control consoles are not successful in shutting down the reactor. SA6.1 An automatic or manual trip fails to shut down the reactor as indicated by reactor power > 5% AND Manual trip actions taken at the reactor control console (C03[2C03/2C14]) (manual reactor trip pushbuttons or DROPS[DSS]) are not successful in shutting down the reactor as indicated by reactor power > 5%. (Note 8) As specified in the generic developers guidance "Developers may include site-specific EOP criteria indicative of a successful reactor shutdown in an EAL statement, the Basis or both (e.g., a reactor power level)." Reactor power < 5% is the site-specific indication of a successful reactor trip. Panels C03[2C03/2C14] are the site-specific reactor control consoles with manual trip capability. Notes Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. N/A Note 8: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. None to 0CAN031801 Page 113 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification SA9 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. MODE: Power Operation, Startup, Hot Standby, Hot Shutdown SA9.1 Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode. MODE: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Pluralized safety systems to be consistent with NRC EP FAQ 2016-002 that specifies degraded performance or visible damage in more than one safety system train. to 0CAN031801 Page 114 of 120 NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 a. The occurrence of ANY of the following hazardous events: Seismic event (earthquake) Internal or external flooding event High winds or tornado strike FIRE EXPLOSION (site-specific hazards) Other events with similar hazard characteristics as determined by the Shift Manager AND b. EITHER of the following: 1. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode. OR 2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode. SA9.1 The occurrence of any Table 1[2]S-5 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER: Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode. (Notes 10, 11) The hazardous events have been tabularized in Table 1[2]S-5. SA9.1 reflects NRC FAQ 2016-002 requiring degraded performance or visible damage to more than one train of a safety system caused by the specified events. This wording is a deviation from NEI 99-01 Revision 6 SA9 generic wording and bases but is deemed acceptable in order to ensure that an Alert is declared only when a hazardous event causes actual or potential performance issues with safety systems. This is consistent with NRC-approved EP FAQ 2016-002. The word "a" is replaced with "the" in the FAQ wording to provide agreement with the FAQ basis information indicating that the criteria is applicable to another train of the same safety system. to 0CAN031801 Page 115 of 120 NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification N/A N/A N/A Note 10: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted Added Note 10 consistent with the recommendation of NRC EP FAQ 2016-002. N/A N/A N/A Note 11: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted. Added Note 11 consistent with the recommendation of NRC EP FAQ 2016-002. Table 1[2]S-5 Hazardous Events Seismic event (earthquake) Internal or external FLOODING event High winds or tornado strike FIRE EXPLOSION Other events with similar hazard characteristics as determined by the Shift Manager to 0CAN031801 Page 116 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification SS1 Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer. MODE: Power Operation, Startup, Hot Standby, Hot Shutdown SS1 Loss of all offsite and all onsite AC power to vital buses for 15 minutes or longer. MODE: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown "vital buses" is the ANO-specific terminology for "emergency buses". NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Loss of ALL offsite and ALL onsite AC power to (site-specific emergency buses) for 15 minutes or longer. SS1.1 Loss of all offsite and all onsite AC power to vital 4.16 KV buses A3[2A3] and A4[2A4] for 15 min. (Note 1) 4.16KV buses A3[2A3] and A4[2A4] are the site-specific emergency buses. Note The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. N/A Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock. to 0CAN031801 Page 117 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification SS5 Inability to shutdown the reactor causing a challenge to (core cooling [PWR] / RPV water level [BWR]) or RCS heat removal. MODE: Power Operation SS6 Inability to shut down the reactor causing a challenge to core cooling or RCS heat removal. MODE: 1 - Power Operation None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 a. An automatic or manual (trip [PWR] / scram [BWR]) did not shutdown the reactor. AND b. All manual actions to shutdown the reactor have been unsuccessful. AND c. EITHER of the following conditions exist: (Site-specific indication of an inability to adequately remove heat from the core) (Site-specific indication of an inability to adequately remove heat from the RCS) SS6.1 An automatic or manual trip fails to shut down the reactor as indicated by reactor power > 5% AND All actions to shut down the reactor are not successful as indicated by reactor power > 5% AND EITHER: CETs > 1200°F RCS heat removal cannot be established using steam generators and HPI[Once Through] cooling initiated As specified in the generic developers guidance "Developers may include site-specific EOP criteria indicative of a successful reactor shutdown in an EAL statement, the Basis or both (e.g., a reactor power level)." Reactor power < 5% is the site-specific indication of a successful reactor trip. Deleted the term "manual actions" from the second condition. For generic IC SS5, all actions to shut down the reactor can be credited, not just those actions from the reactor control panel that may be identified as "manual actions." Indication that heat removal is extremely challenged is manifested by the loss of both steam generators as a heat sink, requiring the initiation of HPI [Once Through] cooling. to 0CAN031801 Page 118 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification SS8 Loss of all Vital DC power for 15 minutes or longer. MODE: Power Operation, Startup, Hot Standby, Hot Shutdown SS2 Loss of all vital DC power for 15 minutes or longer. MODE: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown None NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 Indicated voltage is less than (site-specific bus voltage value) on ALL (site-specific Vital DC busses) for 15 minutes or longer. SS2.1 Indicated voltage is < 105 VDC on D01 [2D01] and D02[2D02] vital 125 VDC buses for 15 min. (Note 1) 105 VDC is the site-specific minimum vital DC bus voltage. DC buses on D01[2D01] and D02[2D02] are the site-specific credited vital DC buses. Note The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. N/A Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock. to 0CAN031801 Page 119 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification SG1 Prolonged loss of all offsite and all onsite AC power to emergency buses. MODE: Power Operation, Startup, Hot Standby, Hot Shutdown SG1a Prolonged loss of all offsite and all onsite AC power to vital buses. MODE: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown "vital buses" is the ANO-specific terminology for "emergency buses". NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 a. Loss of ALL offsite and ALL onsite AC power to (site-specific emergency buses). AND b. EITHER of the following: Restoration of at least one AC emergency bus in less than (site-specific hours) is not likely. (Site-specific indication of an inability to adequately remove heat from the core) SG1.1 Loss of all offsite and all onsite AC power to vital 4.16 KV buses A3[2A3] and A4[2A4] AND EITHER: Restoration of at least one vital 4.16 KV bus in < 4 hrs is not likely (Note 1) CETs > 1200°F 4.16KV buses A3[2A3] and A4[2A4] are the site-specific emergency buses. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is the site-specific SBO coping analysis time. CETs reading > 1200°F indicates significant core exit superheating and core uncovery. Note The Emergency Director should declare the General Emergency promptly upon determining that (site-specific hours) has been exceeded, or will likely be exceeded. N/A Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock. to 0CAN031801 Page 120 of 120 NEI IC# NEI IC Wording ANO IC#(s) ANO IC Wording Difference/Deviation Justification SG8 Loss of all AC and Vital DC power sources for 15 minutes or longer. MODE: Power Operation, Startup, Hot Standby, Hot Shutdown SG1b Loss of all vital AC and vital DC power sources for 15 minutes or longer. MODE: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown "Vital buses" is the ANO-specific terminology for "emergency buses". NEI Ex. EAL # NEI Example EAL Wording ANO EAL # ANO EAL Wording Difference/Deviation Justification 1 a. Loss of ALL offsite and ALL onsite AC power to (site-specific emergency buses) for 15 minutes or longer. AND b. Indicated voltage is less than (site-specific bus voltage value) on ALL (site-specific Vital DC busses) for 15 minutes or longer. SG1.2 Loss of all offsite and all onsite AC power to vital 4.16 KV buses A3[2A3] and A4[2A4] for 15 min. AND Indicated voltage is < 105 VDC on D01 [2D01] and D02[2D02] vital 125 VDC buses for 15 min. (Note 1) 4.16KV buses A3[2A3] and A4[2A4] are the site-specific emergency buses. 105 VDC is the site-specific minimum vital DC bus voltage. D01[2D01] and D02[2D02] are the credited site-specific vital DC buses. Note The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. N/A Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. The classification timeliness note has been standardized across the ANO EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded." To reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock.

Enclosure 5 to 0CAN031801 Proposed EAL Matrix Chart and Review Table (for information only) to 0CAN031801 Page 1 of 22 UNIT 1 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT A Abnorm. Rad Levels / Rad Effluent 1 Rad Effluent Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE Release of gaseous OR liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer AG1.1 Reading on any Table 1A-1 effluent radiation monitor > column "GE" for 15 min. (Notes 1, 2, 3, 4) AG1.2 Dose assessment using actual meteorology indicates doses > 1000 mrem TEDE or 5000 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4) AG1.3 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: - Closed window dose rates > 1000 mR/hr expected to continue for 60 min. - Analyses of field survey samples indicate thyroid CDE > 5000 mrem for 60 min. of inhalation (Notes 1, 2) AS1.1 Reading on any Table 1A-1 effluent radiation monitor > column "SAE" for 15 min. (Notes 1, 2, 3, 4) AS1.2 Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4) AS1.3 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: - Closed window dose rates > 100 mR/hr expected to continue for 60 min. - Analyses of field survey samples indicate thyroid CDE > 500 mrem for 60 min. of inhalation (Notes 1, 2) AA1.1 Reading on any Table 1A-1 effluent radiation monitor > column "ALERT" for 15 min. (Notes 1, 2, 3, 4) AA1.2 Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4) AA1.3 Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses

> 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2) AA1.4 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: - Closed window dose rates > 10 mR/hr expected to continue for 60 min. - Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation (Notes 1, 2) AU1.1 Reading on any Table 1A-1 effluent radiation monitor > column "UE" for 60 min. (Notes 1, 2, 3) AU1.2 Sample analysis for a gaseous or liquid release indicates a concentration or release rate > 2 x ODCM limits for 60 min. (Notes 1, 2) 2 Irradiated Fuel Event Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer Spent fuel pool level at the top of the fuel racks Significant lowering of water level above, or damage to, irradiated fuel UNPLANNED loss of water level above irradiated fuel AG2.1 Spent fuel pool level cannot be restored to at least 377.0 ft. (Alarm 3) on LIT-2020-3(4) for 60 min. (Note 1) AS2.1 Lowering of spent fuel pool level to 377.0 ft. (Alarm 3) on LIT-2020-3(4) AA2.1 IMMINENT uncovery of irradiated fuel in the REFUELING PATHWAY AA2.2 Damage to irradiated fuel resulting in a release of radioactivity AND High alarm on any Table 1A-2 radiation monitor AA2.3 Lowering of spent fuel pool level to 387.0 ft. (Alarm 2) on LIT-2020-3(4) AU2.1 UNPLANNED water level drop in the REFUELING PATHWAY as indicated by low water level alarm, visual observation, or BWST level drop due to makeup demands AND UNPLANNED rise in corresponding area radiation levels as indicated by any of the following radiation monitors: - RE-8009 Spent Fuel Area - RE-8017 Fuel Handling Table 1A-1 Unit 1 Effluent Monitor Classification Thresholds Gaseous Monitor GE SAE Alert UE Containment Purge RX-9820 (SPING 1) 4.15E+01 µCi/cc 4.15E+00 µCi/cc 4.15E-01 µCi/cc 1.21E-03 µCi/cc Radwaste Area RX-9825 (SPING 2) 2.67E+01 µCi/cc 2.67E+00 µCi/cc 2.67E-01 µCi/cc 4.94E-04 µCi/cc Fuel Handling Area RX-9820 (SPING 3) 6.20E+02 µCi/cc 6.20E+01 µCi/cc 6.20E+00 µCi/cc 5.44E-04 µCi/cc Emergency Penetration Room RX-9835 (SPING 4) 6.55E+02 µCi/cc 6.55E+01 µCi/cc 6.55E+00 µCi/cc 1.21E-02 µCi/cc Liquid Liquid Radwaste RE-4642 ---- ---- ---- 2.46E+05 cpm 3 Area Rad Levels Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown Table 1A-2 Unit 1 Fuel Damage Radiation Monitors RE-8009 Spent Fuel Area RE-8017 Fuel Handling RE-8060 Containment High Range Radiation Monitor RE-8061 Containment High Range Radiation Monitor RE-9820 (SPING 1) Containment Purge RE-9825 (SPING 2) Radwaste Area RE-9830 (SPING 3) Fuel Handling Area AA3.1 Dose rate > 15 mR/hr in EITHER of the following areas: - Control Room

- Central Alarm Station (by survey) AA3.2 An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table 1A-3 room or area (Note 5) None Table 1A-3 Unit 1 Safe Operation & Shutdown Rooms/Areas Room / Area Mode A-4 Switchgear Room Upper North Electrical Penetration Room Lower South Electrical Equipment Room 3, 4 3, 4 3, 4 1 2 3 4 5 6 DEF 1 2 3 4 5 6 DEF 1 2 3 4 5 6 DEF 1 2 3 4 5 6 DEF 1 2 3 4 5 6 DEF 1 2 3 4 5 6 DEF 1 2 3 4 5 6 DEF 1 2 3 4 5 6 DEF 1 2 3 4 5 6 DEF 3 4 to 0CAN031801 Page 2 of 22 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT C Cold SD/Refuel System Malfunct. 1 RCS Level Loss of RCS inventory affecting fuel clad integrity with containment challenged Loss of RCS inventory affecting core decay heat removal capability Significant loss of RCS inventory UNPLANNED loss of RCS inventory CG1.1 [Unit 2 ONLY] CG1.2 RCS level cannot be monitored for 30 min. (Note 1) AND Core uncovery is indicated by any of the following: - UNPLANNED rise in any Table 1C-1 sump/tank level of sufficient magnitude to indicate core uncovery - Containment High Range Radiation Monitor RE-8060/8061 reading > 10 R/hr - Erratic Source Range Monitor indication AND Any Containment Challenge indication, Table 1C-2 CS1.1 CONTAINMENT CLOSURE not established AND RVLMS Levels 1 through 9 indicate DRY CS1.2 RCS level cannot be monitored for 30 min. (Note 1) AND Core uncovery is indicated by any of the following: - UNPLANNED rise in any Table 1C-1 sump/tank level of sufficient magnitude to indicate core uncovery - Containment high range radiation monitor RE-8060/8061 reading > 10 R/hr - Erratic Source Range Monitor indication CA1.1 Loss of RCS inventory as indicated by EITHER: - RVLMS Levels 1 through 8 indicate DRY - Reactor vessel level 368.5 ft. (LT-1195/LT-1196) (bottom of hot leg) CA1.2 RCS level cannot be monitored for 15 min. (Note 1) AND EITHER - UNPLANNED rise in any Table 1C-1 Sump/Tank level due to a loss of RCS inventory - Visual observation of UNISOLABLE RCS leakageCU1.1 UNPLANNED loss of reactor coolant results in RCS water level less than a required lower limit for 15 min. (Note 1) CU1.2 RCS level cannot be monitored AND EITHER - UNPLANNED rise in any Table 1C-1 sump/tank level due to loss of RCS inventory - Visual observation of UNISOLABLE RCS leakage 2 Loss of Vital AC Power None None Loss of all offsite and all onsite AC power to vital buses for 15 minutes or longer Loss of all but one AC power source to vital buses for 15 minutes or longer CA2.1 Loss of all offsite and all onsite AC power to vital 4.16 KV buses A3 and A4 for 15 min. (Note 1) CU2.1 AC power capability, Table 1C-3, to vital 4.16 KV buses A3 and A4 reduced to a single power source for 15 min. (Note 1) AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS 3 RCS Temp. None None Inability to maintain plant in cold shutdown UNPLANNED rise in RCS temperature CA3.1 UNPLANNED rise in RCS temperature to > 200°F for > Table 1C-4 duration (Note 1) OR UNPLANNED RCS pressure rise > 10 psig due to a loss of RCS cooling (this EAL does not apply during water-solid plant conditions) CU3.1 UNPLANNED rise in RCS temperature to > 200°F due to loss of decay heat removal capability CU3.2 Loss of all RCS temperature and RCS level indication for 15 min. (Note 1) 4 Loss of Vital DC Power None None None Loss of Vital DC power for 15 minutes or longer CU4.1 Indicated voltage is < 105 VDC on vital 125 VDC buses for 15 min. (Note 1) 5 Loss of Comm. None None None Loss of all onsite or offsite communications capabilities CU5.1 Loss of all Table 1C-5 onsite communication methods OR Loss of all Table 1C-5 State and local agency communication methods OR Loss of all Table 1C-5 NRC communication methods 5 6 5 65 65 65 6 DEF5 6 DEF5 65 65 65 6 DEF to 0CAN031801 Page 3 of 22 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT C Cold SD/Refuel System Malfunct. 6 HazardousEvent Affecting Safety Systems None None Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode None CA6.1 The occurrence of any Table 1C-6 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER - Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode - Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 10, 11)

5 6*Reactor Building Sump *Reactor Drain Tank

• Aux. Building Equipment Drain Tank

• Aux. Building Sump

• Quench Tank Table 1C-1 Unit 1 Sumps / Tanks Offsite *Startup Transformer No. 1 *Startup Transformer No. 2

• Unit Auxiliary Transformer (from 22 KV switchyard) Onsite *DG1 *DG2 *AAC Gen Table 1C-3 Unit 1 AC Power Sources *CONTAINMENT CLOSURE not established (Note 6) *Containment hydrogen concentration > 3% *UNPLANNED rise in containment pressure Table 1C-2 Containment Challenge Indications *Seismic event (earthquake)

• Internal or external FLOODING event

• High winds or tornado strike

• FIRE

• EXPLOSION

• Other events with similar hazard characteristics as determined by the Shift Manager Table 1C-6 Hazardous Events Station radio system ANO plant phone system Gaitronics

Telephone Systems: *Commercial

• Microwave

• Satellite

• VOIP INFORM Notification System

Emergency Notification System (ENS) System State / Local Onsite NRC Table 1C-5 Communication Methods X X X

X X

X

X Intact (but not lowered inventory) Not intact OR Lowered inventory RCS Status CONTAINMENT CLOSURE Status Heat-up Duration N/A Established not established Table 1C-4 RCS Heat-up Duration Thresholds 60 min.* 20 min.*

0 min.

• If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable to 0CAN031801 Page 4 of 22 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT E ISFSI Damage to a loaded cask CONFINEMENT BOUNDARY None None Table 1E-1 ISFSI Dose Rates VSC-24 VCC HI-STORM - 200 mrem/hr on the sides - 400 mrem/hr on the top - 700 mrem/hr at the air inlet - 200 mrem/hr at the air outlet - 60 mrem/hr (gamma + neutron) on the top or outlet vent - 600 mrem/hr (gamma +

neutron on the side of the side of the overpack (excluding inlet and outlet ducts) EU1.1 Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask (VSC-24 VCC or HI-STORM overpack) > any Table 1E-1 value

GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT F Fission Product Barriers None FG1.1 Loss of any two barriers AND Loss or potential loss of the third barrier (Table 1F-1) FS1.1 Loss or potential loss of any two barriers (Table 1F-1) FS1.1 Any loss or any potential loss of either Fuel Clad or RCS barrier (Table 1F-1) 1 2 3 4 5 6 DEF 1 2 3 41 2 3 41 2 3 4 to 0CAN031801 Page 5 of 22 Table 1F-1 Fission Product Barrier Threshold Matrix Category Fuel Clad Barrier (FCB) Reactor Coolant System Barrier (RCB) Containment Barrier (CNB) Loss Potential Loss Loss Potential Loss Loss Potential Loss A. RCS or S/G Tube Leakage None FCB1 RVLMS Levels 1 through 9 indicate DRY RCB1 An automatic or manual ESAS actuation required by EITHER: - UNISOLABLE RCS leakage - S/G tube RUPTURE RCB2 UNISOLABLE RCS leakage or S/G tube leakage > 50 gpm excluding normal reductions in RCS inventory (e.g., letdown, RCP seal leakoff) RCB3 PTS limits apply (RT14) AND RCS pressure and temperature are left of the NDTT/LTOP limit lines on EOP Figure 3 (Note 12) CNB1 A S/G that is leaking > 50 gpm (excluding normal reductions in RCS inventory) or that is RUPTURED is also FAULTED outside of containment None B Inadequate Heat Removal FCB2 CETs > 1200°F FCB3 CETs > 700°F FCB4 RCS heat removal cannot be established using steam generators AND HPI cooling initiated None RCB4 RCS heat removal cannot be established using steam generators AND HPI cooling initiated None CNB2 CETs > 1200°F AND Restoration procedures not effective within 15 min. (Note 1) C CTMT Radiation / RCS Activity FCB5 Containment High Range Radiation Monitor RE-8060/8061 > 750 R/hr FCB6 Coolant activity > 300 Ci/gm dose equivalent I-131 None RCB5 Containment High Range Radiation Monitor RE-8060/8061 > 40 R/hr None None CNB3 Containment High Range Radiation Monitor RE-8060/8061 > 10,000 R/hr D CTMT Integrity or Bypass None None None None CNB4 Containment isolation is required AND EITHER: - Containment integrity has been lost based on Emergency Director judgment - UNISOLABLE pathway from Containment to the environment exists CNB5 Indications of RCS leakage outside of Containment CNB6 Containment pressure > 73.7 psia CNB7 Containment hydrogen concentration > 3% CNB8 Containment pressure > 44.7 psia with < one full train of containment heat removal systems (Note 9) operating per design for 15 min. (Note 1) E Emergency Director Judgment FCB7 Any condition in the opinion of the Emergency Director that indicates loss of the Fuel Clad barrier FCB8 Any condition in the opinion of the Emergency Director that indicates potential loss of the Fuel Clad barrier RCB6 Any condition in the opinion of the Emergency Director that indicates loss of the RCS barrier RCB7 Any condition in the opinion of the Emergency Director that indicates potential loss of the RCS barrier CNB9 Any condition in the opinion of the Emergency Director that indicates loss of the Containment barrier CNB10 Any condition in the opinion of the Emergency Director that indicates potential loss of the Containment barrier to 0CAN031801 Page 6 of 22 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT H Hazards 1 Security None HOSTILE ACTION within the PROTECTED AREA HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes Confirmed SECURITY CONDITION or threat HS1.1 A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by ANO Security Shift Supervision HA1.1 A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by ANO Security Shift Supervision OR A validated notification from NRC of an aircraft attack threat within 30 min. of the site HU1.1 A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by ANO Security Shift Supervision OR Notification of a credible security threat directed at the site OR A validated notification from the NRC providing information of an aircraft threat 2 Seismic Event None None [Refer to CA6.1 or SA9.1 for potential escalation due to a seismic event] None Seismic event greater than OBE levels HU2.1 Seismic event > OBE as indicated by annunciation of the 0.10 g acceleration alarm 3 Natural or Technical Hazard None None [Refer to CA6.1 or SA9.1 for potential escalation due to a hazardous event] None Natural or Technological Hazard HU3.1 A tornado strike within the PROTECTED AREA HU3.2 Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component required by Technical Specifications for the current operating mode HU3.3 Movement of personnel within the PROTECTED AREA is IMPEDED due to an event external to the PROTECTED AREA involving hazardous materials (e.g., an offsite chemical spill or toxic gas release) HU3.4 A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7) 1 2 3 4 5 6 DEF1 2 3 4 5 6 DEF1 2 3 4 5 6 DEF1 2 3 4 5 6 DEF1 2 3 4 5 6 DEFNOTES Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Note 4: The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required. Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents. Note 8: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Note 9: One full train of containment heat removal systems consists of one train of RB [Containment] Spray and one train of RB [Containment] Cooling System. Note 10: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted. Note 11: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted. Note 12: Once PTS limits are first invoked, if RCS temperature and pressure are not brought within the limits within 15 minutes, this threshold is met and an immediate declaration is warranted. This threshold is met immediately upon exceeding the limits after this initial 15 minute period until PTS limits no longer apply. to 0CAN031801 Page 7 of 22 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT H Hazards 4 Fire

None None [Refer to CA6.1 or SA9.1 for potential escalation due to a fire] None Fire potentially degrading the level of safety of the plant HU4.1 A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (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 alarm AND The FIRE is located within any Table 1H-1 area HU4.2 Receipt of a single fire alarm (i.e., no other indications of a FIRE) AND The fire alarm is indicating a FIRE within any Table 1H-1 area AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1) HU4.3 A FIRE within the PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1) HU4.4 A FIRE within the PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish 5 Hazardous Gases None None Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown None HA5.1 Release of a toxic, corrosive, asphyxiant or flammable gas into any Table 1H-2 room or area AND Entry into the room or area is prohibited or IMPEDED (Note 5) 6 Control Room Evacuation None Inability to control a key safety function from outside the Control Room Control Room evacuation resulting in transfer of plant control to alternate locations None HS6.1 An event has resulted in plant control being transferred from the Control Room to alternate locations AND Control of any of the following key safety functions is not re-established within 15 min. (Note 1): - Reactivity (Modes 1, 2 and 3 only) - Core cooling - RCS heat removal HA6.1 An event has resulted in plant control being transferred from the Control Room to alternate locations 1 2 3 4 5 6 DEF1 2 3 4 5 6 DEF3 4DEFReactor Building All elevations Auxiliary Building All elevations including: Penthouse/MSIV Room Exceptions: Boric Acid Mix Tank Room (Chem Add Area), 404' (157-B), EDG Exhaust Fan area on 386' (1-E and 2-E)

Turbine Building All elevations including: Pipechase under ICW Coolers, CRD Pump Pit/T-28 Room/Area under ICW Pumps Outside Areas Manholes adjacent to Startup #2 XFMR (MH-03/MH-04) Manholes adjacent to Intake Structure (MH-05/MH-06)

Intake Structure (354' and 366')

Diesel Fuel Vault Diesel Fuel Vault Pump Manholes (MH-09 and MH-10) Table 1H-1 Unit 1 Fire Areas Table 1H-2 Unit 1 Safe Operation & Shutdown Rooms/Areas A-4 Switchgear Room Upper North Electrical Penetration Room Lower South Electrical Equipment Room 3, 4 3, 4 3, 4 Room/Area Mode to 0CAN031801 Page 8 of 22 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT H Hazards 7 ED Judgment Other conditions exist that in the judgment of the Emergency Director warrant declaration of a GENERAL EMERGENCY Other conditions exist that in the judgment of the Emergency Director warrant declaration of a SITE AREA EMERGENCY Other conditions exist that in the judgment of the Emergency Director warrant declaration of an ALERT Other conditions exist that in the judgment of the Emergency Director warrant declaration of a UE HG7.1 Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area. HS7.1 Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY. HA7.1 Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. HU7.1 Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which indicate a potential degradation of the level of 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 expected unless further degradation of SAFETY SYSTEMS occurs. 1 2 3 4 5 6 DEF1 2 3 4 5 6 DEF1 2 3 4 5 6 DEF1 2 3 4 5 6 DEF to 0CAN031801 Page 9 of 22 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT S System Malfunct. 1 Loss of Vital AC Power Prolonged loss of all offsite and all onsite AC power to vital buses Loss of all offsite and all onsite AC power to vital buses for 15 minutes or longer Loss of all but one AC power source to vital buses for 15 minutes or longer Loss of all offsite AC power capability to vital buses for 15 minutes or longer SG1.1 Loss of all offsite and all onsite AC power to vital 4.16 KV buses A3 and A4 AND EITHER: - Restoration of at least one vital 4.16 KV bus in < 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely (Note 1) - CETs > 1200°F SS1.1 Loss of all offsite and all onsite AC power to vital 4.16 KV buses A3 and A4 for 15 min. (Note 1) SA1.1 AC power capability, Table 1S-1, to vital 4.16 KV buses A3 and A4 reduced to a single power source for 15 min. (Note 1) AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS SU1.1 Loss of all offsite AC power capability, Table 1S-1, to vital 4.16 KV buses A3 and A4 for 15 min. (Note 1) Loss of all vital AC and vital DC power sources for 15 minutes or longer SG1.2 Loss of all offsite and all onsite AC power to vital 4.16 KV buses A3 and A4 for 15 min. (Note 1) AND Indicated voltage is < 105 VDC on D01 and D02 vital 125 VDC buses for 15 min. (Note 1) 2 Loss of Vital DC Power None Loss of all vital DC power for 15 minutes or longer None None SS2.1 Indicated voltage is < 105 VDC on D01 and D02 vital 125 VDC buses for 15 min. (Note 1) 3 Loss of Control Room Indications None None UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress UNPLANNED loss of Control Room indications for 15 minutes or longer SA3.1 An UNPLANNED event results in the inability to monitor one or more Table 1S-2 parameters from within the Control Room for 15 min. (Note 1) AND Any significant transient is in progress, Table 1S-3 SU3.1 An UNPLANNED event results in the inability to monitor one or more Table 1S-2 parameters from within the Control Room for 15 min. (Note 1) 4 RCS Activity None None None RCS activity greater than Technical Specification allowable limits SU4.1 Failed Fuel Iodine radiation monitor RI-1237S > 9.0 E5 cpm SU4.2 RCS sample activity > 1.0 µCi/gm dose equivalent I-131 for > 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> (Note 1) OR RCS sample activity > 60 µCi/gm dose equivalent I-131 OR RCS sample activity > 2200 µCi/gm dose equivalent Xe-133 for > 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> (Note 1) 1 2 3 4 1 2 3 41 2 3 41 2 3 41 2 3 41 2 3 41 2 3 41 2 3 4Offisite*Startup Transformer No. 1

• Startup Transformer No. 2

• Unit Auxiliary Transformer (from 22 KV switchyard) Onsite *Unit Auxiliary Transformer (main generator via main transformer) *DG1

• DG2

• AAC Gen Table 1S-1 Unit 1 AC Power Sources*Reactor power

• RCS level

• RCS pressure

• CET temperature

• Level in at least one S/G

• EFW flow to at least one S/G Table 1S-2 Unit 2 Safety System Parameters*Reactor trip

• Runback > 25% thermal power

• Electrical load rejection > 25% electrical load

• Safety injection actuation Table 1S-3 Significant Transients to 0CAN031801 Page 10 of 22 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT S System Malfunct. 5 RCS Leakage None None None RCS leakage for 15 minutes or longer SU5.1 RCS unidentified or pressure boundary leakage > 10 gpm for 15 min. (Note 1) OR RCS identified leakage > 25 gpm for 15 min. (Note 1)OR Reactor coolant leakage to a location outside containment > 25 gpm for 15 min. (Note 1) 6 RPS Failure None Inability to shut down the reactor causing a challenge to core cooling or RCS heat removal Automatic or manual trip fails to shut down the reactor and subsequent manual actions taken at the reactor control consolers are not successful in shutting down the reactor Automatic or manual trip fails to shut down the reactor SS6.1 An automatic or manual trip fails to shut down the reactor as indicated by reactor power > 5% AND All actions to shut down the reactor are not successful as indicated by reactor power > 5% AND EITHER - CETs >1200°F - RCS heat removal cannot be established using steam generators and HPI cooling initiated SA6.1 An automatic or manual trip fails to shut down the reactor as indicated by reactor power > 5% AND Manual trip actions taken at the reactor control console (C03) (manual reactor trip pushbuttons or DROPS) are not successful in shutting down the reactor as indicated by reactor power > 5% (Note 8) SU6.1 An automatic trip did not shut down the reactor as indicated by reactor power > 5% after any RPS setpoint is exceeded AND A subsequent automatic trip or manual trip action taken at the reactor control console (C03) (manual reactor trip pushbuttons or DROPS) is successful in shutting down the reactor as indicated by reactor power 5% (Note 8) SU6.2 A manual trip did not shut down the reactor as indicated by reactor power > 5% after any manual trip action was initiated AND A subsequent automatic trip or manual trip action taken at the reactor control console (C03 ) (manual reactor trip pushbuttons or DROPS) is successful in shutting down the reactor as indicated by reactor power 5% (Note 8) 7 Loss of Comm. None None None Loss of all onsite or offsite communications capabilities SU7.1 Loss of all Table 1S-4 onsite communication methods OR Loss of all Table 1S-4 State and local agency communication methods OR Loss of all Table 1S-4 NRC communication methods 8 CTMT Failure None None None Failure to isolate containment or loss of containment pressure control SU8.1 Any penetration is not closed within 15 min. of an ESAS actuation signal OR Containment pressure > 44.7 psia with < one full train of containment heat removal systems (Note 9) operating per design for 15 min. (Note 1) 1 2 3 41 2 3 41 2 3 411 1 Station radio system ANO plant phone system Gaitronics Telephone Systems:

• Commercial

• Microwave

• Satellite

• VOIP INFORM Notification System

Emergency Notification System (ENS) X X X

X X

X

X System State / Local Onsite NRC Table 1S-4 Communication Methods to 0CAN031801 Page 11 of 22 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT S System Malfunct. 9 Hazardous Event Affecting Safety Systems None None Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode None SA9.1 The occurrence of any Table 1S-5 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER - Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode - Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 10, 11) 1 2 3 4*Seismic event (earthquake) *Internal or external FLOODING event

• High winds or tornado strike

• FIRE

• EXPLOSION

• Other events with similar hazard characteristics as determined by the Shift Manager Table 1S-5 Hazardous Events to 0CAN031801 Page 12 of 22 UNIT 2 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT A Abnorm. Rad Levels / Rad Effluent 1 Rad Effluent Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE Release of gaseous OR liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer AG1.1 Reading on any Table 2A-1 effluent radiation monitor > column "GE" for 15 min. (Notes 1, 2, 3, 4) AG1.2 Dose assessment using actual meteorology indicates doses > 1000 mrem TEDE or 5000 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4) AG1.3 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: - Closed window dose rates > 1000 mR/hr expected to continue for 60 min. - Analyses of field survey samples indicate thyroid CDE > 5000 mrem for 60 min. of inhalation (Notes 1, 2) AS1.1 Reading on any Table 2A-1 effluent radiation monitor > column "SAE" for 15 min. (Notes 1, 2, 3, 4) AS1.2 Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4) AS1.3 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: - Closed window dose rates > 100 mR/hr expected to continue for 60 min. - Analyses of field survey samples indicate thyroid CDE > 500 mrem for 60 min. of inhalation (Notes 1, 2) AA1.1 Reading on any Table 2A-1 effluent radiation monitor > column "ALERT" for 15 min. (Notes 1, 2, 3, 4) AA1.2 Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4) AA1.3 Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses

> 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2) AA1.4 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: - Closed window dose rates > 10 mR/hr expected to continue for 60 min. - Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation (Notes 1, 2) AU1.1 Reading on any Table 2A-1 effluent radiation monitor > column "UE" for 60 min. (Notes 1, 2, 3) AU1.2 Sample analysis for a gaseous or liquid release indicates a concentration or release rate > 2 x ODCM limits for 60 min. (Notes 1, 2) 2 Irradiated Fuel Event Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer Spent fuel pool level at the top of the fuel racks Significant lowering of water level above, or damage to, irradiated fuel UNPLANNED loss of water level above irradiated fuel AG2.1 Spent fuel pool level cannot be restored to at least 379.5 ft. (Alarm 3) on 2LIT-2020-1(2) for 60 min. (Note 1) AS2.1 Lowering of spent fuel pool level to 379.5 ft. (Alarm 3) on 2LIT-2020-1(2) AA2.1 IMMINENT uncovery of irradiated fuel in the REFUELING PATHWAY AA2.2 Damage to irradiated fuel resulting in a release of radioactivity AND High alarm on any Table 2A-2 radiation monitor AA2.3 Lowering of spent fuel pool level to 389.5 ft. (Alarm 2) on 2LIT-2020-1(2) AU2.1 UNPLANNED water level drop in the REFUELING PATHWAY as indicated by low water level alarm, visual observation, or RWT level drop due to makeup demands AND UNPLANNED rise in corresponding area radiation levels as indicated by any of the following radiation monitors: - 2RE-8914 Spent Fuel Area - 2RE-8915 Spent Fuel Area - 2RE-8916 Spent Fuel Area

- 2RE-8912 Containment Incore Instrumentation Table 2A-1 Unit 1 Effluent Monitor Classification Thresholds Gaseous Monitor GE SAE Alert UE Containment Purge 2RX-9820 (SPING 5) 1.88E+01 µCi/cc 1.88E+00 µCi/cc 1.88E-01 µCi/cc 5.84E-04 µCi/cc Radwaste Area 2RX-9825 (SPING 6) 2.35E+01 µCi/cc 2.35E+00 µCi/cc 2.35E-01 µCi/cc 4.35E-04 µCi/cc Fuel Handling Area 2RX-9820 (SPING 7) 6.86E+02 µCi/cc 6.86E+01 µCi/cc 6.86E+00 µCi/cc 6.04E-04 µCi/cc Emergency Penetration Room 2RX-9835 (SPING 8) 5.88E+02 µCi/cc 5.88E+01 µCi/cc 5.88E+00 µCi/cc 1.09E-02 µCi/cc Liquid BMS Liquid Radwaste 2RE-2330 ---- ---- ---- 2.46E+05 cpm Regenerative Waste Discharge 2RE-4423 ---- ---- ---- 2.46E+05 cpm 3 Area Rad Levels Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown Table 2A-2 Unit 1 Fuel Damage Radiation Monitors 2RE-8905 Containment Equipment Hatch Area 2RE-8909 Containment Personnel Access Area 2RE-8912 Containment Incore Instrumentation 2RE-8914 Spent Fuel Area 2RE-8916 Spent Fuel Area 2RE-9825-1 Containment High Range Radiation Monitors 2RE-9825-2 Containment High Range Radiation Monitors 2Rx-9820 (SPING 5) Containment Purge 2Rx-9825 (SPING 6) Radwaste Area 2Rx-9830 (SPING 7) Fuel Handling Area AA3.1 Dose rate > 15 mR/hr in EITHER of the following areas: - Control Room - Central Alarm Station (by survey) AA3.2 An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table 2A-3 room or area (Note 5) None 1 2 3 4 5 6 DEF 1 2 3 4 5 6 DEF 1 2 3 4 5 6 DEF 1 2 3 4 5 6 DEF 1 2 3 4 5 6 DEF 1 2 3 4 5 6 DEF 1 2 3 4 5 6 DEF 1 2 3 4 5 6 DEF 1 2 3 4 5 6 DEF 3 4 Room / Area Mode Aux Building 317' Emergency Core Cooling Rooms 3, 4 Aux Building 317' Tendon Gallery Access 3, 4 Aux Building 335' Charging Pumps/MCC 2B-52 3, 4 Aux Building 354' MCC 2B-62 Area 3, 4 Emergency Diesel Generator Corridor 3, 4 Lower South Piping Penetration Room 3, 4 Aux Building 386' Containment Hatch 3, 4 Table 2A-3 Unit 2 Safe Operation & Shutdown Rooms/Areas to 0CAN031801 Page 13 of 22 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT C Cold SD/Refuel System Malfunct. 1 RCS Level Loss of RCS inventory affecting fuel clad integrity with containment challenged Loss of RCS inventory affecting core decay heat removal capability Significant loss of RCS inventory UNPLANNED loss of RCS inventory CG1.1 RVLMS Levels 1 through 7 indicate DRY AND Any Containment Challenge indication, Table 2C-2 CG1.2 RCS level cannot be monitored for 30 min. (Note 1) AND Core uncovery is indicated by any of the following: - UNPLANNED rise in any Table 2C-1 sump/tank level of sufficient magnitude to indicate core uncovery - Containment High Range Radiation Monitor 2RE-8925-1/8925-2 reading > 10 R/hr - Erratic Source Range Monitor indication AND Any Containment Challenge indication, Table 2C-2 CS1.1 CONTAINMENT CLOSURE not established AND RVLMS Levels 1 through 6 indicate DRY CS1.2 RVLMS Levels 1 through 7 indicate DRY OR RCS level cannot be monitored for 30 min. (Note 1) AND Core uncovery is indicated by any of the following: - UNPLANNED rise in any Table 2C-1 sump/tank level of sufficient magnitude to indicate core uncovery - Containment high range radiation monitor 2RE-8925-1/8925-2 reading > 10 R/hr - Erratic Source Range Monitor indication CA1.1 Loss of RCS inventory as indicated by EITHER: - RVLMS Levels 1 through 5 indicate DRY

- Reactor vessel level 0 in. (L4791/L4792) (bottom of hot leg) CA1.2 RCS level cannot be monitored for 15 min. (Note 1) AND EITHER - UNPLANNED rise in any Table 2C-1 Sump/Tank level due to a loss of RCS inventory - Visual observation of UNISOLABLE RCS leakageCU1.1 UNPLANNED loss of reactor coolant results in RCS water level less than a required lower limit for 15 min. (Note 1) CU1.2 RCS level cannot be monitored AND EITHER - UNPLANNED rise in any Table 2C-1 sump/tank level due to loss of RCS inventory - Visual observation of UNISOLABLE RCS leakage 2 Loss of Vital AC Power None None Loss of all offsite and all onsite AC power to vital buses for 15 minutes or longer Loss of all but one AC power source to vital buses for 15 minutes or longer CA2.1 Loss of all offsite and all onsite AC power to vital 4.16 KV buses 2A3 and 2A4 for 15 min. (Note 1) CU2.1 AC power capability, Table 2C-3, to vital 4.16 KV buses 2A3 and 2A4 reduced to a single power source for 15 min. (Note 1) AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS 3 RCS Temp. None None Inability to maintain plant in cold shutdown UNPLANNED rise in RCS temperature CA3.1 UNPLANNED rise in RCS temperature to > 200°F for > Table 2C-4 duration (Note 1) OR UNPLANNED RCS pressure rise > 10 psig due to a loss of RCS cooling (this EAL does not apply during water-solid plant conditions) CU3.1 UNPLANNED rise in RCS temperature to > 200°F due to loss of decay heat removal capability CU3.2 Loss of all RCS temperature and RCS level indication for 15 min. (Note 1) 4 Loss of Vital DC Power None None None Loss of Vital DC power for 15 minutes or longer CU4.1 Indicated voltage is < 105 VDC on vital 125 VDC buses for 15 min. (Note 1) 5 Loss of Comm. None None None Loss of all onsite or offsite communications capabilities CU5.1 Loss of all Table 2C-5 onsite communication methods OR Loss of all Table 2C-5 State and local agency communication methods OR Loss of all Table 2C-5 NRC communication methods 5 6 5 65 65 65 6 DEF5 6 DEF5 65 65 65 6 DEF to 0CAN031801 Page 14 of 22 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT C Cold SD/Refuel System Malfunct. 6 HazardousEvent Affecting Safety Systems None None Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode None CA6.1 The occurrence of any Table 2C-6 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER - Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode - Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 10, 11)

5 6*CNTMT Sump *Reactor Drain Tank

• LRW Waste Tank (2T-20)

• Holdup Tank

• Aux. Building Sump

• Quench Tank Table 2C-1 Unit 2 Sumps / Tanks Offsite *Startup Transformer No. 3 *Startup Transformer No. 2

• Unit Auxiliary Transformer (backfed from main transformer) Onsite *2DG1 *2DG2

• AAC Gen Table 2C-3 Unit 2 AC Power Sources *CONTAINMENT CLOSURE not established (Note 6) *Containment hydrogen concentration > 3% *UNPLANNED rise in containment pressure Table 2C-2 Containment Challenge Indications *Seismic event (earthquake) *Internal or external FLOODING event

• High winds or tornado strike

• FIRE

• EXPLOSION

• Other events with similar hazard characteristics as determined by the Shift Manager Table 2C-6 Hazardous Events Station radio system ANO plant phone system Gaitronics Telephone Systems: *Commercial

• Microwave

• Satellite

• VOIP INFORM Notification System

Emergency Notification System (ENS) System State / Local Onsite NRC Table 2C-5 Communication Methods X X X

X X

X

X Intact (but not lowered inventory) Not intact OR Lowered inventory RCS Status CONTAINMENT CLOSURE Status Heat-up Duration N/A Established not established Table 2C-4 RCS Heat-up Duration Thresholds 60 min.* 20 min.*

0 min.

• If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable to 0CAN031801 Page 15 of 22 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT E ISFSI Damage to a loaded cask CONFINEMENT BOUNDARY None None Table 2E-1 ISFSI Dose Rates VSC-24 VCC HI-STORM - 200 mrem/hr on the sides - 400 mrem/hr on the top - 700 mrem/hr at the air inlet - 200 mrem/hr at the air outlet - 60 mrem/hr (gamma + neutron) on the top or outlet vent - 600 mrem/hr (gamma +

neutron on the side of the side of the overpack (excluding inlet and outlet ducts) EU1.1 Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask (VSC-24 VCC or HI-STORM overpack) > any Table 2E-1 value

GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT F Fission Product Barriers None FG1.1 Loss of any two barriers AND Loss or potential loss of the third barrier (Table 2F-1) FS1.1 Loss or potential loss of any two barriers (Table 2F-1) FS1.1 Any loss or any potential loss of either Fuel Clad or RCS barrier (Table 2F-1) 1 2 3 4 5 6 DEF 1 2 3 41 2 3 41 2 3 4 to 0CAN031801 Page 16 of 22 Table 2F-1 Fission Product Barrier Threshold Matrix Category Fuel Clad Barrier (FCB) Reactor Coolant System Barrier (RCB) Containment Barrier (CNB) Loss Potential Loss Loss Potential Loss Loss Potential Loss A. RCS or S/G Tube Leakage None FCB1 RVLMS Levels 1 through 7 indicate DRY RCB1 An automatic or manual ESFAS actuation required by EITHER: - UNISOLABLE RCS leakage - S/G tube RUPTURE RCB2 UNISOLABLE RCS leakage or S/G tube leakage > 44 gpm excluding normal reductions in RCS inventory (e.g., letdown, RCP seal leakoff) RCB3 Uncontrolled RCS cooldown (50°F step change which is below 500°F from NOT) AND RCS pressure and temperature are left of line B (200 degrees MTS)

Standard Attachment 1, P-T Limits (Note 12) CNB1 A S/G that is leaking > 44 gpm (excluding normal reductions in RCS inventory) or that is RUPTURED is also FAULTED outside of containment None B Inadequate Heat Removal FCB2 CETs > 1200°F FCB3 CETs > 700°F FCB4 RCS heat removal cannot be established using steam generators AND Once Through cooling initiated None RCB4 RCS heat removal cannot be established using steam generators AND Once Through cooling initiated None CNB2 CETs > 1200°F AND Restoration procedures not effective within 15 min. (Note 1) C CTMT Radiation / RCS Activity FCB5 Containment High Range Radiation Monitor 2RE-8925-1/ 8925-2 > 700 R/hr FCB6 Coolant activity > 300 Ci/gm dose equivalent I-131 None RCB5 Containment High Range Radiation Monitor 2RE-8925-1/8925-2 > 50 R/hr None None CNB3 Containment High Range Radiation Monitor 2RE-8925-1/8925-2 > 12,000 R/hr D CTMT Integrity or Bypass None None None None CNB4 Containment isolation is required AND EITHER: - Containment integrity has been lost based on Emergency Director judgment - UNISOLABLE pathway from Containment to the environment exists CNB5 Indications of RCS leakage outside of Containment CNB6 Containment pressure > 73.7 psia CNB7 Containment hydrogen concentration > 3% CNB8 Containment pressure > 23.3 psia with < one full train of containment heat removal systems (Note 9) operating per design for 15 min. (Note 1) E Emergency Director Judgment FCB7 Any condition in the opinion of the Emergency Director that indicates loss of the Fuel Clad barrier FCB8 Any condition in the opinion of the Emergency Director that indicates potential loss of the Fuel Clad barrier RCB6 Any condition in the opinion of the Emergency Director that indicates loss of the RCS barrier RCB7 Any condition in the opinion of the Emergency Director that indicates potential loss of the RCS barrier CNB9 Any condition in the opinion of the Emergency Director that indicates loss of the Containment barrier CNB10 Any condition in the opinion of the Emergency Director that indicates potential loss of the Containment barrier to 0CAN031801 Page 17 of 22 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT H Hazards 1 Security None HOSTILE ACTION within the PROTECTED AREA HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes Confirmed SECURITY CONDITION or threat HS1.1 A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by ANO Security Shift Supervision HA1.1 A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by ANO Security Shift Supervision OR A validated notification from NRC of an aircraft attack threat within 30 min. of the site HU1.1 A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by ANO Security Shift Supervision OR Notification of a credible security threat directed at the site OR A validated notification from the NRC providing information of an aircraft threat 2 Seismic Event None None [Refer to CA6.1 or SA9.1 for potential escalation due to a seismic event] None Seismic event greater than OBE levels HU2.1 Seismic event > OBE as indicated by annunciation of the 0.10 g acceleration alarm 3 Natural or Technical Hazard None None [Refer to CA6.1 or SA9.1 for potential escalation due to a hazardous event] None Natural or Technological Hazard HU3.1 A tornado strike within the PROTECTED AREA HU3.2 Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component required by Technical Specifications for the current operating mode HU3.3 Movement of personnel within the PROTECTED AREA is IMPEDED due to an event external to the PROTECTED AREA involving hazardous materials (e.g., an offsite chemical spill or toxic gas release) HU3.4 A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7) 1 2 3 4 5 6 DEF1 2 3 4 5 6 DEF1 2 3 4 5 6 DEF1 2 3 4 5 6 DEF1 2 3 4 5 6 DEFNOTES Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Note 4: The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required. Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents. Note 8: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Note 9: One full train of containment heat removal systems consists of one train of RB [Containment] Spray and one train of RB [Containment] Cooling System. Note 10: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted. Note 11: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted. Note 12: Once PTS limits are first invoked, if RCS temperature and pressure are not brought within the limits within 15 minutes, this threshold is met and an immediate declaration is warranted. This threshold is met immediately upon exceeding the limits after this initial 15 minute period until PTS limits no longer apply. to 0CAN031801 Page 18 of 22 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT H Hazards 4 Fire

None None [Refer to CA6.1 or SA9.1 for potential escalation due to a fire] None Fire potentially degrading the level of safety of the plant HU4.1 A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (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 alarm AND The FIRE is located within any Table 2H-1 area HU4.2 Receipt of a single fire alarm (i.e., no other indications of a FIRE) AND The fire alarm is indicating a FIRE within any Table 2H-1 area AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1) HU4.3 A FIRE within the PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1) HU4.4 A FIRE within the PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish 5 Hazardous Gases None None Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown None HA5.1 Release of a toxic, corrosive, asphyxiant or flammable gas into any Table 2H-2 room or area AND Entry into the room or area is prohibited or IMPEDED (Note 5) 6 Control Room Evacuation None Inability to control a key safety function from outside the Control Room Control Room evacuation resulting in transfer of plant control to alternate locations None HS6.1 An event has resulted in plant control being transferred from the Control Room to alternate locations AND Control of any of the following key safety functions is not re-established within 15 min. (Note 1): - Reactivity (Modes 1, 2 and 3 only) - Core cooling - RCS heat removal HA6.1 An event has resulted in plant control being transferred from the Control Room to alternate locations 1 2 3 4 5 6 DEF1 2 3 4 5 6 DEF3 4DEFReactor Building All elevations Auxiliary Building All elevations including: Aux Extension Turbine Building All elevations Outside Areas Intake Structure (354' and 366') Concrete Manhole East, NE of intake (2MH-01)

Concrete Manhole East of Turbine Building next to train bay (2MH-03) Diesel Fuel Vault Diesel Fuel Vault Pump Manholes (MH-09 and MH-10) Table 2H-1 Unit 2Fire Areas Table 2H-2 Unit 2Safe Operation & Shutdown Rooms/Areas Aux Building 317' Emergency Core Cooling Rooms Aux Building 317' Tendon Gallery Access Aux Building 335' Charging Pumps / MCC 2B-52 Aux Building 354' MCC 2B-62 Area Emergency Diesel Generator Corridor Lower South Piping Penetration Room Aux Building 386' Containment Hatch 3, 4 3, 4 3, 4 3, 4 3, 4 3, 4 3, 4 Room/Area Mode to 0CAN031801 Page 19 of 22 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT H Hazards 7 ED Judgment Other conditions exist that in the judgment of the Emergency Director warrant declaration of a GENERAL EMERGENCY Other conditions exist that in the judgment of the Emergency Director warrant declaration of a SITE AREA EMERGENCY Other conditions exist that in the judgment of the Emergency Director warrant declaration of an ALERT Other conditions exist that in the judgment of the Emergency Director warrant declaration of a UE HG7.1 Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area. HS7.1 Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY. HA7.1 Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. HU7.1 Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which indicate a potential degradation of the level of 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 expected unless further degradation of SAFETY SYSTEMS occurs. 1 2 3 4 5 6 DEF1 2 3 4 5 6 DEF1 2 3 4 5 6 DEF1 2 3 4 5 6 DEF to 0CAN031801 Page 20 of 22 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT S System Malfunct. 1 Loss of Vital AC Power Prolonged loss of all offsite and all onsite AC power to vital buses Loss of all offsite and all onsite AC power to vital buses for 15 minutes or longer Loss of all but one AC power source to vital buses for 15 minutes or longer Loss of all offsite AC power capability to vital buses for 15 minutes or longer SG1.1 Loss of all offsite and all onsite AC power to vital 4.16 KV buses 2A3 and 2A4 AND EITHER: - Restoration of at least one vital 4.16 KV bus in < 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely (Note 1) - CETs > 1200°F SS1.1 Loss of all offsite and all onsite AC power to vital 4.16 KV buses 2A3 and 2A4 for 15 min. (Note 1) SA1.1 AC power capability, Table 1S-1, to vital 4.16 KV buses 2A3 and 2A4 reduced to a single power source for 15 min. (Note 1) AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS SU1.1 Loss of all offsite AC power capability, Table 2S-1, to vital 4.16 KV buses 2A3 and 2A4 for 15 min. (Note 1) Loss of all vital AC and vital DC power sources for 15 minutes or longer SG1.2 Loss of all offsite and all onsite AC power to vital 4.16 KV buses A3 and A4 for 15 min. (Note 1) AND Indicated voltage is < 105 VDC on 2D01 and 2D02 vital 125 VDC buses for 15 min. (Note 1) 2 Loss of Vital DC Power None Loss of all vital DC power for 15 minutes or longer None None SS2.1 Indicated voltage is < 105 VDC on 2D01 and 2D02 vital 125 VDC buses for 15 min. (Note 1) 3 Loss of Control Room Indications None None UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress UNPLANNED loss of Control Room indications for 15 minutes or longer SA3.1 An UNPLANNED event results in the inability to monitor one or more Table 2S-2 parameters from within the Control Room for 15 min. (Note 1) AND Any significant transient is in progress, Table 2S-3 SU3.1 An UNPLANNED event results in the inability to monitor one or more Table 2S-2 parameters from within the Control Room for 15 min. (Note 1) 4 RCS Activity None None None RCS activity greater than Technical Specification allowable limits SU4.1 Failed Fuel Iodine radiation monitor 2RITS-4806B > 9.0 E5 cpm SU4.2 RCS sample activity > 1.0 µCi/gm dose equivalent I-131 for > 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> (Note 1) OR RCS sample activity > 60 µCi/gm dose equivalent I-131 OR RCS sample activity > 3100 µCi/gm dose equivalent Xe-133 for > 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> (Note 1) 1 2 3 4 1 2 3 41 2 3 41 2 3 41 2 3 41 2 3 41 2 3 41 2 3 4Offisite*Startup Transformer No. 3

• Startup Transformer No. 2

• Unit Auxiliary Transformer (backfed from main transformer) Onsite *Unit Auxiliary Transformer (main generator via main transformer) *2DG1

• 2DG2

• AAC Gen Table 2S-1 Unit 2 AC Power Sources*Reactor power

• RCS level

• RCS pressure

• CET temperature

• Level in at least one S/G

• EFW flow to at least one S/G Table 1S-2 Unit 2 Safety System Parameters*Reactor trip

• Runback > 25% thermal power

• Electrical load rejection > 25% electrical load

• Safety injection actuation Table 1S-3 Significant Transients to 0CAN031801 Page 21 of 22 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT S System Malfunct. 5 RCS Leakage None None None RCS leakage for 15 minutes or longer SU5.1 RCS unidentified or pressure boundary leakage > 10 gpm for 15 min. (Note 1) OR RCS identified leakage > 25 gpm for 15 min. (Note 1)OR Reactor coolant leakage to a location outside containment > 25 gpm for 15 min. (Note 1) 6 RPS Failure None Inability to shut down the reactor causing a challenge to core cooling or RCS heat removal Automatic or manual trip fails to shut down the reactor and subsequent manual actions taken at the reactor control consolers are not successful in shutting down the reactor Automatic or manual trip fails to shut down the reactor SS6.1 An automatic or manual trip fails to shut down the reactor as indicated by reactor power > 5% AND All actions to shut down the reactor are not successful as indicated by reactor power > 5% AND EITHER - CETs >1200°F - RCS heat removal cannot be established using steam generators and Once Through cooling initiated SA6.1 An automatic or manual trip fails to shut down the reactor as indicated by reactor power > 5% AND Manual trip actions taken at the reactor control console (2C03/2C14) (manual reactor trip pushbuttons or DSS) are not successful in shutting down the reactor as indicated by reactor power > 5% (Note 8) SU6.1 An automatic trip did not shut down the reactor as indicated by reactor power > 5% after any RPS setpoint is exceeded AND A subsequent automatic trip or manual trip action taken at the reactor control console (2C03/2C14)

(manual reactor trip pushbuttons or DSS) is successful in shutting down the reactor as indicated by reactor power 5% (Note 8) SU6.2 A manual trip did not shut down the reactor as indicated by reactor power > 5% after any manual trip action was initiated AND A subsequent automatic trip or manual trip action taken at the reactor control console (2C03/2C14)

(manual reactor trip pushbuttons or DSS) is successful in shutting down the reactor as indicated by reactor power 5% (Note 8) 7 Loss of Comm. None None None Loss of all onsite or offsite communications capabilities SU7.1 Loss of all Table 2S-4 onsite communication methods OR Loss of all Table 2S-4 State and local agency communication methods OR Loss of all Table 2S-4 NRC communication methods 8 CTMT Failure None None None Failure to isolate containment or loss of containment pressure control SU8.1 Any penetration is not closed within 15 min. of an CIAS actuation signal OR Containment pressure > 23.3 psia with < one full train of containment heat removal systems (Note 9) operating per design for 15 min. (Note 1) 1 2 3 41 2 3 41 2 3 411 1 Station radio system ANO plant phone system Gaitronics Telephone Systems:

• Commercial

• Microwave

• Satellite

• VOIP INFORM Notification System

Emergency Notification System (ENS) X X X

X X

X

X System State / Local Onsite NRC Table 2S-4 Communication Methods to 0CAN031801 Page 22 of 22 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT S System Malfunct. 9 Hazardous Event Affecting Safety Systems None None Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode None SA9.1 The occurrence of any Table 2S-5 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER - Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode - Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 10, 11)

1 2 3 4*Seismic event (earthquake) *Internal or external FLOODING event

• High winds or tornado strike

• FIRE

• EXPLOSION

• Other events with similar hazard characteristics as determined by the Shift Manager Table 1S-5 Hazardous Events Enclosure 6 to 0CAN031801 Supporting Referenced Document Pages Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 1 of 46 Table of Contents Section Page 1. PURPOSE ............................................................................................................................2

2. DEVELOPMENT METHODOLOGY AND BASES ................................................................2 2.1 Threshold Limits ..........................................................................................................2 2.2 Effluent Release Points ...............................................................................................5 2.3 Source Term ................................................................................................................8 2.4 Effluent Flow ..............................................................................................................10 2.5 Release Duration .......................................................................................................11 2.6 Meteorology ...............................................................................................................12 3. DESIGN INPUTS ................................................................................................................13 3.1 General Constants and Conversion Factors ..............................................................13 3.2 Liquid Effluent ............................................................................................................13 3.3 Gaseous Effluent .......................................................................................................14

4. CALCULATIONS ................................................................................................................16 4.1 AU1.1 Liquid Release ................................................................................................16 4.2 AU1.1 Gaseous Release ...........................................................................................17 4.3 AA1.1, AS1.1 and AG1.1 Gaseous Release .............................................................17

5. CONCLUSIONS .................................................................................................................18 6. REFERENCES ...................................................................................................................18 ATTACHMENTS Attachment 1, AU1.1 Liquid Effluent EAL Calculations .........................................................20 , AU1.1 Gaseous Effluent EAL Calculations .....................................................21 Attachment 3, AA1.1, AS1.1 and AG1.1 URI Gaseous Effluent EAL Calculations ................22

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 2 of 46 1. PURPOSE The Arkansas Nuclear One (ANO) Emergency Action Level (EAL) Technical Bases Manual contains background information, event declaration thresholds, bases and references for the EAL and Fission Product Barrier (FPB) values used to implement the Nuclear Energy Institute (NEI) 99-01 Revision 6 EAL guidance. This calculation document provides additional technical detail specific to the derivation of the gaseous and liquid radiological effluent EAL values developed in accordance with the guidance in NEI 99-01 Revision 6.

Documentation of the assumptions, calculations and results are provided for the Ax1 series EAL effluent monitor values associated with the following NEI 99-01 Revision 6 EALs: NEI EAL AU1.1 (gaseous and liquid) NEI EAL AA1.1 (gaseous and liquid) NEI EAL AS1.1 (gaseous) NEI EAL AG1.1 (gaseous) 2. DEVELOPMENT METHODOLOGY AND BASES 2.1 Threshold Limits 2.1.1 AU1.1 Liquid Threshold Limits Guidance Criteria AU1 addresses a release of gaseous or liquid radioactivity greater than 2 times the Offsite Dose Calculation Manual (ODCM) limits for 60 minutes or longer. ANO Bases ODCM Section L 2.3.1 states that the concentration limits for the radioactive liquid effluents released to the discharge canal are as follows: Less than or equal to the concentrations specified in 10 CFR Part 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases Less than or equal to 2.0E-04 Ci/ml total activity for dissolved and entrained noble gases The site specific AU1.1 liquid effluent EAL threshold values will equate to 2 times the ODCM limit.

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 3 of 46 2.1.2 AU1.1 Gaseous Threshold Limits Guidance Criteria AU1 addresses a release of gaseous or liquid radioactivity greater than 2 times the Offsite Dose Calculation Manual (ODCM) limits for 60 minutes or longer. ANO Bases ODCM Section L 2.4.1 states that the dose rate limits for the radioactive gaseous effluents released from the site to unrestricted areas are as follows: Less than or equal to 500 mrem/yr to the total body (Noble Gasses) Less than or equal to 3000 mrem/yr to the skin (Noble Gasses) Less than or equal to 1500 mrem/yr to any organ (I-131, H-3, and for all radionuclides in particulate form with half-lives greater than 8 days) ODCM gaseous setpoint calculations are based on the noble gas limits. Organ dose includes inhalation, ingestion and deposition pathways and are applied in unrestricted area site boundary gaseous effluent dose calculations used in the Annual Radioactive Effluent Release Report. Ingestion pathway bases are not compatible or directly comparable with short term event considerations, and are not a significant contribution to the total dose (total body or skin dose limits from noble gas are the major exposure pathway). Thus, the organ dose limit is not applicable for EAL threshold determination. The site specific AU1.1 gaseous effluent EAL threshold values will equate to 2 times the ODCM limit for the lesser of the total body or skin exposure pathways.

2.1.3 AA1.1 Liquid Threshold Limits Guidance Criteria AA1 addresses a release of radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE. This is based on values at 1% of the EPA Protective Action Guides (PAGs). Per NEI 99-01, the effluent monitor readings should correspond to the above dose limits at the "site-specific dose receptor point" (consistent with the calculation methodology employed) for one hour of exposure.

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 4 of 46 ANO Bases The liquid effluent limits are based on the water concentration values given in 10 CFR 20 Appendix B Table 2 Column 2 (see Section 2.1.1 above). The 10 CFR 20 values are equivalent to the radionuclide concentrations which, if ingested continuously over the course of a year, would produce a total effective dose equivalent of 0.05 rem (50 mrem). The EPA PAGs are based on a TEDE dose from immersion, inhalation and deposition.

The 10 CFR 20 limits and the EPA limits do not represent the same type of exposure and thus cannot be compared on a one to one basis. Thus, the site specific EALs will not contain an AA1.1 liquid effluent monitor threshold value that equates to 1% of the EPA PAG. However, EALs AA1.3 (liquid effluent sample analysis) and AA1.4 (field survey results) will remain applicable for liquid effluent releases that exceed their respective thresholds. 2.1.4 AA1.1 Gaseous Threshold Limits Guidance Criteria AA1 addresses a release of radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE. Per NEI 99-01, the effluent monitor readings are based on values at 1% of the EPA Protective Action Guides (PAGs) at the "site-specific dose receptor point" (consistent with the calculation methodology employed) for one hour of exposure. ANO Bases The gaseous effluent limits for AA1.1 are based on values that equate to an offsite dose greater than 10 mrem TEDE or 50 mrem CDE thyroid, which are 1% of the EPA PAGs.

2.1.5 AS1.1 Gaseous Threshold Limits Guidance Criteria AS1 addresses a release of radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE. This is based on values at 10% of the EPA Protective Action Guides (PAGs) at the "site-specific dose receptor point" (consistent with the calculation methodology employed) for one hour of exposure. ANO Bases The gaseous effluent limits for AS1.1 are based on values that equate to an offsite dose greater than 100 mrem TEDE or 500 mrem CDE thyroid, which are 10% of the EPA PAGs.

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 5 of 46 2.1.6 AG1.1 Gaseous Threshold Limits Guidance Criteria AG1 addresses a release of radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE. This is based on values at 100% of the EPA Protective Action Guides (PAGs) at the "site-specific dose receptor point" (consistent with the calculation methodology employed) for one hour of exposure. ANO Bases The gaseous effluent limits for AG1.1 are based on values that equate to an offsite dose greater than 1,000 mrem TEDE or 5,000 mrem CDE thyroid, which are 100% of the EPA PAGs.

2.2 Effluent Release Points Note - All effluent release points assume a background reading of zero to conservatively account for all modes of operation applicable to the EALs.

2.2.1 Liquid Release Points Guidance Criteria Per NEI 99-01, the AU1 IC addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways (EAL AU1.1) and planned batch releases from non-continuous release pathways (EAL AU1.2). Per NEI 99-01, the AA1 IC includes events or conditions involving a radiological release, whether gaseous or liquid, monitored or un-monitored. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. The "site-specific monitor list and threshold values" should be determined with consideration of the selection of the appropriate installed gaseous and liquid effluent monitors. ANO Bases There are three monitored liquid effluent lines that discharge to the environment at ANO (ODCM Section 2.1): ANO-1: RE-4642 - Liquid Radwaste Monitor ANO-2: 2RE-2330 - Liquid Radwaste Monitor 2RE-4423 - Liquid Radwaste Monitor Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 6 of 46 2.2.2 Gaseous Release Points Guidance Criteria Per NEI 99-01, the AU1 IC addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways (EAL #1) and planned batch releases from non-continuous release pathways (EAL #2). Per NEI 99-01, the AA1 IC includes events or conditions involving a radiological release, whether gaseous or liquid, monitored or un-monitored. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. Per NEI 99-01, the AS1 and AG1 ICs address monitored and un-monitored releases of gaseous radioactivity. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. The "site-specific monitor list and threshold values" should include the effluent monitors described in emergency plan and emergency dose assessment procedures. ANO Bases There are ten monitored gaseous effluent lines that discharge to the environment at ANO (ODCM Section 3.1.2.a): ANO-1: RX-9820 (SPING 1) - Unit 1 Containment Purge Exhaust RX-9825 (SPING 2) - Unit 1 Radwaste Area Exhaust (RE-4830 - Waste Gas Holdup System Monitor is upstream to RX-9825) RX-9830 (SPING 3) - Unit 1 Fuel Handling Area Exhaust RX-9835 (SPING 4) - Unit 1 Penetration Room Exhaust ANO-2: 2RX-9820 (SPING 5) - Unit 2 Containment Purge Exhaust (2RE-8233 - Containment Building Purge Monitor is upstream to 2RX-9820) 2RX-9825 (SPING 6) - Unit 2 Radwaste Area Exhaust (2RE-2429 - Waste Gas Holdup System Monitor is upstream to 2RX-9825) 2RX-9830 (SPING 7) - Unit 2 Fuel Handling Area Exhaust 2RX-9835 (SPING 8) - Unit 2 Penetration Room Exhaust 2RX-9845 (SPING 10) - Auxiliary Building Extension Exhaust (ABE) Per STM 2 2, Turbine Building ABE Ventilation, the most probably release of radioactivity through this pathway would be due to a primary-to-secondary leak where radioactivity from the steam generators would be sent to the Start Up / Blowdown DI and possibly released via the Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 7 of 46 ABE ventilation. However, Operations has administrative controls in place via Abnormal Operating Procedures (AOP) and Emergency Operating Procedures (EOP) to control / limit any release via this pathway due to this type of event. Radioactivity from the Auxiliary Building Extension (ABE) exhaust is normally below detectable limits. Review of the 2017 Unit 2 gaseous release permits revealed that the weekly analyses of the ABE pathway did not contain any measurable radioactivity from the vent and thus no release permits were required during 2017. Since this pathway does not typically discharge radioactivity above detection limits during normal operations, the pathway does not meet the NEI 99-01 criteria for an EAL threshold: normally occurring continuous radioactivity release or a planned batch release from non-continuous release pathways.

Therefore, no threshold will be developed for EAL AU1.1 for this pathway. EALs AU1.2 and AU1.3 will continue to be valid for a release from the ABE. 2RX-9850 (SPING 11) - Low Level Radwaste Building Exhaust Per U2 SAR, Section 11.5.6.1, Low Level Radioactive Waste Storage Building (LLRWSB), the LLRWSB is designed to provide a controlled environment for receiving and shipping, inspection, equipment sorting, compaction, and decontamination activities associated with on-site storage and off-site shipment of LLRW. The only potential release of radioactivity would occur during compacting operations; however, this process is not currently used. Radioactivity from the Low Level Radwaste Building exhaust is normally below detectable limits. Review of the 2017 Unit 2 gaseous release permits revealed that the LLRWSB weekly analyses did not contain any measurable radioactivity from the vent and thus no release permits were required during 2017. Since this pathway does not typically discharge radioactivity above detection limits during normal operations, the pathway does not meet the NEI 99-01 criteria for an EAL threshold:

normally occurring continuous radioactivity release or a planned batch release from non-continuous release pathways. Therefore, no threshold will be developed for EAL AU1.1 for this pathway. EALs AU1.2 and AU1.3 will continue to be valid for a release from the LLRWSB.

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 8 of 46 2.3 Source Term 2.3.1 AU1.1 Liquid Source Term Guidance Criteria NEI 99-01 does not provide specific guidance for AU1 liquid source term assumptions. ANO Bases The ODCM setpoint method calls for a sample to be obtained and a gross gamma (Cs-137 equivalent) count and a gamma isotopic analysis performed. Since liquid releases vary in isotopic composition, Cs-137 is considered a reasonable representative isotope and is used as the assumed source term for the liquid effluent EAL calculations. 2.3.2 AU1.1 Gaseous Source Term Guidance Criteria NEI 99-01 does not provide specific guidance for AU1 gaseous source term assumptions. ANO Bases The gaseous source term is based upon the NUREG-1940 Table 1-6 noble gas fraction of activity available at shutdown. 2.3.3 AA1.1, AS1.1 and AG1.1 Gaseous Source Terms Guidance Criteria NEI 99-01 specifies that the calculation of monitor readings will require use of an assumed release isotopic mix; the selected mix should be the same for ICs AA1, AS1 and AG1. ANO Bases The AA1.1, AS1.1 and AG1.1 gaseous EAL thresholds are based upon the ANO URI dose model results using input assumptions applicable to the event, pathway and particular monitor. The source term used in the URI dose model is taken from NUREG-1940 Table 1.1 (URI Requirements Specification Appendix A Section A.2). The process reductions used in the URI dose model are taken from NUREG-1228 and NUREG-1465 (URI Requirements Specification Appendix A Sections A.4 and A.5). Note - HUT is hold-up time.

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 9 of 46 URI input assumptions for the gaseous release points are as follows: Unit 1 RCS Containment HUT < 2 hrs Sprays On CP Filter Working CP Exhaust Env RX-9820 (SPING 1) is modeled to release path 'A' utilizing an event with fuel clad damage. RCS Containment HUT < 2 hrs Sprays On Aux Building HUT < 2 hrs RWA Filter Working RWA Exhaust Env RX-9825 (SPING 2) is modeled to release path 'D' utilizing an event with fuel clad damage. SFP Fuel Building HUT < 2 hrs FHA Filter Working SFA Exhaust Env RX-9830 (SPING 3) is modeled to release path 'M' utilizing an event with underwater fuel clad damage. RCS Containment HUT < 2 hrs Sprays On Aux Building HUT < 2 hrs EPPR FilterWorking PPR Exhaust Env RX-9835 (SPING 4) is modeled to release path 'C' utilizing an event with fuel clad damage. Unit 2 RCS Containment HUT < 2 hrs Sprays On CP Filter Working CP Exhaust Env 2RX-9820 (SPING 5) is modeled to release path 'A' utilizing an event with fuel clad damage. RCS Containment HUT < 2 hrs Sprays On Aux Building HUT < 2 hrs RWA Filter Working RWA Exhaust Env 2RX-9825 (SPING 6) is modeled to release path 'D' utilizing an event with fuel clad damage.

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 10 of 46 SFP Fuel Building HUT < 2 hrs FHA Filter Working SFA Exhaust Env 2RX-9830 (SPING 7) is modeled to release path 'M' utilizing an event with underwater fuel clad damage. RCS Containment HUT < 2 hrs Sprays On Aux Building HUT < 2 hrs EPPR FilterWorking PPR Exhaust Env 2RX-9835 (SPING 8) is modeled to release path 'C' utilizing an event with fuel clad damage. For RCS initiated accidents, 0 time after shutdown (TAS) is used as no credit is taken for source term decay. For the spent fuel accident, the new fuel age option is used with 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> for time after shutdown (TAS).

2.4 Effluent Flow 2.4.1 Effluent Liquid Discharge Flow Guidance Criteria NEI 99-01 does not provide specific guidance for effluent liquid flow assumptions. ANO Bases Discharge Flow Unit 1 SAR Section 11.1.1 describes the liquid waste processing system for Unit 1. The radioactive waste disposal systems are designed to collect, store, process, monitor, and safely dispose all liquids, gases and solids which are potentially radioactive. U1 SAR Table 11-6 discharge flows are as follows: Treated Waste Pumps, P-47A/B - 85 gpm Max Flow Rate to Discharge Flume Filtered Waste Pumps P-53A/B - 38 gpm Max Flow Rate to Discharge Flume Laundry Drain Pump, P 50 gpm Max Flow Rate to Discharge Flume A representative discharge flow rate of 85 gpm is used as the input for purposes of the U1 liquid effluent EAL calculations. Unit 2 SAR Section 11.2.1 describes the liquid waste processing system for Unit 2.

Radioactive liquid wastes which are discharged from the plant are first processed by the Waste Management System (WMS) or the Boron Management System (BMS). The Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 11 of 46 contents of turbine building sumps and detergent wastes are routinely discharged unprocessed due to their very small potential for radioactive contamination. Discharge flows per U2 SAR are as follows: Treated Waste Pumps, 2P-47A/B - 50 gpm Max Flow Rate to Discharge Flume (Table 11.2-2) Treated Waste Pumps, 2P-53A/B - 50 gpm Max Flow Rate to Discharge Flume (Table 11.2-8) Regenerative Waste Tank Pumps, 2P-135A/B - 100 gpm Max Flow Rate to Discharge Flume (Table 11.2-23) A representative discharge flow rate of 100 gpm is used as the input for purposes of the U2 liquid effluent EAL calculations. Dilution Flow Per ODCM Section 2.1.4, dilution volume is the number of circulating water pumps in operation multiplied by the approximate flowrate of a circulating water pump (normally 191,500 gpm) or an indicated flow rate. A normal dilution flow rate of one circulating water pump (191,500 gpm) is selected for EAL calculations.

2.4.2 Effluent Gaseous Vent Flow Guidance Criteria NEI 99-01 does not provide specific guidance for effluent gaseous vent flow assumptions. ANO Bases Vent flow values for AU1.1, AA1.1, AS1.1 and AG1.1 use the URI default flow values. 2.5 Release Duration Guidance Criteria Per NEI 99-01, the effluent monitor readings for AA1.1, AS1.1 and AG1.1 gaseous EAL threshold values should correspond to a dose at the "site-specific dose receptor point" (consistent with the calculation methodology employed) for one hour of exposure. ANO Bases The effluent monitor readings for AA1.1, AS1.1 and AG1.1 gaseous EAL threshold values are calculated for a release duration of one hour.

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 12 of 46 2.6 Meteorology Guidance Criteria The effluent monitor readings should correspond to the applicable dose limit at the "site-specific dose receptor point." The "site-specific dose receptor point" is the distance(s) and/or locations used by the licensee to distinguish between on-site and offsite doses. The selected distance(s) and/or locations should reflect the content of the emergency plan, and the procedural methodology used to determine offsite doses and protective action recommendations. This is typically the boundary of the Owner Controlled Area. Monitor readings will be calculated using a set of assumed meteorological data or atmospheric dispersion factors; the data or factors selected for use should be the same for ICs AA1, AS1 and AG1. ANO Bases The site specific meteorology used for the EAL calculation inputs are based upon the FSAR and ODCM as documented below. 2.6.1 ODCM Gaseous Dispersion Factor (U1 SAR 2.3.6.2.5) The highest annual average X/Q value for a ground level release, derived using the methodology of Regulatory Guide 1.111, is 2.0E-5 sec/m3 in the west-southwest sector at a distance of 0.65 miles. For purposes of EALs thresholds the SAR historical stability class is used as a basis and any changes in SAR stability results need not require a recalculation of this input.

2.6.2 ODCM Gaseous Dispersion Factor (U1 SAR 2.3.6.2.5)

The predominant stability class of 'E' is determined from the seasonal and annual percent frequency tables. For purposes of EAL thresholds a long term historical meteorological stability of class E is used as a basis and any changes in seasonal or annual stability results need not require a redetermination of this input.

2.6.3 Wind Speed 190' (elevated) wind speed having the greatest number of intervals for all stability classes is 5 kts or 5.75 mph (U1 SAR Table 2-40). 40' (ground level) wind speed having the greatest number of intervals for all stability classes is 4 mph (U1 SAR Table 2-41). For purposes of EALs thresholds the SAR historical meteorological wind speeds are used as a basis and any changes in annual predominant wind speed results need not require a recalculation of this input.

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 13 of 46 2.6.4 Wind Direction 190' (elevated) wind direction having the greatest number of intervals for all stability classes is sector E (U1 SAR Table 2-40). 40' (elevated) wind direction having the greatest number of intervals for all stability classes is sector E (U1 SAR Table 2-41). For purposes of EALs thresholds the SAR historical meteorological wind directions are used as a basis and any changes in annual predominant wind direction results need not require a recalculation of this input.

2.6.5 Other Parameters No precipitation is assumed to occur for the duration of the release and plume transport across the EPZ.

3. DESIGN INPUTS 3.1 General Constants and Conversion Factors 3.1.1 472 cc/sec per cfm

3.1.2 106 Ci per Ci 3.2 Liquid Effluent 3.2.1 Liquid Effluent Monitor Ranges 1) RE-4642 (STM 1-62 Table 62.2, TDL185 0080/0090) ................... 1E+0 to 1E+8 cpm

2) 2RE-2330 (TDW120 2020 Section 1.1.2.1) ................................... 1E+0 to 1E+6 cpm 3) 2RE-4423 (TDW120 2020 Section 1.1.2.1) ................................... 1E+0 to 1E+6 cpm 3.2.2 Offset Factor (EMS Report) 1) RE-4642 .............................................................................................. 3.027E+07 cpm

2) 2RE-2330 ............................................................................................ 2.054E+06 cpm 3) 2RE-4423 ............................................................................................ 2.054E+06 cpm 3.2.3 Slope Factor (EMS Report) 1) RE-4642 ............................................................................. 9.572E-01 cpm per µCi/ml

2) 2RE-2330 ........................................................................... 9.178E-01 cpm per µCi/ml

3) 2RE-4423 ........................................................................... 9.178E-01 cpm per µCi/ml Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 14 of 46 3.2.4 Liquid Effluent Dilution Flowrate - DV (ODCM 2.1.4) 1) Minimum Flow ......................................................................................... 191,500 gpm 3.2.5 Liquid Effluent Actual Flowrate - FA 1) RE-4642 Liquid Radwaste Discharge (see section 2.4.1) ............................... 85 gpm 2) 2RE-2330 Liquid Radwaste Discharge (see section 2.4.1) ........................... 100 gpm 3) 2RE-4423 Liquid Radwaste Discharge (see section 2.4.1) ........................... 100 gpm 3.2.6 10 CFR 20 Appendix B, Table 2, Column 2 Source Term Limit (MPCi) Cs-137 .............................................................................................................. 1E-6 Ci/ml 3.3 Gaseous Effluent 3.3.1 Gaseous Effluent Monitor Ranges (TDE070 0290) Note - Containment Purge, Radwaste Area, Fuel Handling Area and Penetration Room SPING monitors have identical ranges. 1) Low Range ............................................................. 1.12E-7 to 3.63E-2 Ci/cc Xe-133 2) Medium Range ...................................................... 4.16E-5 to 6.76E+1 Ci/cc Xe-133 3) High Range ........................................................... 4.57E-3 to 5.51E+3 Ci/cc Xe-133 3.3.2 Gaseous Effluent Source Flow (URI default values) 1) Unit 1 RX-9820 (Containment Purge) ...................................................... 1.8E+04 cfm RX-9825 (Radwaste Area) ............................................................ 4.4E+04 cfm RX-9830 (Fuel Handling Area) ...................................................... 4.0E+04 cfm RX-9835 (Penetration Room) ........................................................ 1.8E+03 cfm 2) Unit 2 2RX-9820 (Containment Purge) .................................................. 3.97E+04 cfm 2RX-9825 (Radwaste Area) .......................................................... 5.0E+04 cfm 2RX-9830 (Fuel Handling Area) .................................................... 3.6E+04 cfm 2RX-9835 (Penetration Room) ...................................................... 2.0E+03 cfm 3.3.3 AU1.1 Dispersion Factor (X/Q)

Dispersion Factor (SAR 2.3.6.2.5) .............................................................. 2.0E-05 sec/m3 Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 15 of 46 3.3.4 AU1.1 Gaseous Source Term Fraction (Ai) NUREG-1940 Table 1-6 noble gas fraction of activity available at shutdown. Isotopic Fraction Ai (unitless) Kr-83m 1.83E-02 Kr-85 1.70E-03 Kr-85m 3.71E-02 Kr-87 7.40E-02 Kr-88 1.02E-01 Xe-131m 2.20E-03 Xe-133 3.26E-01 Xe-133m 1.03E-02 Xe-135 8.54E-02 Xe-135m 6.90E-02 Xe-138 2.74E-01 1.00E+00 3.3.5 ODCM Dose Factors (Regulatory Guide 1.109 Table B-1) Note - RG 1.109 values converted from mRem/yr per Ci/m3 to mRem/yr per µCi/m3. Total Body Dose Factor - Ki (mRem/yr / µCi/m3) Skin Beta Dose Factor - Li (mRem/yr / µCi/m3) Gamma Air Dose Factor - Mi (mRad/yr / µCi/m3) Kr-83m 7.56E-02 0.00E+00 1.93E+01 Kr-85 1.61E+01 1.34E+03 1.72E+01 Kr-85m 1.17E+03 1.46E+03 1.23E+03 Kr-87 5.92E+03 9.73E+03 6.17E+03 Kr-88 1.47E+04 2.37E+03 1.52E+04 Xe-131m 9.15E+01 4.76E+02 1.56E+02 Xe-133 2.94E+02 3.06E+02 3.53E+02 Xe-133m 2.51E+02 9.94E+02 3.27E+02 Xe-135 1.81E+03 1.86E+03 1.92E+03 Xe-135m 3.12E+03 7.11E+02 3.36E+03 Xe-138 8.83E+03 4.13E+03 9.21E+03

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 16 of 46 4. CALCULATIONS 4.1 AU1.1 Liquid Release 4.1.1 ODCM Liquid Release Limit (ODCM Section 2.1) Where: ML Radiation monitor setpoint equivalent to the ODCM limit (cpm) A Allocation fraction for the specific unit (1 selected for EAL calculations) Offset Detector response for the minimum detectable sample activity calculated from the calibration data (cpm) SA Gross gamma (Cs-137 equivalent) activity for the tank (µCi/ml) (Cs-137 10 CFR 20 limit of 1E-6 Ci/ml selected for EAL calculations) Slope Log of the detector response (cpm) / Log of activity concentration (µCi/ml) DV Dilution volume flowrate (gpm) Ci Concentration of isotope "i" (µCi/ml) (Cs-137 10 CFR 20 limit of 1E-6 Ci/ml selected for EAL calculations) MPCi 10 CFR 20 isotope "i" concentration limit (Ci/ml) (Cs-137 10 CFR 20 limit of 1E-6 Ci/ml selected for EAL calculations) CTNG Total concentration of noble gases (µCi/ml) (No noble gas assumed in the liquid discharge for EAL calculations) MPCTNG 10 CFR 20 Total Noble Gas limit of 2E-4 (Ci/ml) FA Actual flowrate (gpm) - maximum flowrate of the discharge pump B background countrate prior to the release (cpm) (0 cpm selected for EAL calculations) 4.1.2 AU1.1 Liquid Release EAL Threshold AU1.1 liquid is two times (2x) the calculated ODCM limit setpoint. See Attachment 1 for the spreadsheet calculations that develop the AU1.1 liquid effluent EAL threshold values for each applicable monitor.

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 17 of 46 4.2 AU1.1 Gaseous Release 4.2.1 ODCM Gaseous Release Limit 472500 Limitbody total

iiiKAQf

iiiLAQf 1.1M 4723000 LimitiSkin Where: Limit radiation monitor reading equivalent to the ODCM limit (Ci/cc) 500/3000 ODCM Limit - 500 total body or 3000 skin (mrem/yr) 472 conversion factor (cc/ft3 per sec/min) f vent flow (cfm) X/Q highest annual average gaseous site boundary dispersion factor (sec/m3) Ai isotopic fraction of the mix activity released (unitless) Ki total body dose factor (mrem/yr per µCi/m3) Li + 1.1Mi skin dose factor (mrem/yr per µCi/m3) 4.2.2 AU1.1 Gaseous Release EAL Threshold AU1.1 is two times (2x) the lesser of the calculated total body or skin value ODCM limit setpoint. See Attachment 2 for the spreadsheet calculations that develop the AU1.1 gaseous effluent EAL threshold values for each applicable monitor.

4.3 AA1.1, AS1.1 and AG1.1 Gaseous Release The AA1.1, AS1.1 and AG1.1 gaseous release EAL threshold are developed using the site specific URI dose assessment models with the inputs described in Section 2 above. Note - URI calculations were performed for each unit. There was no difference in results between units. Refer to Attachment 3 for the results of the URI gaseous effluent EAL threshold calculations.

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 18 of 46 5. CONCLUSIONS 5.1 Unit 1 Monitor # Monitor Name GE SAE Alert UE Gaseous RX-9820 Containment Purge 4.15E+1 (Ci/cc) 4.15E+0 (Ci/cc) 4.15E-1 (Ci/cc) 1.21E-3 (Ci/cc) RX-9825 Radwaste Area 2.67E+1 (Ci/cc) 2.67E+0 (Ci/cc) 2.67E-1 (Ci/cc) 4.94E-4 (Ci/cc) RX-9830 Fuel Handling Area 6.20E+2 (Ci/cc) 6.20E+1 (Ci/cc) 6.20E+0 (Ci/cc) 5.44E-4 (Ci/cc) RX-9835 Penetration Room 6.55E+2 (Ci/cc) 6.55E+1 (Ci/cc) 6.55E+0 (Ci/cc) 1.21E-2 (Ci/cc) Liquid RE-4642 Liquid Radwaste N/A N/A N/A 2.46E+05 (cpm) 5.2 Unit 2 Monitor # Monitor Name GE SAE Alert UE Gaseous 2RX-9820 Containment Purge 1.88+01 (Ci/cc) 1.88E+00 (Ci/cc) 1.88E-01 (Ci/cc) 5.48E-4 (Ci/cc) 2RX-9825 Radwaste Area 2.35+01 (Ci/cc) 2.35E+00 (Ci/cc) 2.35E-01 (Ci/cc) 4.35E-4 (Ci/cc) 2RX-9830 Fuel Handling Area 6.86E+02 (Ci/cc) 6.86E+01 (Ci/cc) 6.86E+00 (Ci/cc) 6.04E-4 (Ci/cc) 2RX-9835 Penetration Room 5.88E+02 (Ci/cc) 5.88E+01 (Ci/cc) 5.88E+00 (Ci/cc) 1.09E-2 (Ci/cc) Liquid 2RE-2330 Liquid Radwaste N/A N/A N/A 2.45E+04 (cpm) 2RE-4423 Liquid Radwaste N/A N/A N/A 2.45E+05 (cpm)

6. REFERENCES

6.1 10 CFR Part 20, Appendix B, Table 2, Column 2 6.2 Regulatory Guide 1.109 Table B-1

6.3 NEI 99-01 Revision 6, Methodology for Development of Emergency Action Levels, November 2012 6.4 NUREG-1940, RASCAL 4, Description of Models and Methods, December 2012

6.5 NUREG-1228, Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents, October 1988 Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 19 of 46 6.6 ANO Offsite Dose Calculation Manual (ODCM), Revision 25

6.7 Unified RASCAL Interface Requirement Specification, Draft 051611 6.8 Unified RASCAL Interface Requirement Specification ANO Unit 1 Site Annex, Version 1.1 6.9 Unified RASCAL Interface Requirement Specification ANO Unit 2 Site Annex, Version 1.1

6.10 ANO1 Safety Analysis Report (SAR), Amendment 26 6.11 ANO2 Safety Analysis Report (SAR), Amendment 27

6.12 STM 1-62, Radiation Monitoring, Revision 13 6.13 STM 2 2, Turbine Building and Auxiliary Building Extension Ventilation, Revision 15 6.14 TD L185.0080, Operation and Maintenance Samplers, Revision 0

6.15 TD L185 0080.0090, Operation and Maintenance Digital Ratemeter Model DRM-100 and DRM-100S, Revision 4 6.16 TDW120 2020, Installation, Operation, and Maintenance Instructions Switchboard Edgewise Instruments Five Inch Classification 252 Line, Revision 0 6.17 TDE070 0290, Instruction Manual Particulate, Iodine and Noble Gas Air Monitor Model SPING-3A/SPING-4A, Revision 15 6.18 EMS Activity Monitors Report dated 10/07/15

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 20 of 46 ATTACHMENT 1 AU1.1 Liquid Effluent EAL Calculations

MonitorOffset SlopeExpected Countrate (cpm)K=Offset

• S^SlopeActual Flowrate (Fa)Monitor Setpoint - M=A*(K*Fm/Fa)+BRU1.1 EAL Threshold Value (cpm)RE-46423.03E+079.57E-015.47E+01851.23E+052.46E+05 2RE-23302.05E+069.18E-016.39E+001001.22E+042.45E+042RE-44232.05E+069.18E-016.39E+001001.22E+042.45E+04Gross gamma Cs-137 Equivalent Activity - S (µCi/ml):1.00E-06Dilution Volume Flowrate - DV (gpm):1.92E+05Cs-137 Activity - C (µCi/ml):1.00E-06Cs-137 10 CFR 20 Limit - MPC (µCi/ml):1.00E-06Dilution Factor - DF=i(Ci/MPCi) + CTNG/MPCTNG:1.00E+00Theoretical Release Rate - Fm=DV/DF (gpm):1.92E+05Allocation Fraction - A:100%Background (cpm):0 Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 21 of 46 ATTACHMENT 2 AU1.1 Gaseous Effluent EAL Calculations Total Body Dose Factor - Ki(mRem/yr per µCi/m3)Skin Beta Dose Factor - Li(mRem/yr per µCi/m3)Gamma Air Dose Factor - Mi(mRad/yr per µCi/m3)Source Term Fraction - AiAi x Ki(mRem/yr per µCi/m3)Ai x (Li + 1.1Mi)(mRem/yr per µCi/m3)Kr-83m7.56E-020.00E+001.93E+011.83E-021.38E-033.89E-01Kr-851.61E+011.34E+031.72E+011.70E-032.74E-022.31E+00Kr-85m1.17E+031.46E+031.23E+033.71E-024.34E+011.04E+02Kr-875.92E+039.73E+036.17E+037.40E-024.38E+021.22E+03Kr-881.47E+042.37E+031.52E+041.02E-011.50E+031.95E+03Xe-131m9.15E+014.76E+021.56E+022.20E-032.01E-011.42E+00Xe-1332.94E+023.06E+023.53E+023.26E-019.59E+012.26E+02Xe-133m2.51E+029.94E+023.27E+021.03E-022.59E+001.39E+01Xe-1351.81E+031.86E+031.92E+038.54E-021.55E+023.39E+02Xe-135m3.12E+037.11E+023.36E+036.90E-022.15E+023.04E+02Xe-1388.83E+034.13E+039.21E+032.74E-012.42E+033.90E+031.00E+004.87E+038.07E+03Calculation ConstantsTotal Body-DDE (mRem/yr):500Skin Dose-SDE (mRem/yr):3000UCF (CFM to cc/sec):472X/Q (sec/m3):2.00E-05DCF (mRad to mRem):1.1Calculation ResultsUnit 1:RX-9820RX-9825RX-9830RX-9835Flow (cfm):1.80E+044.40E+044.00E+041.80E+03Limit-TB (µCi/cc):6.04E-042.47E-042.72E-046.04E-03Limit-Skin (µCi/cc):2.19E-038.95E-049.85E-042.19E-022x ODCM (µCi/cc):1.21E-034.94E-045.44E-041.21E-02Unit 2:2RX-98202RX-98252RX-98302RX-9835Flow (cfm):3.97E+045.00E+043.60E+042.00E+03Limit-TB (µCi/cc):2.74E-042.18E-043.02E-045.44E-03Limit-Skin (µCi/cc):9.92E-047.88E-041.09E-031.97E-022x ODCM (µCi/cc):5.48E-044.35E-046.04E-041.09E-02 Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 22 of 46

ATTACHMENT 3 AA1.1, AS1.1 and AG1.1 Gaseous Effluent EAL Calculations Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 23 of 46 U1 RX-9820 (SPING 1) Containment Purge Exhaust - Alert Dose Assessment ANO Unit 1 Wednesday, October 07, 2015 15:52 Method: Detailed Assessment - Monitored Release Release Pathway: <RCS> <Containment> <CP Filters> < Env> PRF: 3.00E-03 Containment HUT: = < 2 Hours Containment Sprays: = ON Purge Filters: = Working Steam Gen: = N/A Aux / FHB HUT: = N/A Filters: = N/A Turb Bldg HUT: = N/A Source Term: Reactor Core Accident - Clad On Site 57 m Time After S/D (hh:mm): 0:00 Wind: From 270° @ 5.8 mph Release Duration (hh:mm): 1.:00 ETE (hh:mm): [N/A] Stability Class: E Precipitation: None Monitor: CP Gas Conc Readings: 4.15E-01 µCI/cc Flowrate: 18000 CFM Distance (Miles) Exposure Rate (mR/hr) External Plume DDE (mRem) Inhalation CEDE (mRem) Deposition Ground DDE (mRem) TEDE (mRem) CDE Thyroid (mRem Evacuation Areas From 0 to 10 Miles S.B. 8.00E+00 5.32E+00 1.19E+00 6.77E-01 7.19E+00 5.01E+01 0.7 7.44E+00 4.92E+00 1.11E+00 6.66E-01 6.70E+00 4.56E+01 1.0 5.36E+00 3.47E+00 8.68E-01 5.91E-01 4.93E+00 3.26E+01 1.5 3.55E+00 2.26E+00 6.52E-01 4.47E-01 3.36E+00 2.35E+01 2.0 2.40E+00 1.52E+00 5.12E-01 3.15E-01 2.35E+00 1.86E+01 3.0 2.14E+00 1.41E+00 3.61E-01 2.20E-01 1.99E+00 1.32E+01 4.0 1.63E+00 1.05E+00 2.89E-01 1.67E-01 1.50E+00 1.05E+01 5.0 1.36E+00 8.64E-01 2.53E-01 1.35E-01 1.25E+00 9.29E+00 7.0 9.84E-01 6.27E-01 2.01E-01 0.00E+00 8.29E-01 7.67E+00 10.0 5.72E-01 3.56E-01 1.24E-01 0.00E+00 4.80E-01 5.00E+00 Assessment Data Results Save to File: ANO Unit 1 10Miles Monitored Release 10072015 155234.UR17 No PAGs Exceeded Release Rates (Ci/sec) * *

• Classification: Validate against Emergency Action Levels * *

• Particulate 6.56E-04 (0.0%) Iodine 1.61E-02 (0.5%) Reviewed By: Noble Gas 3.53E+00 (99.5%) P ML KNQUHJ I S OTRG Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 24 of 46 U1 RX-9820 (SPING 1) Containment Purge Exhaust - Site Area Emergency Dose Assessment ANO Unit 1 Wednesday, October 07, 2015 15:52 Method: Detailed Assessment - Monitored Release Release Pathway: <RCS> <Containment> <CP Filters> < Env> PRF: 3.00E-03 Containment HUT: = < 2 Hours Containment Sprays: = ON Purge Filters: = Working Steam Gen: = N/A Aux / FHB HUT: = N/A Filters: = N/A Turb Bldg HUT: = N/A Source Term: Reactor Core Accident - Clad On Site 57 m Time After S/D (hh:mm): 0:00 Wind: From 270° @ 5.8 mph Release Duration (hh:mm): 1.:00 ETE (hh:mm): [N/A] Stability Class: E Precipitation: None Monitor: CP Gas Conc Readings: 4.15E+00 µCI/cc Flowrate: 18000 CFM Distance (Miles) Exposure Rate (mR/hr) External Plume DDE (mRem) Inhalation CEDE (mRem) Deposition Ground DDE (mRem) TEDE (mRem) CDE Thyroid (mRem Evacuation Areas From 0 to 10 Miles S.B. 8.00E+01 5.32E+01 1.19E+01 6.77E+00 7.19E+01 5.01E+02 0.7 7.44E+01 4.92E+01 1.11E+01 6.66E+00 6.70E+01 4.56E+02 1.0 5.36E+01 3.47E+01 8.68E+00 5.91E+00 4.93E+01 3.26E+02 1.5 3.55E+01 2.26E+01 6.52E+00 4.47E+00 3.36E+01 2.35E+02 2.0 2.40E+01 1.52E+01 5.12E+00 3.15E+00 2.35E+01 1.86E+02 3.0 2.14E+01 1.41E+01 3.61E+00 2.20E+00 1.99E+01 1.32E+02 4.0 1.63E+01 1.05E+01 2.89E+00 1.67E+00 1.50E+01 1.05E+02 5.0 1.36E+01 8.64E+00 2.53E+00 1.35E+00 1.25E+01 9.29E+01 7.0 9.84E+00 6.27E+00 2.01E+00 9.23E-01 9.21E+00 7.67E+01 10.0 5.72E+00 3.56E+00 1.24E+00 4.53E-01 5.25E+00 5.00E+00 Assessment Data Results Save to File: ANO Unit 1 10Miles Monitored Release 10072015 155209.UR17 No PAGs Exceeded Release Rates (Ci/sec) * *

• Classification: Site Area Emergency * *

• Particulate 6.56E-03 (0.0%) Iodine 1.61E-01 (0.5%) Reviewed By: Noble Gas 3.53E+01 (99.5%) P ML KNQUHJ IS OTRG Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 25 of 46 U1 RX-9820 (SPING 1) Containment Purge Exhaust - General Emergency Dose Assessment ANO Unit 1 Wednesday, October 07, 2015 15:52 Method: Detailed Assessment - Monitored Release Release Pathway: <RCS> <Containment> <CP Filters> < Env> PRF: 3.00E-03 Containment HUT: = < 2 Hours Containment Sprays: = ON Purge Filters: = Working Steam Gen: = N/A Aux / FHB HUT: = N/A Filters: = N/A Turb Bldg HUT: = N/A Source Term: Reactor Core Accident - Clad On Site 57 m Time After S/D (hh:mm): 0:00 Wind: From 270° @ 5.8 mph Release Duration (hh:mm): 1.:00 ETE (hh:mm): [N/A] Stability Class: E Precipitation: None Monitor: CP Gas Conc Readings: 4.15E+01 µCI/cc Flowrate: 18000 CFM Distance (Miles) Exposure Rate (mR/hr) External Plume DDE (mRem) Inhalation CEDE (mRem) Deposition Ground DDE (mRem) TEDE (mRem) CDE Thyroid (mRem Evacuation Areas From 0 to 10 Miles S.B. 8.00E+01 5.32E+01 1.19E+01 6.77E+00 7.19E+01 5.01E+02 0.7 7.44E+01 4.92E+01 1.11E+01 6.66E+00 6.70E+01 4.56E+02 1.0 5.36E+01 3.47E+01 8.68E+00 5.91E+00 4.93E+01 3.26E+02 1.5 3.55E+01 2.26E+01 6.52E+00 4.47E+00 3.36E+01 2.35E+02 2.0 2.40E+01 1.52E+01 5.12E+00 3.15E+00 2.35E+01 1.86E+02 3.0 2.14E+01 1.41E+01 3.61E+00 2.20E+00 1.99E+01 1.32E+02 4.0 1.63E+01 1.05E+01 2.89E+00 1.67E+00 1.50E+01 1.05E+02 5.0 1.36E+01 8.64E+00 2.53E+00 1.35E+00 1.25E+01 9.29E+01 7.0 9.84E+00 6.27E+00 2.01E+00 9.23E-01 9.21E+00 7.67E+01 10.0 5.72E+00 3.56E+00 1.24E+00 4.53E-01 5.25E+00 5.00E+01 Assessment Data Results Save to File: ANO Unit 1 10Miles Monitored Release 10072015 155647.UR17 No PAGs Exceeded Release Rates (Ci/sec) * *

• Classification: General Emergency * *

• Particulate 6.56E-02 (0.0%) Iodine 1.61E+00 (0.5%) Reviewed By: Noble Gas 3.53E+02 (99.5%) P ML KNQUHJ IS OTRG Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 26 of 46 U1 RX-9825 (SPING 2) Radwaste Area Exhaust - Alert Dose Assessment ANO Unit 1 Wednesday, October 07, 2015 15:59 Method: Detailed Assessment - Monitored Release Release Pathway: <RCS> <Containment> <Aux Bldg> <RWA Filters> < Env> PRF: 1.20E-03 Containment HUT: = < 2 Hours Containment Sprays: = ON Purge Filters: = N/A Steam Gen: = N/A Aux / FHB HUT: = < 2 Hours Filters: = Working Turb Bldg HUT: = N/A Source Term: Reactor Core Accident - Clad On Site 57 m Time After S/D (hh:mm): 0:00 Wind: From 270° @ 5.8 mph Release Duration (hh:mm): 1.:00 ETE (hh:mm): [N/A] Stability Class: E Precipitation: None Monitor: RWA Gas Conc Readings: 2.67E-01 µCI/cc Flowrate: 44000 CFM Distance (Miles) Exposure Rate (mR/hr) External Plume DDE (mRem) Inhalation CEDE (mRem) Deposition Ground DDE (mRem) TEDE (mRem) CDE Thyroid (mRem Evacuation Areas From 0 to 10 Miles S.B. 1.24E+01 8.31E+00 9.39E-01 7.70E-01 1.00E+01 3.15E+01 0.7 1.16E+01 7.68E+00 9.00E-01 7.78E-01 9.36E+00 2.87E+01 1.0 8.36E+00 5.40E+00 7.60E-01 7.38E-01 6.90E+00 2.06E+01 1.5 5.52E+00 3.52E+00 5.92E-01 5.66E-01 4.67E+00 1.49E+01 2.0 3.74E+00 2.38E+00 4.56E-01 3.91E-01 3.22E+00 1.18E+01 3.0 3.33E+00 2.20E+00 3.23E-01 2.78E-01 2.80E+00 8.30E+00 4.0 2.54E+00 1.64E+00 2.60E-01 2.10E-01 2.11E+00 6.61E+00 5.0 2.12E+00 1.35E+00 2.26E-01 1.68E-01 1.74E+00 5.85E+00 7.0 1.54E+00 9.82E-01 1.75E-01 1.09E-01 1.27E+00 4.83E+00 10.0 8.92E-01 5.56E-01 1.03E-01 0.00E+00 6.59E-01 3.15E+00 Assessment Data Results Save to File: ANO Unit 1 10Miles Monitored Release 10072015 155932.UR17 No PAGs Exceeded Release Rates (Ci/sec) * *

• Classification: Validate against Emergency Action Levels * *

• Particulate 4.13E-04 (0.0%) Iodine 1.01E-02 (0.2%) Reviewed By: Noble Gas 5.54E+00 (99.8%) OTG Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 27 of 46 U1 RX-9825 (SPING 2) Radwaste Area Exhaust - Site Area Emergency Dose Assessment ANO Unit 1 Wednesday, October 07, 2015 15:59 Method: Detailed Assessment - Monitored Release Release Pathway: <RCS> <Containment> <Aux Bldg> <RWA Filters> < Env> PRF: 1.20E-03 Containment HUT: = < 2 Hours Containment Sprays: = ON Purge Filters: = N/A Steam Gen: = N/A Aux / FHB HUT: = < 2 Hours Filters: = Working Turb Bldg HUT: = N/A Source Term: Reactor Core Accident - Clad On Site 57 m Time After S/D (hh:mm): 0:00 Wind: From 270° @ 5.8 mph Release Duration (hh:mm): 1.:00 ETE (hh:mm): [N/A] Stability Class: E Precipitation: None Monitor: RWA Gas Conc Readings: 2.67E+00 µCI/cc Flowrate: 44000 CFM Distance (Miles) Exposure Rate (mR/hr) External Plume DDE (mRem) Inhalation CEDE (mRem) Deposition Ground DDE (mRem) TEDE (mRem) CDE Thyroid (mRem Evacuation Areas From 0 to 10 Miles S.B. 1.24E+02 8.31E+01 9.39E+00 7.70E+00 1.00E+02 3.15E+02 0.7 1.16E+02 7.68E+01 9.00E+00 7.78E+00 9.36E+01 2.87E+02 1.0 8.36E+01 5.40E+01 7.60E+00 7.38E+00 6.90E+01 2.06E+02 1.5 5.52E+01 3.52E+01 5.92E+00 5.66E+00 4.67E+01 1.49E+02 2.0 3.74E+01 2.38E+01 4.56E+00 3.91E+00 3.22E+01 1.18E+02 3.0 3.33E+01 2.20E+01 3.23E+00 2.78E+00 2.80E+01 8.30E+01 4.0 2.54E+01 1.64E+01 2.60E+00 2.10E+00 2.11E+01 6.61E+01 5.0 2.12E+01 1.35E+01 2.26E+00 1.68E+00 1.74E+01 5.85E+01 7.0 1.54E+01 9.82E+00 1.75E+00 1.09E+00 1.27E+01 4.83E+01 10.0 8.92E+00 5.56E+00 1.03E+00 4.90E+01 7.08E+00 3.15E+01 Assessment Data Results Save to File: ANO Unit 1 10Miles Monitored Release 10072015 155911.UR17 No PAGs Exceeded Release Rates (Ci/sec) * *

• Classification: Site Area Emergency * *

• Particulate 4.13E-03 (0.0%) Iodine 1.01E-01 (0.2%) Reviewed By: Noble Gas 5.54E+01 (99.8%)

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 28 of 46 U1 RX-9825 (SPING 2) Radwaste Area Exhaust - General Emergency Dose Assessment ANO Unit 1 Wednesday, October 07, 2015 15:58 Method: Detailed Assessment - Monitored Release Release Pathway: <RCS> <Containment> <Aux Bldg> <RWA Filters> < Env> PRF: 1.20E-03 Containment HUT: = < 2 Hours Containment Sprays: = ON Purge Filters: = N/A Steam Gen: = N/A Aux / FHB HUT: = < 2 Hours Filters: = Working Turb Bldg HUT: = N/A Source Term: Reactor Core Accident - Clad On Site 57 m Time After S/D (hh:mm): 0:00 Wind: From 270° @ 5.8 mph Release Duration (hh:mm): 1.:00 ETE (hh:mm): [N/A] Stability Class: E Precipitation: None Monitor: RWA Gas Conc Readings: 2.67E+01 µCI/cc Flowrate: 44000 CFM Distance (Miles) Exposure Rate (mR/hr) External Plume DDE (mRem) Inhalation CEDE (mRem) Deposition Ground DDE (mRem) TEDE (mRem) CDE Thyroid (mRem Evacuation Areas From 0 to 10 Miles S.B. 1.24E+03 8.31E+02 9.39E+01 7.70E+01 1.00E+03 3.15E+03 0.7 1.16E+03 7.68E+02 9.00E+01 7.78E+01 9.36E+02 2.87E+03 1.0 8.36E+02 5.40E+02 7.60E+01 7.38E+01 6.90E+02 2.06E+03 1.5 5.52E+02 3.52E+02 5.92E+01 5.66E+01 4.67E+02 1.49E+03 2.0 3.74E+02 2.38E+02 4.56E+01 3.91E+01 3.22E+02 1.18E+03 3.0 3.33E+02 2.20E+02 3.23E+01 2.78E+01 2.80E+02 8.30E+02 4.0 2.54E+02 1.64E+02 2.60E+01 2.10E+01 2.11E+02 6.61E+02 5.0 2.12E+02 1.35E+02 2.26E+01 1.68E+01 1.74E+02 5.85E+02 7.0 1.54E+02 9.82E+01 1.75E+01 1.09E+01 1.27E+02 4.83E+02 10.0 8.92E+01 5.56E+01 1.03E+01 4.90E+00 7.08E+01 3.15E+02 Assessment Data Results Save to File: ANO Unit 1 10Miles Monitored Release 10072015 155849.UR17 No PAGs Exceeded Release Rates (Ci/sec) * *

• Classification: General Emergency * *

• Particulate 4.13E-02 (0.0%) Iodine 1.01E+00 (0.2%) Reviewed By: Noble Gas 5.54E+02 (99.8%)

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 29 of 46 U1 RX-9830 (SPING 3) Spent Fuel Area Exhaust - Alert Dose Assessment ANO Unit 1 Wednesday, October 07, 2015 16:05 Method: Detailed Assessment - Monitored Release Release Pathway: <SF> <Under Water> <Fuel Bldg> <FHA Filters> < Env> PRF: 4.00E-04 Containment HUT: = N/A Containment Sprays: = N/A Purge Filters: = N/A Steam Gen: = N/A Aux / FHB HUT: = < 2 Hours Filters: = Working Turb Bldg HUT: = N/A Source Term: Spent Fuel Accident - Under Water Damage: 0.560% On Site 57 m Time After S/D (hh:mm): 80:01 Wind: From 270° @ 5.8 mph Release Duration (hh:mm): 1.:00 ETE (hh:mm): [N/A] Stability Class: E Precipitation: None Monitor: FH Gas Conc Readings: 6.20E+00 µCI/cc Flowrate: 40000 CFM Distance (Miles) Exposure Rate (mR/hr) External Plume DDE (mRem) Inhalation CEDE (mRem) Deposition Ground DDE (mRem) TEDE (mRem) CDE Thyroid (mRem Evacuation Areas From 0 to 10 Miles S.B. 1.42E+01 9.98E+00 0.00E+00 0.00E+00 9.98E+00 3.89E+00 0.7 1.31E+01 9.20E+00 0.00E+00 0.00E+00 9.20E+00 3.54E+00 1.0 9.60E+00 6.72E+00 0.00E+00 0.00E+00 6.72E+00 2.54E+00 1.5 6.80E+00 4.76E+00 0.00E+00 0.00E+00 4.76E+00 1.84E+00 2.0 5.36E+00 3.75E+00 0.00E+00 0.00E+00 3.75E+00 1.46E+00 3.0 3.69E+00 2.56E+00 0.00E+00 0.00E+00 2.56E+00 1.03E+00 4.0 3.11E+00 2.10E+00 0.00E+00 0.00E+00 2.10E+00 8.19E-01 5.0 2.87E+00 1.92E+00 0.00E+00 0.00E+00 1.92E+00 7.27E-01 7.0 2.54E+00 1.69E+00 0.00E+00 0.00E+00 1.69E+00 6.03E-01 10.0 1.88E+00 1.20E+00 0.00E+00 0.00E+00 1.20E+00 3.95E-01 Assessment Data Results Save to File: ANO Unit 1 10Miles Monitored Release 10072015 160534.UR17 No PAGs Exceeded Release Rates (Ci/sec) * *

• Classification: Validate against Emergency Action Levels * *

• Particulate 9.72E-05 (0.0%) Iodine 2.31E-04 (0.0%) Reviewed By: Noble Gas 1.17E+02 (100.0%)

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 30 of 46 U1 RX-9830 (SPING 3) Spent Fuel Area Exhaust - Site Area Emergency Dose Assessment ANO Unit 1 Wednesday, October 07, 2015 16:04 Method: Detailed Assessment - Monitored Release Release Pathway: <SF> <Under Water> <Fuel Bldg> <FHA Filters> < Env> PRF: 4.00E-04 Containment HUT: = N/A Containment Sprays: = N/A Purge Filters: = N/A Steam Gen: = N/A Aux / FHB HUT: = < 2 Hours Filters: = Working Turb Bldg HUT: = N/A Source Term: Spent Fuel Accident - Under Water Damage: 0.560% On Site 57 m Time After S/D (hh:mm): 80:00 Wind: From 270° @ 5.8 mph Release Duration (hh:mm): 1.:00 ETE (hh:mm): [N/A] Stability Class: E Precipitation: None Monitor: FH Gas Conc Readings: 6.20E+01 µCI/cc Flowrate: 40000 CFM Distance (Miles) Exposure Rate (mR/hr) External Plume DDE (mRem) Inhalation CEDE (mRem) Deposition Ground DDE (mRem) TEDE (mRem) CDE Thyroid (mRem Evacuation Areas From 0 to 10 Miles S.B. 1.42E+02 9.98E+01 8.74E-01 2.18E-01 1.01E+02 3.89E+01 0.7 1.31E+02 9.20E+01 7.96E-01 1.98E-01 9.30E+01 3.54E+01 1.0 9.60E+01 6.72E+01 5.72E-01 1.42E-01 6.79E+01 2.54E+01 1.5 6.80E+01 4.76E+01 4.12E-01 1.03E-01 4.81E+01 1.84E+01 2.0 5.36E+01 3.75E+01 3.28E-01 0.00E+00 3.78E+01 1.46E+01 3.0 3.69E+01 2.56E+01 2.30E-01 0.00E+00 2.58E+01 1.03E+01 4.0 3.11E+01 2.10E+01 1.83E-01 0.00E+00 2.12E+01 8.19E+00 5.0 2.87E+01 1.92E+01 1.62E-01 0.00E+00 1.94E+01 7.27E+00 7.0 2.54E+01 1.69E+01 1.34E-01 0.00E+00 1.70E+01 6.03E+00 10.0 1.88E+01 1.20E+01 0.00E+00 0.00E+00 1.20E+01 3.95E+00 Assessment Data Results Save to File: ANO Unit 1 10Miles Monitored Release 10072015 160416.UR17 No PAGs Exceeded Release Rates (Ci/sec) * *

• Classification: Site Area Emergency * *

• Particulate 9.72E-04 (0.0%) Iodine 2.31E-03 (0.0%) Reviewed By: Noble Gas 1.17E+03 (100.0%)

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 31 of 46 U1 RX-9830 (SPING 3) Spent Fuel Area Exhaust - General Emergency Dose Assessment ANO Unit 1 Wednesday, October 07, 2015 16:03 Method: Detailed Assessment - Monitored Release Release Pathway: <SF> <Under Water> <Fuel Bldg> <FHA Filters> < Env> PRF: 4.00E-04 Containment HUT: = N/A Containment Sprays: = N/A Purge Filters: = N/A Steam Gen: = N/A Aux / FHB HUT: = < 2 Hours Filters: = Working Turb Bldg HUT: = N/A Source Term: Spent Fuel Accident - Under Water Damage: 0.560% On Site 57 m Time After S/D (hh:mm): 80:00 Wind: From 270° @ 5.8 mph Release Duration (hh:mm): 1.:00 ETE (hh:mm): [N/A] Stability Class: E Precipitation: None Monitor: FH Gas Conc Readings: 6.20E+02 µCI/cc Flowrate: 40000 CFM Distance (Miles) Exposure Rate (mR/hr) External Plume DDE (mRem) Inhalation CEDE (mRem) Deposition Ground DDE (mRem) TEDE (mRem) CDE Thyroid (mRem Evacuation Areas From 0 to 10 Miles S.B. 1.42E+03 9.98E+02 8.74E+00 2.18E+00 1.01E+03 3.89E+02 0.7 1.31E+03 9.20E+02 7.96E+00 1.98E+00 9.30E+02 3.54E+02 1.0 9.60E+02 6.72E+02 5.72E+00 1.42E+00 6.79E+02 2.54E+02 1.5 6.80E+02 4.76E+02 4.12E+00 1.03E+00 4.81E+02 1.84E+02 2.0 5.36E+02 3.75E+02 3.28E+00 8.16E-01 3.79E+02 1.46E+02 3.0 3.69E+02 2.56E+02 2.30E+00 5.06E-01 2.58E+02 1.03E+02 4.0 3.11E+02 2.10E+02 1.83E+00 3.93E-01 2.13E+02 8.19E+01 5.0 2.87E+02 1.92E+02 1.62E+00 3.44E-01 1.94E+02 7.27E+01 7.0 2.54E+02 1.69E+02 1.34E+00 2.79E-01 1.70E+02 6.03E+01 10.0 1.88E+02 1.20E+02 8.79E-01 1.75E-01 1.21E+02 3.95E+01 Assessment Data Results Save to File: ANO Unit 1 10Miles Monitored Release 10072015 160358.UR17 No PAGs Exceeded Release Rates (Ci/sec) * *

• Classification: General Emergency * *

• Particulate 9.72E-03 (0.0%) Iodine 2.31E-02 (0.0%) Reviewed By: Noble Gas 1.17E+04 (100.0%)

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 32 of 46 U1 RX-9835 (SPING 4) Penetration Room Exhaust - Alert Dose Assessment ANO Unit 1 Wednesday, October 07, 2015 16:10 Method: Detailed Assessment - Monitored Release Release Pathway: <RCS> <Containment> <Aux Bldg> <EPPR Filters> < Env> PRF: 1.20E-03 Containment HUT: = < 2 Hours Containment Sprays: = ON Purge Filters: = N/A Steam Gen: = N/A Aux / FHB HUT: = < 2 Hours Filters: = Working Turb Bldg HUT: = N/A Source Term: Reactor Core Accident - Clad On Site 57 m Time After S/D (hh:mm): 0:00 Wind: From 270° @ 5.8 mph Release Duration (hh:mm): 1.:00 ETE (hh:mm): [N/A] Stability Class: E Precipitation: None Monitor: EPPR Gas Conc Readings: 6.55E+00 µCI/cc Flowrate: 1800 CFM Distance (Miles) Exposure Rate (mR/hr) External Plume DDE (mRem) Inhalation CEDE (mRem) Deposition Ground DDE (mRem) TEDE (mRem) CDE Thyroid (mRem Evacuation Areas From 0 to 10 Miles S.B. 1.25E+01 8.35E+00 9.40E-01 7.74E-01 1.01E+01 3.16E+01 0.7 1.16E+01 7.72E+00 9.00E-01 7.83E-01 9.40E+00 2.88E+01 1.0 8.40E+00 5.44E+00 7.64E-01 7.41E-01 6.95E+00 2.06E+01 1.5 5.56E+00 3.54E+00 5.92E-01 5.69E-01 4.70E+00 1.49E+01 2.0 3.75E+00 2.39E+00 4.60E-01 3.93E-01 3.24E+00 1.18E+01 3.0 3.34E+00 2.21E+00 3.24E-01 2.80E-01 2.82E+00 8.31E+00 4.0 2.55E+00 1.64E+00 2.61E-01 2.11E-01 2.12E+00 6.62E+00 5.0 2.13E+00 1.36E+00 2.27E-01 1.69E-01 1.75E+00 5.86E+00 7.0 1.54E+00 9.87E-01 1.76E-01 1.10E-01 1.27E+00 4.84E+00 10.0 8.96E-01 5.59E-01 1.04E-01 0.00E+00 6.62E-01 3.15E+00 Assessment Data Results Save to File: ANO Unit 1 10Miles Monitored Release 10072015 161015.UR17 No PAGs Exceeded Release Rates (Ci/sec) * *

• Classification: Validate against Emergency Action Levels * *

• Particulate 4.14E-04 (0.0%) Iodine 1.02E-02 (0.2%) Reviewed By: Noble Gas 5.56E+00 (99.8%)

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 33 of 46 U1 RX-9835 (SPING 4) Penetration Room Exhaust - Site Area Emergency Dose Assessment ANO Unit 1 Wednesday, October 07, 2015 16:09 Method: Detailed Assessment - Monitored Release Release Pathway: <RCS> <Containment> <Aux Bldg> <EPPR Filters> < Env> PRF: 1.20E-03 Containment HUT: = < 2 Hours Containment Sprays: = ON Purge Filters: = N/A Steam Gen: = N/A Aux / FHB HUT: = < 2 Hours Filters: = Working Turb Bldg HUT: = N/A Source Term: Reactor Core Accident - Clad On Site 57 m Time After S/D (hh:mm): 0:00 Wind: From 270° @ 5.8 mph Release Duration (hh:mm): 1.:00 ETE (hh:mm): [N/A] Stability Class: E Precipitation: None Monitor: EPPR Gas Conc Readings: 6.55E+01 µCI/cc Flowrate: 1800 CFM Distance (Miles) Exposure Rate (mR/hr) External Plume DDE (mRem) Inhalation CEDE (mRem) Deposition Ground DDE (mRem) TEDE (mRem) CDE Thyroid (mRem Evacuation Areas From 0 to 10 Miles S.B. 1.25E+02 8.35E+01 9.40E+00 7.74E+00 1.01E+02 3.16E+02 0.7 1.16E+02 7.72E+01 9.00E+00 7.83E+00 9.40E+01 2.88E+02 1.0 8.40E+01 5.44E+01 7.64E+00 7.41E+00 6.95E+01 2.06E+02 1.5 5.56E+01 3.54E+01 5.92E+00 5.69E+00 4.70E+01 1.49E+02 2.0 3.75E+01 2.39E+01 4.60E+00 3.93E+00 3.24E+01 1.18E+02 3.0 3.34E+01 2.21E+01 3.24E+00 2.80E+00 2.82E+01 8.31E+01 4.0 2.55E+01 1.64E+01 2.61E+00 2.11E+00 2.12E+01 6.62E+01 5.0 2.13E+01 1.36E+01 2.27E+00 1.69E+00 1.75E+01 5.86E+01 7.0 1.54E+01 9.87E+00 1.76E+00 1.10E+00 1.27E+01 4.84E+01 10.0 8.96E+00 5.59E+00 1.04E+00 4.92E-01 6.62E+00 3.15E+01 Assessment Data Results Save to File: ANO Unit 1 10Miles Monitored Release 10072015 160956.UR17 No PAGs Exceeded Release Rates (Ci/sec) * *

• Classification: Site Area Emergency * *

• Particulate 4.14E-03 (0.0%) Iodine 1.02E-01 (0.2%) Reviewed By: Noble Gas 5.56E+01 (99.8%)

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 34 of 46 U1 RX-9835 (SPING 4) Penetration Room Exhaust - General Emergency Dose Assessment ANO Unit 1 Wednesday, October 07, 2015 16:09 Method: Detailed Assessment - Monitored Release Release Pathway: <RCS> <Containment> <Aux Bldg> <EPPR Filters> < Env> PRF: 1.20E-03 Containment HUT: = < 2 Hours Containment Sprays: = ON Purge Filters: = N/A Steam Gen: = N/A Aux / FHB HUT: = < 2 Hours Filters: = Working Turb Bldg HUT: = N/A Source Term: Reactor Core Accident - Clad On Site 57 m Time After S/D (hh:mm): 0:00 Wind: From 270° @ 5.8 mph Release Duration (hh:mm): 1.:00 ETE (hh:mm): [N/A] Stability Class: E Precipitation: None Monitor: EPPR Gas Conc Readings: 6.55E+02 µCI/cc Flowrate: 1800 CFM Distance (Miles) Exposure Rate (mR/hr) External Plume DDE (mRem) Inhalation CEDE (mRem) Deposition Ground DDE (mRem) TEDE (mRem) CDE Thyroid (mRem Evacuation Areas From 0 to 10 Miles S.B. 1.25E+03 8.35E+02 9.40E+01 7.74E+01 1.01E+03 3.16E+03 0.7 1.16E+02 7.72E+02 9.00E+01 7.83E+01 9.40E+02 2.88E+03 1.0 8.40E+02 5.44E+02 7.64E+01 7.41E+01 6.95E+02 2.06E+03 1.5 5.56E+02 3.54E+02 5.92E+01 5.69E+01 4.70E+02 1.49E+03 2.0 3.75E+02 2.39E+02 4.60E+01 3.93E+01 3.24E+02 1.18E+03 3.0 3.34E+02 2.21E+02 3.24E+01 2.80E+01 2.82E+02 8.31E+02 4.0 2.55E+02 1.64E+02 2.61E+01 2.11E+01 2.12E+02 6.62E+02 5.0 2.13E+02 1.36E+02 2.27E+01 1.69E+01 1.75E+02 5.86E+02 7.0 1.54E+02 9.87E+01 1.76E+01 1.10E+01 1.27E+02 4.84E+02 10.0 8.96E+01 5.59E+01 1.04E+01 4.92E+00 7.11E+01 3.15E+02 Assessment Data Results Save to File: ANO Unit 1 10Miles Monitored Release 10072015 160932.UR17 No PAGs Exceeded Release Rates (Ci/sec) * *

• Classification: General Emergency * *

• Particulate 4.14E-02 (0.0%) Iodine 1.02E+00 (0.2%) Reviewed By: Noble Gas 5.56E+02 (99.8%)

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 35 of 46 U2 2RX-9820 (SPING 5) Containment Purge Exhaust - Alert Dose Assessment ANO Unit 2 Wednesday, October 07, 2015 16:15 Method: Detailed Assessment - Monitored Release Release Pathway: <RCS> <Containment> <CP Filters> < Env> PRF: 3.00E-03 Containment HUT: = < 2 Hours Containment Sprays: = ON Purge Filters: = Working Steam Gen: = N/A Aux / FHB HUT: = N/A Filters: = N/A Turb Bldg HUT: = N/A Source Term: Reactor Core Accident - Clad On Site 57 m Time After S/D (hh:mm): 0:00 Wind: From 270° @ 5.8 mph Release Duration (hh:mm): 1.:00 ETE (hh:mm): [N/A] Stability Class: E Precipitation: None Monitor: CP Gas Conc Readings: 1.88E-01 µCI/cc Flowrate: 39700 CFM Distance (Miles) Exposure Rate (mR/hr) External Plume DDE (mRem) Inhalation CEDE (mRem) Deposition Ground DDE (mRem) TEDE (mRem) CDE Thyroid (mRem Evacuation Areas From 0 to 10 Miles S.B. 8.00E+00 5.32E+00 1.19E+00 6.77E-01 7.19E+00 5.01E+01 0.7 7.44E+00 4.92E+00 1.11E+00 6.66E-01 6.70E+00 4.56E+01 1.0 5.36E+00 3.47E+00 8.68E-01 5.90E-01 4.93E+00 3.26E+01 1.5 3.55E+00 2.26E+00 6.52E-01 4.46E-01 3.35E+00 2.35E+01 2.0 2.40E+00 1.52E+00 5.12E-01 3.15E-01 2.35E+00 1.86E+01 3.0 2.14E+00 1.41E+00 3.61E-01 2.20E-01 1.99E+00 1.32E+01 4.0 1.63E+00 1.05E+00 2.89E-01 1.67E-01 1.50E+00 1.05E+01 5.0 1.36E+00 8.64E-01 2.53E-01 1.35E-01 1.25E+00 9.29E+00 7.0 9.84E-01 6.27E-01 2.01E-01 0.00E+00 8.29E-01 7.67E+00 10.0 5.72E-01 3.56E-01 1.24E-01 0.00E+00 4.80E-01 5.00E+00 Assessment Data Results Save to File: ANO Unit 1 10Miles Monitored Release 10072015 161518.UR17 No PAGs Exceeded Release Rates (Ci/sec) * *

• Classification: Validate against Emergency Action Levels * *

• Particulate 6.56E-04 (0.0%) Iodine 1.61E-02 (0.5%) Reviewed By: Noble Gas 3.52E+00 (99.5%)

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 36 of 46 U2 2RX-9820 (SPING 5) Containment Purge Exhaust - Site Area Emergency Dose Assessment ANO Unit 2 Wednesday, October 07, 2015 16:15 Method: Detailed Assessment - Monitored Release Release Pathway: <RCS> <Containment> <CP Filters> < Env> PRF: 3.00E-03 Containment HUT: = < 2 Hours Containment Sprays: = ON Purge Filters: = Working Steam Gen: = N/A Aux / FHB HUT: = N/A Filters: = N/A Turb Bldg HUT: = N/A Source Term: Reactor Core Accident - Clad On Site 57 m Time After S/D (hh:mm): 0:00 Wind: From 270° @ 5.8 mph Release Duration (hh:mm): 1.:00 ETE (hh:mm): [N/A] Stability Class: E Precipitation: None Monitor: CP Gas Conc Readings: 1.88E+00 µCI/cc Flowrate: 39700 CFM Distance (Miles) Exposure Rate (mR/hr) External Plume DDE (mRem) Inhalation CEDE (mRem) Deposition Ground DDE (mRem) TEDE (mRem) CDE Thyroid (mRem Evacuation Areas From 0 to 10 Miles S.B. 8.00E+01 5.32E+01 1.19E+01 6.77E+00 7.19E+01 5.01E+02 0.7 7.44E+01 4.92E+01 1.11E+01 6.66E+00 6.70E+01 4.56E+02 1.0 5.36E+01 3.47E+01 8.68E+00 5.90E+00 4.93E+01 3.26E+02 1.5 3.55E+01 2.26E+01 6.52E+00 4.46E+00 3.35E+01 2.35E+02 2.0 2.40E+01 1.52E+01 5.12E+00 3.15E+00 2.35E+01 1.86E+02 3.0 2.14E+01 1.41E+01 3.61E+00 2.20E+00 1.99E+01 1.32E+02 4.0 1.63E+01 1.05E+01 2.89E+00 1.67E+00 1.50E+01 1.05E+02 5.0 1.36E+01 8.64E+00 2.53E+00 1.35E+00 1.25E+01 9.29E+01 7.0 9.84E+00 6.27E+00 2.01E+00 9.22E-01 9.21E+00 7.67E+01 10.0 5.72E+00 3.56E+00 1.24E+00 4.52E-01 5.25E+00 5.00E+01 Assessment Data Results Save to File: ANO Unit 1 10Miles Monitored Release 10072015 161502.UR17 No PAGs Exceeded Release Rates (Ci/sec) * *

• Classification: Site Area Emergency * *

• Particulate 6.56E-03 (0.0%) Iodine 1.61E-01 (0.5%) Reviewed By: Noble Gas 3.52E+01 (99.5%)

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 37 of 46 U2 2RX-9820 (SPING 5) Containment Purge Exhaust - General Emergency Dose Assessment ANO Unit 2 Wednesday, October 07, 2015 16:14 Method: Detailed Assessment - Monitored Release Release Pathway: <RCS> <Containment> <CP Filters> < Env> PRF: 3.00E-03 Containment HUT: = < 2 Hours Containment Sprays: = ON Purge Filters: = Working Steam Gen: = N/A Aux / FHB HUT: = N/A Filters: = N/A Turb Bldg HUT: = N/A Source Term: Reactor Core Accident - Clad On Site 57 m Time After S/D (hh:mm): 0:00 Wind: From 270° @ 5.8 mph Release Duration (hh:mm): 1.:00 ETE (hh:mm): [N/A] Stability Class: E Precipitation: None Monitor: CP Gas Conc Readings: 1.88E+01 µCI/cc Flowrate: 39700 CFM Distance (Miles) Exposure Rate (mR/hr) External Plume DDE (mRem) Inhalation CEDE (mRem) Deposition Ground DDE (mRem) TEDE (mRem) CDE Thyroid (mRem Evacuation Areas From 0 to 10 Miles S.B. 8.00E+02 5.32E+02 1.19E+02 6.77E+01 7.19E+02 5.01E+03 0.7 7.44E+02 4.92E+02 1.11E+02 6.66E+01 6.70E+02 4.56E+03 1.0 5.36E+02 3.47E+02 8.68E+01 5.90E+01 4.93E+02 3.26E+03 1.5 3.55E+02 2.26E+02 6.52E+01 4.46E+01 3.35E+02 2.35E+03 2.0 2.40E+02 1.52E+02 5.12E+01 3.15E+01 2.35E+02 1.86E+03 3.0 2.14E+02 1.41E+02 3.61E+01 2.20E+01 1.99E+02 1.32E+03 4.0 1.63E+02 1.05E+02 2.89E+01 1.67E+01 1.50E+02 1.05E+03 5.0 1.36E+02 8.64E+01 2.53E+01 1.35E+01 1.25E+02 9.29E+02 7.0 9.84E+01 6.27E+01 2.01E+01 9.22E+00 9.21E+01 7.67E+02 10.0 5.72E+01 3.56E+01 1.24E+01 4.52E+00 5.25E+01 5.00E+02 Assessment Data Results Save to File: ANO Unit 1 10Miles Monitored Release 10072015 161434.UR17 No PAGs Exceeded Release Rates (Ci/sec) * *

• Classification: General Emergency * *

• Particulate 6.56E-02 (0.0%) Iodine 1.61E+00 (0.5%) Reviewed By: Noble Gas 3.52E+02 (99.5%)

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 38 of 46 U2 2RX-9825 (SPING 6) Radwaste Area Exhaust - Alert Dose Assessment ANO Unit 2 Wednesday, October 07, 2015 16:18 Method: Detailed Assessment - Monitored Release Release Pathway: <RCS> <Containment> <Aux Bldg> <RWA Filters> < Env> PRF: 1.20E-03 Containment HUT: = < 2 Hours Containment Sprays: = ON Purge Filters: = N/A Steam Gen: = N/A Aux / FHB HUT: = < 2 Hours Filters: = Working Turb Bldg HUT: = N/A Source Term: Reactor Core Accident - Clad On Site 57 m Time After S/D (hh:mm): 0:00 Wind: From 270° @ 5.8 mph Release Duration (hh:mm): 1.:00 ETE (hh:mm): [N/A] Stability Class: E Precipitation: None Monitor: RWA Gas Conc Readings: 2.35E-01 µCI/cc Flowrate: 50000 CFM Distance (Miles) Exposure Rate (mR/hr) External Plume DDE (mRem) Inhalation CEDE (mRem) Deposition Ground DDE (mRem) TEDE (mRem) CDE Thyroid (mRem Evacuation Areas From 0 to 10 Miles S.B. 1.24E+01 8.31E+00 9.39E-01 7.70E-01 1.00E+01 3.15E+01 0.7 1.16E+01 7.68E+00 9.00E-01 7.78E-01 9.36E+00 2.87E+01 1.0 8.36E+00 5.40E+00 7.60E-01 7.38E-01 6.90E+00 2.06E+01 1.5 5.52E+00 3.52E+00 5.92E-01 5.66E-01 4.67E+00 1.49E+01 2.0 3.74E+00 2.38E+00 4.56E-01 3.91E-01 3.22E+00 1.18E+01 3.0 3.33E+00 2.20E+00 3.23E-01 2.78E-01 2.80E+00 8.30E+00 4.0 2.54E+00 1.64E+00 2.60E-01 2.10E-01 2.11E+00 6.61E+00 5.0 2.12E+00 1.35E+00 2.26E-01 1.68E-01 1.74E+00 5.85E+00 7.0 1.54E+00 9.82E-01 1.75E-01 1.09E-01 1.27E+00 4.83E+00 10.0 8.92E-01 5.56E-01 1.03E-01 0.00E+00 6.59E-01 3.15E+00 Assessment Data Results Save to File: ANO Unit 1 10Miles Monitored Release 10072015 161818.UR17 No PAGs Exceeded Release Rates (Ci/sec) * *

• Classification: Validate against Emergency Action Levels * *

• Particulate 4.13E-04 (0.0%) Iodine 1.01E-02 (0.2%) Reviewed By: Noble Gas 5.55E+00 (99.8%)

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 39 of 46 U2 2RX-9825 (SPING 6) Radwaste Area Exhaust - Site Area Emergency Dose Assessment ANO Unit 2 Wednesday, October 07, 2015 16:17 Method: Detailed Assessment - Monitored Release Release Pathway: <RCS> <Containment> <Aux Bldg> <RWA Filters> < Env> PRF: 1.20E-03 Containment HUT: = < 2 Hours Containment Sprays: = ON Purge Filters: = N/A Steam Gen: = N/A Aux / FHB HUT: = < 2 Hours Filters: = Working Turb Bldg HUT: = N/A Source Term: Reactor Core Accident - Clad On Site 57 m Time After S/D (hh:mm): 0:00 Wind: From 270° @ 5.8 mph Release Duration (hh:mm): 1.:00 ETE (hh:mm): [N/A] Stability Class: E Precipitation: None Monitor: RWA Gas Conc Readings: 2.35E+00 µCI/cc Flowrate: 50000 CFM Distance (Miles) Exposure Rate (mR/hr) External Plume DDE (mRem) Inhalation CEDE (mRem) Deposition Ground DDE (mRem) TEDE (mRem) CDE Thyroid (mRem Evacuation Areas From 0 to 10 Miles S.B. 1.24E+02 8.31E+01 9.39E+00 7.70E+00 1.00E+02 3.15E+02 0.7 1.16E+02 7.68E+01 9.00E+00 7.78E+00 9.36E+01 2.87E+02 1.0 8.36E+01 5.40E+01 7.60E+00 7.38E+00 6.90E+01 2.06E+02 1.5 5.52E+01 3.52E+01 5.92E+00 5.66E+00 4.67E+01 1.49E+02 2.0 3.74E+01 2.38E+01 4.56E+00 3.91E+00 3.22E+01 1.18E+02 3.0 3.33E+01 2.20E+01 3.23E+00 2.78E+00 2.80E+01 8.30E+01 4.0 2.54E+01 1.64E+01 2.60E+00 2.10E+00 2.11E+01 6.61E+01 5.0 2.12E+01 1.35E+01 2.26E+00 1.68E+00 1.74E+01 5.85E+01 7.0 1.54E+01 9.82E+00 1.75E+00 1.09E+00 1.27E+01 4.83E+01 10.0 8.92E+00 5.56E+00 1.03E+00 4.90E-01 7.08E+00 3.15E+01 Assessment Data Results Save to File: ANO Unit 1 10Miles Monitored Release 10072015 161756.UR17 No PAGs Exceeded Release Rates (Ci/sec) * *

• Classification: Site Area Emergency * *

• Particulate 4.13E-03 (0.0%) Iodine 1.01E-01 (0.2%) Reviewed By: Noble Gas 5.55E+01 (99.8%)

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 40 of 46 U2 2RX-9825 (SPING 6) Radwaste Area Exhaust - General Emergency Dose Assessment ANO Unit 2 Wednesday, October 07, 2015 16:17 Method: Detailed Assessment - Monitored Release Release Pathway: <RCS> <Containment> <Aux Bldg> <RWA Filters> < Env> PRF: 1.20E-03 Containment HUT: = < 2 Hours Containment Sprays: = ON Purge Filters: = N/A Steam Gen: = N/A Aux / FHB HUT: = < 2 Hours Filters: = Working Turb Bldg HUT: = N/A Source Term: Reactor Core Accident - Clad On Site 57 m Time After S/D (hh:mm): 0:00 Wind: From 270° @ 5.8 mph Release Duration (hh:mm): 1.:00 ETE (hh:mm): [N/A] Stability Class: E Precipitation: None Monitor: RWA Gas Conc Readings: 2.35E+01 µCI/cc Flowrate: 50000 CFM Distance (Miles) Exposure Rate (mR/hr) External Plume DDE (mRem) Inhalation CEDE (mRem) Deposition Ground DDE (mRem) TEDE (mRem) CDE Thyroid (mRem Evacuation Areas From 0 to 10 Miles S.B. 1.24E+03 8.31E+02 9.39E+01 7.70E+01 1.00E+03 3.15E+03 0.7 1.16E+03 7.68E+02 9.00E+01 7.78E+01 9.36E+02 2.87E+03 1.0 8.36E+02 5.40E+02 7.60E+01 7.38E+01 6.90E+02 2.06E+03 1.5 5.52E+02 3.52E+02 5.92E+01 5.66E+01 4.67E+02 1.49E+03 2.0 3.74E+02 2.38E+02 4.56E+01 3.91E+01 3.22E+02 1.18E+03 3.0 3.33E+02 2.20E+02 3.23E+01 2.78E+01 2.80E+02 8.30E+02 4.0 2.54E+02 1.64E+02 2.60E+01 2.10E+01 2.11E+02 6.61E+02 5.0 2.12E+02 1.35E+02 2.26E+01 1.68E+01 1.74E+02 5.85E+02 7.0 1.54E+02 9.82E+01 1.75E+01 1.09E+01 1.27E+02 4.83E+02 10.0 8.92E+01 5.56E+01 1.03E+01 4.90E+00 7.08E+01 3.15E+02 Assessment Data Results Save to File: ANO Unit 1 10Miles Monitored Release 10072015 161733.UR17 No PAGs Exceeded Release Rates (Ci/sec) * *

• Classification: General Emergency * *

• Particulate 4.13E-02 (0.0%) Iodine 1.01E+00 (0.2%) Reviewed By: Noble Gas 5.55E+02 (99.8%)

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 41 of 46 U2 2RX-9830 (SPING 7) Spent Fuel Area Exhaust - Alert Dose Assessment ANO Unit 2 Wednesday, October 07, 2015 16:28 Method: Detailed Assessment - Monitored Release Release Pathway: <SF> <Under Water> <Fuel Bldg> <FHA Filters> < Env> PRF: 4.00E-04 Containment HUT: = N/A Containment Sprays: = N/A Purge Filters: = N/A Steam Gen: = N/A Aux / FHB HUT: = < 2 Hours Filters: = Working Turb Bldg HUT: = N/A Source Term: Spent Fuel Accident - Under Water Damage: 0.560% On Site 57 m Time After S/D (hh:mm): 80:02 Wind: From 270° @ 5.8 mph Release Duration (hh:mm): 1.:00 ETE (hh:mm): [N/A] Stability Class: E Precipitation: None Monitor: FH Gas Conc Readings: 6.86E+00 µCI/cc Flowrate: 36000 CFM Distance (Miles) Exposure Rate (mR/hr) External Plume DDE (mRem) Inhalation CEDE (mRem) Deposition Ground DDE (mRem) TEDE (mRem) CDE Thyroid (mRem Evacuation Areas From 0 to 10 Miles S.B. 1.42E+01 9.98E+00 0.00E+00 0.00E+00 9.98E+00 3.89E+00 0.7 1.31E+01 9.20E+00 0.00E+00 0.00E+00 9.20E+00 3.54E+00 1.0 9.60E+00 6.72E+00 0.00E+00 0.00E+00 6.72E+00 2.54E+00 1.5 6.80E+00 4.76E+00 0.00E+00 0.00E+00 4.76E+00 1.84E+00 2.0 5.36E+00 3.75E+00 0.00E+00 0.00E+00 3.75E+00 1.46E+00 3.0 3.69E+00 2.56E+00 0.00E+00 0.00E+00 2.56E+00 1.03E+00 4.0 3.11E+00 2.10E+00 0.00E+00 0.00E+00 2.10E+00 8.19E-01 5.0 2.87E+00 1.92E+00 0.00E+00 0.00E+00 1.92E+00 7.27E-01 7.0 2.54E+00 1.69E+00 0.00E+00 0.00E+00 1.69E+00 6.03E-01 10.0 1.88E+00 1.20E+00 0.00E+00 0.00E+00 1.20E+00 3.95E-01 Assessment Data Results Save to File: ANO Unit 1 10Miles Monitored Release 10072015 162831.UR17 No PAGs Exceeded Release Rates (Ci/sec) * *

• Classification: Validate against Emergency Action Levels * *

• Particulate 9.68E-05 (0.0%) Iodine 2.30E-04 (0.0%) Reviewed By: Noble Gas 1.17E+02 (100.0%)

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 42 of 46 U2 2RX-9830 (SPING 7) Spent Fuel Area Exhaust - Site Area Emergency Dose Assessment ANO Unit 2 Wednesday, October 07, 2015 16:28 Method: Detailed Assessment - Monitored Release Release Pathway: <SF> <Under Water> <Fuel Bldg> <FHA Filters> < Env> PRF: 4.00E-04 Containment HUT: = N/A Containment Sprays: = N/A Purge Filters: = N/A Steam Gen: = N/A Aux / FHB HUT: = < 2 Hours Filters: = Working Turb Bldg HUT: = N/A Source Term: Spent Fuel Accident - Under Water Damage: 0.560% On Site 57 m Time After S/D (hh:mm): 80:02 Wind: From 270° @ 5.8 mph Release Duration (hh:mm): 1.:00 ETE (hh:mm): [N/A] Stability Class: E Precipitation: None Monitor: FH Gas Conc Readings: 6.86E+01 µCI/cc Flowrate: 36000 CFM Distance (Miles) Exposure Rate (mR/hr) External Plume DDE (mRem) Inhalation CEDE (mRem) Deposition Ground DDE (mRem) TEDE (mRem) CDE Thyroid (mRem Evacuation Areas From 0 to 10 Miles S.B. 1.42E+02 9.98E+01 8.73E-01 2.18E-01 1.01E+02 3.89E+01 0.7 1.31E+02 9.20E+01 7.96E-01 1.98E-01 9.30E+01 3.54E+01 1.0 9.60E+01 6.72E+01 5.68E-01 1.42E-01 6.79E+01 2.54E+01 1.5 6.80E+01 4.76E+01 4.12E-01 1.03E-01 4.81E+01 1.84E+01 2.0 5.36E+01 3.75E+01 3.27E-01 0.00E+00 3.78E+01 1.46E+01 3.0 3.69E+01 2.56E+01 2.29E-01 0.00E+00 2.58E+01 1.03E+01 4.0 3.11E+01 2.10E+01 1.83E-01 0.00E+00 2.12E+01 8.19E+00 5.0 2.87E+01 1.92E+01 1.62E-01 0.00E+00 1.94E+01 7.26E+00 7.0 2.54E+01 1.69E+01 1.34E-01 0.00E+00 1.70E+01 6.03E+00 10.0 1.88E+01 1.20E+01 0.00E+00 0.00E+00 1.20E+01 3.95E+00 Assessment Data Results Save to File: ANO Unit 1 10Miles Monitored Release 10072015 162808.UR17 No PAGs Exceeded Release Rates (Ci/sec) * *

• Classification: Site Area Emergency * *

• Particulate 9.68E-04 (0.0%) Iodine 2.30E-03 (0.0%) Reviewed By: Noble Gas 1.17E+03 (100.0%)

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 43 of 46 U2 2RX-9830 (SPING 7) Spent Fuel Area Exhaust - General Emergency Dose Assessment ANO Unit 2 Wednesday, October 07, 2015 16:27 Method: Detailed Assessment - Monitored Release Release Pathway: <SF> <Under Water> <Fuel Bldg> <FHA Filters> < Env> PRF: 4.00E-04 Containment HUT: = N/A Containment Sprays: = N/A Purge Filters: = N/A Steam Gen: = N/A Aux / FHB HUT: = < 2 Hours Filters: = Working Turb Bldg HUT: = N/A Source Term: Spent Fuel Accident - Under Water Damage: 0.560% On Site 57 m Time After S/D (hh:mm): 80:02 Wind: From 270° @ 5.8 mph Release Duration (hh:mm): 1.:00 ETE (hh:mm): [N/A] Stability Class: E Precipitation: None Monitor: FH Gas Conc Readings: 6.86E+02 µCI/cc Flowrate: 36000 CFM Distance (Miles) Exposure Rate (mR/hr) External Plume DDE (mRem) Inhalation CEDE (mRem) Deposition Ground DDE (mRem) TEDE (mRem) CDE Thyroid (mRem Evacuation Areas From 0 to 10 Miles S.B. 1.42E+03 9.98E+02 8.73E+00 2.18E+00 1.01E+03 3.89E+02 0.7 1.31E+03 9.20E+02 7.96E+00 1.98E+00 9.30E+02 3.54E+02 1.0 9.60E+02 6.72E+02 5.68E+00 1.42E+00 6.79E+02 2.54E+02 1.5 6.80E+02 4.76E+02 4.12E+00 1.03E+00 4.81E+02 1.84E+02 2.0 5.36E+02 3.75E+02 3.27E+00 8.14E-01 3.79E+02 1.46E+02 3.0 3.69E+02 2.56E+02 2.29E+00 5.05E-01 2.58E+02 1.03E+02 4.0 3.11E+02 2.10E+02 1.83E+00 3.93E-01 2.13E+02 8.19E+01 5.0 2.87E+02 1.92E+02 1.62E+00 3.43E-01 1.94E+02 7.26E+01 7.0 2.54E+02 1.69E+02 1.34E+00 2.78E-01 1.70E+02 6.03E+01 10.0 1.88E+02 1.20E+02 8.76E-01 1.75E-01 1.21E+02 3.95E+01 Assessment Data Results Save to File: ANO Unit 1 10Miles Monitored Release 10072015 162750.UR17 No PAGs Exceeded Release Rates (Ci/sec) * *

• Classification: General Emergency * *

• Particulate 9.68E-03 (0.0%) Iodine 2.30E-02 (0.0%) Reviewed By: Noble Gas 1.17E+04 (100.0%)

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 44 of 46 U2 2RX-9835 (SPING 8) Penetration Room Exhaust - Alert Dose Assessment ANO Unit 2 Wednesday, October 07, 2015 16:31 Method: Detailed Assessment - Monitored Release Release Pathway: <RCS> <Containment> <Aux Bldg> <EPPR Filters> < Env> PRF: 1.20E-03 Containment HUT: = < 2 Hours Containment Sprays: = ON Purge Filters: = N/A Steam Gen: = N/A Aux / FHB HUT: = < 2 Hours Filters: = Working Turb Bldg HUT: = N/A Source Term: Reactor Core Accident - Clad On Site 57 m Time After S/D (hh:mm): 0:00 Wind: From 270° @ 5.8 mph Release Duration (hh:mm): 1.:00 ETE (hh:mm): [N/A] Stability Class: E Precipitation: None Monitor: EPPR Gas Conc Readings: 5.88E+00 µCI/cc Flowrate: 2000 CFM Distance (Miles) Exposure Rate (mR/hr) External Plume DDE (mRem) Inhalation CEDE (mRem) Deposition Ground DDE (mRem) TEDE (mRem) CDE Thyroid (mRem Evacuation Areas From 0 to 10 Miles S.B. 1.25E+01 8.32E+00 9.40E-01 7.74E-01 1.01E+01 3.16E+01 0.7 1.16E+01 7.68E+00 9.00E-01 7.83E-01 9.36E+00 2.88E+01 1.0 8.40E+00 5.44E+00 7.60E-01 7.41E-01 6.94E+00 2.06E+01 1.5 5.56E+00 3.53E+00 5.92E-01 5.69E-01 4.69E+00 1.49E+01 2.0 3.74E+00 2.38E+00 4.56E-01 3.93E-01 3.23E+00 1.18E+01 3.0 3.34E+00 2.21E+00 3.24E-01 2.80E-01 2.81E+00 8.31E+00 4.0 2.55E+00 1.64E+00 2.61E-01 2.11E-01 2.11E+00 6.62E+00 5.0 2.12E+00 1.35E+00 2.27E-01 1.69E-01 1.75E+00 5.86E+00 7.0 1.54E+00 9.84E-01 1.75E-01 1.09E-01 1.27E+00 4.84E+00 10.0 8.92E-01 5.57E-01 1.03E-01 0.00E+00 6.60E-01 3.15E+00 Assessment Data Results Save to File: ANO Unit 1 10Miles Monitored Release 10072015 163140.UR17 No PAGs Exceeded Release Rates (Ci/sec) * *

• Classification: Validate against Emergency Action Levels * *

• Particulate 4.13E-04 (0.0%) Iodine 1.01E-02 (0.2%) Reviewed By: Noble Gas 5.55E+00 (99.8%)

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 45 of 46 U2 2RX-9835 (SPING 8) Penetration Room Exhaust - Site Area Emergency Dose Assessment ANO Unit 2 Wednesday, October 07, 2015 16:31 Method: Detailed Assessment - Monitored Release Release Pathway: <RCS> <Containment> <Aux Bldg> <EPPR Filters> < Env> PRF: 1.20E-03 Containment HUT: = < 2 Hours Containment Sprays: = ON Purge Filters: = N/A Steam Gen: = N/A Aux / FHB HUT: = < 2 Hours Filters: = Working Turb Bldg HUT: = N/A Source Term: Reactor Core Accident - Clad On Site 57 m Time After S/D (hh:mm): 0:00 Wind: From 270° @ 5.8 mph Release Duration (hh:mm): 1.:00 ETE (hh:mm): [N/A] Stability Class: E Precipitation: None Monitor: EPPR Gas Conc Readings: 5.88E+01 µCI/cc Flowrate: 2000 CFM Distance (Miles) Exposure Rate (mR/hr) External Plume DDE (mRem) Inhalation CEDE (mRem) Deposition Ground DDE (mRem) TEDE (mRem) CDE Thyroid (mRem Evacuation Areas From 0 to 10 Miles S.B. 1.25E+02 8.32E+01 9.40E+00 7.74E+00 1.01E+02 3.16E+02 0.7 1.16E+02 7.68E+01 9.00E+00 7.83E+00 9.36E+01 2.88E+02 1.0 8.40E+01 5.44E+01 7.60E+00 7.41E+00 6.94E+01 2.06E+02 1.5 5.56E+01 3.53E+01 5.92E+00 5.69E+00 4.69E+01 1.49E+02 2.0 3.74E+01 2.38E+01 4.56E+00 3.93E+00 3.23E+01 1.18E+02 3.0 3.34E+01 2.21E+01 3.24E+00 2.80E+00 2.81E+01 8.31E+01 4.0 2.55E+01 1.64E+01 2.61E+00 2.11E+00 2.11E+01 6.62E+01 5.0 2.12E+01 1.35E+01 2.27E+00 1.69E+00 1.75E+01 5.86E+01 7.0 1.54E+01 9.84E+00 1.75E+00 1.09E+00 1.27E+01 4.84E+01 10.0 8.92E+00 5.57E+00 1.03E+00 4.92E-01 7.09E+00 3.15E+01 Assessment Data Results Save to File: ANO Unit 1 10Miles Monitored Release 10072015 163125.UR17 No PAGs Exceeded Release Rates (Ci/sec) * *

• Classification: Site Area Emergency * *

• Particulate 4.13E-03 (0.0%) Iodine 1.01E-01 (0.2%) Reviewed By: Noble Gas 5.55E+01 (99.8%)

Radiological Effluent EAL Values EP-CALC-ANO-1701, Rev. 0 Page 46 of 46 U2 2RX-9835 (SPING 8) Penetration Room Exhaust - General Emergency Dose Assessment ANO Unit 2 Wednesday, October 07, 2015 16:31 Method: Detailed Assessment - Monitored Release Release Pathway: <RCS> <Containment> <Aux Bldg> <EPPR Filters> < Env> PRF: 1.20E-03 Containment HUT: = < 2 Hours Containment Sprays: = ON Purge Filters: = N/A Steam Gen: = N/A Aux / FHB HUT: = < 2 Hours Filters: = Working Turb Bldg HUT: = N/A Source Term: Reactor Core Accident - Clad On Site 57 m Time After S/D (hh:mm): 0:00 Wind: From 270° @ 5.8 mph Release Duration (hh:mm): 1.:00 ETE (hh:mm): [N/A] Stability Class: E Precipitation: None Monitor: EPPR Gas Conc Readings: 5.88E+02 µCI/cc Flowrate: 2000 CFM Distance (Miles) Exposure Rate (mR/hr) External Plume DDE (mRem) Inhalation CEDE (mRem) Deposition Ground DDE (mRem) TEDE (mRem) CDE Thyroid (mRem Evacuation Areas From 0 to 10 Miles S.B. 1.25E+03 8.32E+02 9.40E+01 7.74E+01 1.00E+03 3.16E+03 0.7 1.16E+02 7.68E+02 9.00E+01 7.83E+01 9.36E+02 2.88E+03 1.0 8.40E+02 5.44E+02 7.60E+01 7.41E+01 6.94E+02 2.06E+03 1.5 5.56E+02 3.53E+02 5.92E+01 5.69E+01 4.69E+02 1.49E+03 2.0 3.74E+02 2.38E+02 4.56E+01 3.93E+01 3.23E+02 1.18E+03 3.0 3.34E+02 2.21E+02 3.24E+01 2.80E+01 2.81E+02 8.31E+02 4.0 2.55E+02 1.64E+02 2.61E+01 2.11E+01 2.11E+02 6.62E+02 5.0 2.12E+02 1.35E+02 2.27E+01 1.69E+01 1.75E+02 5.86E+02 7.0 1.54E+02 9.84E+01 1.75E+01 1.09E+01 1.27E+02 4.84E+02 10.0 8.92E+01 5.57E+01 1.03E+01 4.92E+00 7.09E+01 3.15E+02 Assessment Data Results Save to File: ANO Unit 1 10Miles Monitored Release 10072015 163106.UR17 No PAGs Exceeded Release Rates (Ci/sec) * *

• Classification: General Emergency * *

• Particulate 4.13E-02 (0.0%) Iodine 1.01E+00 (0.2%) Reviewed By: Noble Gas 5.55E+02 (99.8%)

Containment High Range Radiation Monitor EAL Values EP-CALC-ANO-1702, Rev. 0 Page 1 of 15 Table of Contents Section Page 1. PURPOSE ............................................................................................................................2 2. DEVELOPMENT METHODOLOGY AND BASES ................................................................2 2.1 Fuel Clad Loss .............................................................................................................2 2.2 Reactor Coolant System Loss .....................................................................................4 2.3 Containment Potential Loss .........................................................................................4 2.4 Source Term ................................................................................................................6 2.5 Decay Considerations ..................................................................................................6

3. DESIGN INPUTS ..................................................................................................................6 3.1 Constants and Conversion Factors .............................................................................6 3.2 Plant Inputs ..................................................................................................................6 3.3 Source Term ................................................................................................................7

4. CALCULATIONS ..................................................................................................................8 4.1 Fuel Clad Damage Estimate Based on 300 µCi/gm DEI-131 ......................................8 4.2 Fission Product Barrier Thresholds .............................................................................9 5. CONCLUSIONS .................................................................................................................10

6. REFERENCES ...................................................................................................................11 ATTACHMENTS Attachment 1, 300 µCi/gm DEI-131 Equivalent Fuel Clad Damage ......................................12 , Fission Product Barrier Threshold Values ......................................................14 , NUREG-1940 Figure 1-1 PWR Containment Monitor Response ...................15

Containment High Range Radiation Monitor EAL Values EP-CALC-ANO-1702, Rev. 0 Page 2 of 15 1. PURPOSE The Arkansas Nuclear One (ANO) Emergency Action Level (EAL) Technical Bases Manual contains background information, event declaration thresholds, bases and references for the EAL and Fission Product Barrier (FPB) values used to implement the Nuclear Energy Institute (NEI) 99-01 Revision 6 EAL guidance. This calculation document provides additional technical detail specific to the derivation of the FPB containment high range radiation monitor (CHRRM) readings developed in accordance with the guidance in NEI 99-01 Revision 6.

Documentation of the assumptions, calculations and results are provided for the values associated the NEI 99-01 Revision 6 Table 9-F-3, PWR EAL Fission Product Barrier Table, thresholds listed below.

NEI Fuel Clad Loss 3.A NEI Reactor Coolant Loss 3.A NEI Containment Potential Loss 3.A

2. DEVELOPMENT METHODOLOGY AND BASES

2.1 Fuel Clad Loss Guidance Criteria Per NEI 99-01 Revision 6, this radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals 300 µCi/gm dose equivalent I-131 (DEI-131). Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the fuel clad barrier. The radiation monitor reading in this threshold is higher than that specified for RCS barrier loss threshold since it indicates a loss of both the fuel clad barrier and the RCS barrier. The reading should be determined assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory, with RCS radioactivity concentration equal to 300 µCi/gm dose equivalent I-131, into the primary containment atmosphere. ANO Bases The fuel clad FPB threshold value is based on an instantaneous release of reactor coolant into the containment at a percent fuel clad damage equivalent to 300 µCi/gm DEI-131 RCS activity. That percent fuel clad damage value is ratioed to a containment radiation reading for 100% fuel clad damage to determine the fuel clad FPB threshold value in R/hr.

Containment High Range Radiation Monitor EAL Values EP-CALC-ANO-1702, Rev. 0 Page 3 of 15 2.2 Reactor Coolant System Loss Guidance Criteria Per NEI 99-01 Revision 6, this radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for fuel clad barrier loss threshold since it indicates a loss of the RCS barrier only. The reading should be determined assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory, with RCS activity at Technical Specification allowable limits, into the primary containment atmosphere. RCS activity at this level will typically result in primary containment radiation levels that can be more readily detected by primary containment radiation monitors, and more readily differentiated from those caused by piping or component "shine" sources. If desired, a plant may use a lesser value of RCS activity for determining this value. In some cases, the site-specific physical location and sensitivity of the containment radiation monitor(s) may be such that radiation from a cloud of released RCS gases cannot be distinguished from radiation emanating from piping and components containing elevated reactor coolant activity. If so, determine if an alternate indication is available. ANO Bases The ANO technical specification high value for DEI-131 is 60 µCi/gm. This activity would yield a containment radiation monitor reading approximately 5x lower than the fuel clad loss fission product barrier containment radiation reading equivalent to 300 µCi/gm. NUREG-1940 Figure 1-1 provides estimates for standard plant containment radiation based on spiked RCS activity, which is slightly less than half the value obtained by the 300 µCi/gm to 60 Ci/gm DEI-131 ratio. NUREG-1940 Figure 1-1 models a spiked RCS activity that is lower than the RCS activity equivalent to 60 µCi/gm DEI-131 described above (the NUREG-1940 graph is based on a release into containment of 100 times the non-noble gas fission products normally found in the coolant). This is the preferred value for the RCS loss threshold as it provides for a containment monitor escalation of approximately one decade between fission product barrier thresholds at the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> point. NEI 99-01 guidance criteria allows the use of a lesser value for RCS activity (see guidance criteria section above). The ANO RCS FPB threshold value is based on NUREG-1940 standard plant containment radiation readings for an instantaneous release of spiked reactor coolant, which is lower than 60 µCi/gm DEI-131 Technical Specification allowable limits, and is adjusted for the site specific power rating.

Containment High Range Radiation Monitor EAL Values EP-CALC-ANO-1702, Rev. 0 Page 4 of 15 2.3 Containment Potential Loss Guidance Criteria Per NEI 99-01 Revision 6, this radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous fuel clad and RCS barrier loss thresholds. NUREG-1228 indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS and the fuel clad barriers. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the classification level to a General Emergency. NUREG-1228 provides the basis for using the 20% fuel cladding failure value. Unless there is a site-specific analysis justifying a different value, the reading should be determined assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with 20% fuel clad failure into the primary containment atmosphere. ANO Bases The Containment FPB threshold value is based on an instantaneous release of all reactor coolant into the containment at an equivalent of 20% clad damage. The ANO FPB containment radiation reading value equivalent to 20% fuel clad damage is obtained by ratio of the 100% fuel clad damage containment radiation reading value to 20% fuel clad damage.

2.4 Source Term Guidance Criteria NEI 99-01 does not specify a basis for the source term activity or the reduction factors. RG 1.183 provides assumptions for a LOCA used as a reference for FSAR design basis event analysis. Per RG 1.183 Section 1.1.4, Emergency Preparedness Applications: Requirements for emergency preparedness at nuclear power plants are set forth in 10 CFR 50.47, "Emergency Plans." Additional requirements are set forth in Appendix E, "Emergency Planning and Preparedness for Production and Utilization Facilities," to 10 CFR Part 50. The planning basis for many of these requirements was published in NUREG-0396, "Planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support of Light Water Nuclear Power Plants". This joint effort by the Environmental Protection Agency (EPA) and the NRC considered the principal characteristics (such as nuclides released and distances) likely to be involved for a spectrum of design basis and severe (core melt) accidents. No single accident scenario is the basis of the required preparedness. The Containment High Range Radiation Monitor EAL Values EP-CALC-ANO-1702, Rev. 0 Page 5 of 15 objective of the planning is to provide public protection that would encompass a wide spectrum of possible events with a sufficient basis for extension of response efforts for unanticipated events. These requirements were issued after a long period of involvement by numerous stakeholders, including the Federal Emergency Management Agency, other Federal agencies, local and State governments (and in some cases, foreign governments), private citizens, utilities, and industry groups. Although the AST provided in this guide was based on a limited spectrum of severe accidents, the particular characteristics have been tailored specifically for DBA analysis use. The AST is not representative of the wide spectrum of possible events that make up the planning basis of emergency preparedness. Therefore, the AST is insufficient by itself as a basis for requesting relief from the emergency preparedness requirements of 10 CFR 50.47 and Appendix E to 10 CFR Part 50. Thus, RG 1.183 is not used as a basis for the containment radiation monitor thresholds. Guidance contained in NUREG-1940 is considered representative of the wide spectrum of possible events that make up the emergency preparedness planning basis and provides radiological consequence assessment methods which are acceptable to the NRC. Additionally, the source term used to develop the effluent EAL thresholds and in the Unified RASCAL Interface/Radiological Assessment System for Consequence Analysis (URI/RASCAL) dose assessment model is from NUREG-1940. Thus, NUREG-1940 has been selected as a source term basis for the fission product barrier containment radiation thresholds for conformance to NRC guidance and consistency with other source term bases used within the Entergy emergency preparedness program. ANO Bases 2.4.1 The NUREG-1940 source term inputs used for the fission product barrier containment radiation thresholds are as follows: Fuel Clad Damage Equivalent to 300 µCi/g DEI-131 - NUREG-1940 Table 1-1 equilibrium core activity, in conjunction with the NUREG-1940 Table 1-5 non-noble gas release fraction, is used to develop the site specific iodine source term. Fuel Clad and Containment Barrier Thresholds - NUREG-1940 Figure 1-1 for cladding failure is used as a basis to establish these thresholds. RCS Barrier Threshold - NUREG-1940 Figure 1-1 for spiked coolant is used as a basis to establish this threshold. Note - Source term reduction from containment spray is not included as an assumption for these thresholds. 2.4.2 NUREG-1940 source term is based on a generic plant with a power rating of 3000 MWt. The ANO site specific source term is derived from the licensed core thermal power output of 2,568 megawatts for Unit 1 and 3,026 megawatts for Unit 2 (SAR Section 1.1).

Containment High Range Radiation Monitor EAL Values EP-CALC-ANO-1702, Rev. 0 Page 6 of 15 2.4.3 Dose equivalent iodine 131 (DEI-131) dose conversion factors (DCFs) are developed from EPA-400-R 001 isotopic DCFs. EPA-400 is the basis for the protective action guidelines and is the appropriate source for DCFs used in emergency preparedness. The DEI-131 dose conversion factors are not based on the FSAR Chapter 15 or other 10 CFR 20 reference sources as those are not reflective of the exposure assumptions used within the EPA guidance for emergency preparedness use.

2.5 Decay Considerations Guidance Criteria Fission product barrier thresholds and their associated EALs are applicable only when the plant is in Hot Shutdown, Startup, or Power Operation modes (known as the hot operating modes). The events for these thresholds correspond to an instantaneous release of all reactor coolant mass into the primary containment. ANO Bases Consistent with the NUREG-1940 graphs, the instantaneous release of the RCS to the containment is assumed to occur one hour after the damage event / reactor scram to account for damage progression, dispersion of activity and decay of the very short half-life isotopes. 3. DESIGN INPUTS 3.1 Constants and Conversion Factors None 3.2 Plant Inputs 3.2.1 Rated Power 1) Standard Plant (NUREG-1940 Section 1.2.4) ........................................... 3,000 MWt 2) Unit 1 (SAR Section 1.1) ........................................................................... 2,568 MWt 3) Unit 2 (SAR Section 1.1) ........................................................................... 3,026 MWt 3.2.2 RCS Water Mass at STP (1302.022 Attachment 3 Section 2.3) 1) Unit 1 ...................................................................................................... 2.41E+8 gm 2) Unit 2 ...................................................................................................... 2.14E+8 gm 3.2.3 Standard Plant Containment Radiation Reading (NUREG-1940 Figure 1-1) 1) 100% fuel clad damage (spray off) ................... 60,000 R/hr (@ 1 hr after shutdown) 2) 100% spiked coolant (spray off) .............................. 50 R/hr (@ 1 hr after shutdown)

Containment High Range Radiation Monitor EAL Values EP-CALC-ANO-1702, Rev. 0 Page 7 of 15 3.2.4 Monitor Range 1) RE-8060/8061 (TM GO63.0010/TD G063 0020) ............... 1.00E+0 to 1.00E+8 R/hr 2) 2RE-8925-1/2RE-8925-2 (TM GO63.0010/TD G063 0020) 1.00E+0 to 1.00E+8 R/hr 3.3 Source Term 3.3.1 Source Term Activity (NUREG-1940 Table 1-1)

Core Activity (Ci/MWt) 1-131 2.67E+04 1-132 3.88E+04 1-133 5.42E+04 1-134 5.98E+04 1-135 5.18E+04 3.3.2 Monitor Range 1) Non-Noble Gasses (I, Cs, Rb) - Fuel Clad Damage .................................. 0.05 (5%) 3.3.3 Release Fractions - RFCore (NUREG-1940 Table 1-5) Rem/hr per µCi/cc1-131 1.3E+06 1-132 7.7E+03 1-133 2.2E+05 1-134 1.3E+03 1-135 3.8E+04

Containment High Range Radiation Monitor EAL Values EP-CALC-ANO-1702, Rev. 0 Page 8 of 15 4. CALCULATIONS 4.1 Fuel Clad Damage Estimate Based on 300 µCi/gm DEI-131 4.1.1 Equivalent Iodine Core Activity 100% Core Activityi(Ci) = Core Activityi(Ci/MWt) x Unit MWt Core Activity (Ci) Unit 1 Unit 2 1-131 6.86E+07 8.08E+07 1-132 9.96E+07 1.17E+08 1-133 1.39E+08 1.64E+08 1-134 1.54E+08 1.81E+08 1-135 1.33E+08 1.57E+08 Total 5.94E+08 7.00E+08 4.1.2 100% Core Activity Equivalent Reactor Coolant Iodine Concentrations 100% Core RCS Activityi(µCi/gm) = 100% Core RCS Activityi(Ci) x 106 RCS Mass (gm) RCS Activity (Ci/gm) Unit 1 Unit 2 1-131 2.85E+05 3.78E+05 1-132 4.13E+05 5.49E+05 1-133 5.78E+05 7.66E+05 1-134 6.37E+05 8.46E+05 1-135 5.52E+05 7.32E+05 Total 2.46E+06 3.27E+06 4.1.3 100% Core Activity Equivalent Reactor Coolant Iodine Concentrations 100% Clad Damage RCS Activityi(µCi/gm) = 100% Core RCS Activityi(µCi/gm) x RFCore RCS Activity (Ci/gm) Unit 1 Unit 2 1-131 1.42E+04 1.89E+04 1-132 2.07E+04 2.74E+04 1-133 2.89E+04 3.83E+04 1-134 3.19E+04 4.23E+04 1-135 2.76E+04 3.66E+04 Total 1.23E+05 1.64E+05 Containment High Range Radiation Monitor EAL Values EP-CALC-ANO-1702, Rev. 0 Page 9 of 15 4.1.4 100% Fuel Clad Damage Activity Equivalent Reactor Coolant DEI-131 Concentrations 100% DEI RCS Activityi(µCi/gm) = = 100% Clad Damage RCS Activityi(µCi/gm) x DEI DCFi Note - The DEI DCF value for each iodine isotope is determined as follows: DEI DCFi = EPA - 400 Table 5 - 2 Iodine DCFi(Rem/hr per µCi/cc) EPA - 400 Table 5 - 2 Iodine DCFI-131(Rem/hr per µCi/cc) DEI DCF (unit less) RCS Activity (Ci/gm) Unit 1 Unit 2 1-131 1.00E+00 1.42E+04 1.89E+04 1-132 5.92E-03 1.22E+02 1.62E+02 1-133 1.69E-01 4.89E+03 6.48E+03 1-134 1.00E-03 3.19E+01 4.23E+01 1-135 2.92E-02 8.07E+02 1.07E+03 Total 2.01E+04 2.66E+04 4.1.5 % Fuel Clad Damage Activity Equivalent Reactor Coolant at 300 µCi/gm DEI-131

% Clad Damage = 300 µCi/gm 100% DEI RCS Activity(µCi/gm) U1 300 Ci/gm DEI-131 ..................................................... 1.49% Fuel Clad Damage U2 300 Ci/gm DEI-131 ..................................................... 1.13% Fuel Clad Damage See Attachment 1 for the spreadsheet calculations that develop the fuel clad source term activity and the % clad damage.

4.2 Fission Product Barrier Thresholds See Attachment 2 for the spreadsheet calculations that develop the FPB threshold monitor readings.

4.2.1 Containment Potential Loss (20% Fuel Clad Damage Monitor Reading)

Unit20% clad(R/hr) = Std Plant100% clad(R/hr) x 20% x MWtUnit MWtStd Plant U1 Containment Potential Loss Threshold Monitor Reading ........... 1.03E+4 R/hr U2 Containment Potential Loss Threshold Monitor Reading ........... 1.21E+4 R/hr Containment High Range Radiation Monitor EAL Values EP-CALC-ANO-1702, Rev. 0 Page 10 of 15 4.2.2 Fuel Clad Loss (300 µCi/gm DEI-131 Equivalent Clad Damage Monitor Reading) Unit1.49% clad(R/hr) = Std Plant100% clad(R/hr) x 1.49% x MWtU1 MWtStd Plant Unit1.13% clad(R/hr) = Std Plant100% clad(R/hr) x 1.13% x MWtU2 MWtStd Plant U1 Fuel Clad Loss Threshold Monitor Reading ................................. 7.68E+2 R/hr U2 Fuel Clad Loss Threshold Monitor Reading ................................. 6.82E+2 R/hr 4.2.3 RCS Loss (Spiked Coolant Monitor Reading)

UnitSpiked(R/hr) = Std Plant100% Spiked(R/hr) x MWtUnit MWtStd Plant U1 RCS Loss Threshold Monitor Reading ......................................... 4.28E+1 R/hr U2 RCS Loss Threshold Monitor Reading ......................................... 5.04E+1 R/hr

5. CONCLUSIONS 5.1 300 µCi/gm DEI-131 is equivalent to: 1) Unit 1 .............................................................................. 1.49% fuel clad (gap) damage 2) Unit 2 .............................................................................. 1.13% fuel clad (gap) damage 5.2 Calculated containment high range radiation monitor values are as follows: Fuel Clad Loss RCS Loss Containment Potential Loss Unit 1 7.68E+2 R/hr 4.28E+1 R/hr 1.03E+4 R/hr Unit 2 6.82E+2 R/hr 5.04E+1 R/hr 1.21E+4 R/hr Based on monitor accuracy/readability and human factors, the EAL Fission Product Barrier thresholds are established as follows: Fuel Clad Loss RCS Loss Containment Potential Loss Unit 1 750 R/hr 40 R/hr 10,000 R/hr Unit 2 700 R/hr 50 R/hr 12,000 R/hr

Containment High Range Radiation Monitor EAL Values EP-CALC-ANO-1702, Rev. 0 Page 11 of 15 6. REFERENCES 6.1 NEI 99-01 R6, Development of Emergency Action Levels for Non-Passive Reactors, September 2012 6.2 EPA-400-R 001, Manual of Protective action Guides and Protective Actions for Nuclear Incidents, May 1992 6.3 NUREG-1940, RASCAL 4: Description of Models and Methods, December 2012 6.4 NUREG-1228, Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents, October 1988 6.5 Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000 6.6 ANO1 Safety Analysis Report (SAR) 1) Section 1.1, Introduction, Amendment 20 6.7 ANO2 Safety Analysis Report (SAR) 1) Section 1.1, Introduction, Amendment 17 6.8 ANO1 Technical Specifications 1) Section 3.4.12, RCS Specific Activity, Amendment 243 6.9 ANO2 Technical Specifications 1) Section 3.4.8, RCS Specific Activity, Amendment 293 6.10 1302.022, Core Damage Assessment, Change 5 6.11 TM GO63.00, General Atomic Tech Manual, Revision 2

6.12 TD G063 0020, General Atomic Tech Manual, Revision 2 Containment High Range Radiation Monitor EAL Values EP-CALC-ANO-1702, Rev. 0 Page 12 of 15 ATTACHMENT 1 300 µCi/gm DEI-131 Equivalent Clad Damage Unit 1 NUREG-1940 Table 1-1 Core Activity (Ci/MWt) U1 Core Activity (Ci) U1 RCS Activity (µCi/gm 100% Core) U1 RCS Activity (µCi/gm 100% Clad) EPA-400 Table 5-2 Dose Conversion Factors (Rem/hr per µCi/cc) DEI DCF U1 RCS Activity (µCi/gm 100% Gap DEI) I-131 2.67E+04 6.86E+07 2.85E+05 1.42E+04 1.30E+06 1.00E+00 1.42E+04 I-132 3.88E+04 9.96E+07 4.13E+05 2.07E+04 7.70E+03 5.92E-03 1.22E+02 I-133 5.42E+04 1.39E+08 5.78E+05 2.89E+04 2.20E+05 1.69E-01 4.89E+03 I-134 5.98E+04 1.54E+08 6.37E+05 3.19E+04 1.30E+03 1.00E-03 3.19E+01 I-135 5.18E+04 1.33E+08 5.52E+05 2.76E+04 3.80E+04 2.92E-02 8.07E+02 Total 2.31E+04 5.94E+08 2.46E+06 1.23E+05 2.01E+04 U1 Rate Power (MWt): 2568 RCS Liquid Volume (gm): 2.41E+08 Halogen Release Fraction: 5.0% Target DEI: 3.00E+02 %Clad Damage: 1.49%

Containment High Range Radiation Monitor EAL Values EP-CALC-ANO-1702, Rev. 0 Page 13 of 15 Unit 2 NUREG-1940 Table 1-1 Core Activity (Ci/MWt) U2 Core Activity (Ci) U2 RCS Activity (µCi/gm 100% Core) U2 RCS Activity (µCi/gm 100% Clad) EPA-400 Table 5-2 Dose Conversion Factors (Rem/hr per µCi/cc) DEI DCF U2 RCS Activity (µCi/gm 100% Gap DEI) I-131 2.67E+04 8.08E+07 3.78E+05 1.89E+04 1.30E+06 1.00E+00 1.89E+04 I-132 3.88E+04 1.17E+08 5.49E+05 2.74E+04 7.70E+03 5.92E-03 1.62E+02 I-133 5.42E+04 1.64E+08 7.66E+05 3.83E+04 2.20E+05 1.69E-01 6.48E+03 I-134 5.98E+04 1.81E+08 8.46E+05 4.23E+04 1.30E+03 1.00E-03 4.23E+01 I-135 5.18E+04 1.57E+08 7.32E+05 3.66E+04 3.80E+04 2.92E-02 1.07E+03 Total 2.31E+05 7.00E+08 3.27E+06 1.64E+05 2.66E+04 U2Rate Power (MWt): 3026 RCS Liquid Volume (gm): 2.14E+08 Halogen Release Fraction: 5.0% Target DEI: 3.00E+02 %Clad Damage: 1.13%

Containment High Range Radiation Monitor EAL Values EP-CALC-ANO-1702, Rev. 0 Page 14 of 15 ATTACHMENT 2 Fission Product Barrier Threshold Values Rated Power (MWt) Reading for 100% Clad Failure (R/hr) Reading for 20% Clad Failure (R/hr) % Damage for 300 µCi/gm RCS Activity Reading for 300 µCi/gm RCS Activity (R/hr) Reading for Spiked RCS Activity (R/hr) Unit 1 2568 5.14E+04 1.03E+04 1.49% 7.68E+02 4.28E+01 Unit 2 3026 6.05E+04 1.21E+04 1.13% 6.82E+02 5.04E+01 NUREG-1940 100% Clad Failure (R/hr): 6.00E+04 NUREG-1940 100% Spiked Coolant (R/hr): 5.00E+01 Standard Plant (MWt): 3000

Containment High Range Radiation Monitor EAL Values EP-CALC-ANO-1702, Rev. 0 Page 15 of 15 ATTACHMENT 3 NUREG-1940 Figure 1-1 PWR Containment Monitor Response