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Response to Request for Additional Information Regarding License Amendment Request (TSCR-166), Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors. (Pa
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Issue date:	07/26/2018
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I ATTACHMENT 2 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELATING TO LICENSE AMENDMENT REQUEST TSCR-166 UPDATED CLEAN COPY OF THE PROPOSED DAEC EAL SCHEME 125 pages follow Duane Arnold Energy Center (DAEC) Emergency Action Levels Technical Bases Document TBD,2018 TABLE OF CONTENTS 1 BASIS FOR EMERGENCY ACTION LEVELS .................................................................

1 1.1 OPERATING REACTORS ..................................................................................................

1 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSl) .....................................

2 1.3 NRC ORDEREA-12-051

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3 2 KEY TERMINOLOGY USED IN DAEC EAL SCHEME .....................................................

4 2.1 EMERGENCY CLASSIFICATION LEVEL (ECL) ..............

.' ................................................

4 2.2 INITIATING CONDITION (IC) ..........................................................................................

6 2.3 EMERGENCY ACTION LEVEL (EAL) .............................................................................

6 2.4 FISSION PRODUCT BARRIER THRESHOLD

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6 3 DESIGN OF THE DAEC EMERGENCY CLASSIFICATION SCHEME .............................

7 3.1 ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS (ECLS) ...............................

7 3.2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS ....................

10 3.3 DAEC-SPECIFIC ORGANIZATION AND PRESENTATION OF GENERIC INFORMATIONll 3.4 IC AND EAL MODE APPLICABILITY

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12 4 DAEC SCHEME DEVELOPMENT

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13 4.1 GENERAL DEVELOPMENT PROCESS ............................................................................

13 4.2 CRITICAL CHARACTERISTICS

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13 4.3 INSTRUMENTATION USED FOR EALS ..........................................................................

14 4.4 EAL/THRESHOLD REFERENCES TO AOP AND EOP SETPOINTS/CRITERIA

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14 5 GUIDANCE ON USING THE DAEC EALS ....******.......******....

ll ****************************************

15 5.1 GENERAL CONSIDERATIONS

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15 5.2 CLASSIFICATION METHODOLOGY

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16 5.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS

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16 5 .4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION

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17 5.5 CLASSIFICATION OF IMMINENT CONDITIONS

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17 5.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING

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17 5.7 CLASSIFICATION OF SHORT-LIVED EVENTS ...............................................................

18 5.8 CLASSIFICATION OF TRANSIENT CONDITIONS

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18 5 .9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION

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19 5.10 RETRACTION OF AN EMERGENCY DECLARATION

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19 11 6 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS ........................

20 7 COLD SHUTDOWN/

REFUELING SYSTEM MALFUNCTION ICS/EALS ....................

36 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS ..............

58 9 FISSION PRODUCT BARRIER ICS/EALS ******************************************************************

60 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS .........

75 11 SYSTEM MALFUNCTION ICS/EALS ***************************************************************************

96 APPENDIX A -ACRONYMS AND ABBREVIATIONS

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A-1 APPENDIX B -DEFINITIONS

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DUANE ARNOLD EMERGENCY ACTION LEVELS TECHNICAL BASIS DOCUMENT 1 BASIS FOR EMERGENCY ACTION LEVELS 1.1 OPERATING REACTORS Title 10, Code of Federal Regulations (CFR), Energy, contains the U.S. Nuclear Regulatory Commission (NRC) regulations that apply to nuclear power facilities.

Several of these regulations govern various aspects of an emergency classification scheme. A review of the relevant sections listed below will aid the reader in understanding the key terminology provided in Section 3.0 of this document.

10 CFR § 50.47(a)(l)(i)

10 CFR § 50.47(b)(4)

10 CFR § 50.54(q)

10 CFR § 50.72(a)

10 CFR § 50, Appendix E, IV.B, Assessment Actions

10 CFR § 50, Appendix E, IV.C, Activation of Emergency Organization The above regulations are supplemented by various regulatory guidance documents.

Three documents of particular relevance to NEI 99-01 are: NUREG-0654/FEMA-REP-1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, October 1980. [Refer to Appendix 1, Emergency Action Level Guidelines for Nuclear Power Plants] NUREG-1022, Event Reporting Guidelines 10 CFR § 50. 72 and§ 50. 73 Regulatory Guide 1.101, Emergency Response Planning and Preparedness for Nuclear Power Reactors 1 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) Selected guidance in NEI 99-01 is applicable to licensees electing to use their 10 CFR 50 emergency plan to fulfill the requirements of 10 CFR 72.32 for a stand-alone ISFSI. The emergency classification levels applicable to an ISFSI are consistent with the requirements of 10 CFR 50 and the guidance in NUREG 0654/FEMA-REP-1.

The initiating conditions germane to a 10 CFR 72.32 emergency plan (as described in NUREG-1567) are subsumed within the classification scheme for a 10 CFR 50.47 emergency plan. The generic ICs and EALs for an ISFSI are presented in Section 8, ISFSI ICs/EALs.

IC E-HUl covers the spectrum of credible natural and man-made events included within the scope of an ISFSI design. This IC is not applicable to installations or facilities that may process and/or repackage spent fuel (e.g., a Monitored Retrievable Storage Facility or an ISFSI at a spent fuel processing facility).

In addition, appropriate aspects of IC HUI and IC HAI should also be included to address a HOSTILE ACTION directed against an ISFSI. The analysis of potential onsite and off site consequences of accidental releases associated with the operation of an ISFSI is contained in NUREG-1140, A Regulatory Analysis on Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees.

NUREG-1140 concluded that the postulated worst-case accident involving an ISFSI has insignificant consequences to public health and safety. This evaluation shows that the maximum offsite dose to a member of the public due to an accidental release of radioactive materials would not exceed 1 rem Effective Dose Equivalent.

Regarding the above information, the expectations for an offsite response to an Alert classified under a 10 CFR 72.32 emergency plan are generally consistent with those for a Notification of Unusual Event in a 10 CFR 50.47 emergency plan (e.g., to provide assistance ifrequested).

Also, the licensee's Emergency Response Organization (ERO) required for 10 CFR 72.32 emergency plan is different than that prescribed for a 10 CFR 50.47 emergency plan (e.g., no emergency technical support function).

2 1.3 NRC ORDER EA-12-051 The Fukushima Daiichi accident of March 11, 2012, was the result of a tsunami that exceeded the plant's design basis and flooded the site's emergency electrical power supplies and distribution systems. This caused an extended loss of power that severely compromised the key safety functions of core cooling and containment integrity, and ultimately led to core damage in three reactors.

While the loss of power also impaired the spent fuel pool cooling function, sufficient water inventory was maintained in the pools to preclude fuel damage from the loss of cooling. Following a review of the Fukushima Daiichi accident, the NRC concluded that several measures were necessary to ensure adequate protection of public health and safety under the provisions of the backfit rule, 10 CFR 50.109(a)(4)(ii).

Among them was to provide each spent fuel pool with reliable level instrumentation to significantly enhance the ability of key decision-makers to allocate resources effectively following a beyond design basis event. To this end, the NRC issued Order EA-12-051, Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, on March 12, 2012, to all US nuclear plants with an operating license, construction permit, or combined construction and operating license. NRC Order EA-12-051 states, in part, "All licensees

... shall have a reliable indication of the water level in associated spent fuel storage pools capable of supporting identification of the following pool water level conditions by trained personnel:

(1) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck, and (3) level where fuel remains covered and actions to implement make-up water addition should no longer be deferred." To this end, all licensees must provide:

A primary and back-up level instrument that will monitor water level from the normal level to the top of the used fuel rack in the pool;

A display in an area accessible following a severe event; and

Independent electrical power to each instrument channel and provide an alternate remote power connection capability.

NEI 12-02, Industry Guidance for Compliance with NRC Order EA-12-051, "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, "provides guidance for complying with NRC Order EA-12-051.

NEI 99-01, Revision 6, includes three EALs that reflect the availability of the enhanced spent fuel pool level instrumentation associated with NRC Order EA-12-051.

These EALs are included within ICs RA2, RS2, and RG2. 3 2 KEY TERMINOLOGY USED IN DAEC EAL SCHEME There are several key terms that appear throughout the EAL methodology.

These terms are introduced in this section to support understanding of subsequent material.

As an aid to the reader, the following table is provided as an overview to illustrate the relationship of the terms to each other. Emergency Classification Level Unusual Event I Alert I SAE I GE Initiating Condition Initiating Condition Initiating Condition Initiating Condition Emergency Action Emergency Action Emergency Action Emergency Action Level (1) Level (1) Level (1) Level (1)

Operating Mode

Operating Mode

Operating Mode

Operating Mode Applicability Applicability Applicability Applicability

Notes

Notes

Notes

Notes

Basis

Basis

Basis

Basis (1) -When making an emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of an Initiating Condition.

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 referred to as Fission Product Barrier Thresholds; the thresholds serve the same function as an EAL. 2.1 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) 2.1.1 Notification of Unusual Event (NOUE) 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. Purpose: The purpose of this classification is to assure that the first step in future response has been carried out, to bring the operations staff to a state of readiness, and to provide systematic handling of unusual event information and decision-making.

4

2.1.2 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 PAG exposure levels. Purpose: The purpose of this classification is to assure that emergency personnel are readily available to respond if the situation becomes more serious or to perform confirmatory radiation monitoring if required, and provide off site authorities current information on plant status and parameters.

2.1.3 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.

Purpose: The purpose of the Site Area Emergency declaration is to assure that emergency response centers are staffed, to assure that monitoring teams are dispatched, to assure that personnel required for evacuation of near-site areas are at duty stations if the situation becomes more serious, to provide consultation with offsite authorities, and to provide updates to the public through government authorities.

2.1.4 General Emergency (GE) 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 P AG exposure levels off site for more than the immediate site area. Purpose: The purpose of the General Emergency declaration is to initiate predetennined protective actions for the public, to provide continuous assessment of information from the licensee and offsite organizational measurements, to initiate additional measures as indicated by actual or potential releases, to provide consultation with offsite authorities, and to provide updates for the public through government authorities.

5 2.2 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.

Discussion:

An IC describes an event or condition, the severity or consequences of which meets the definition of an emergency classification level. An IC can be expressed as a continuous, measurable parameter (e.g., RCS leakage), an event (e.g., an earthquake) or the status of one or more fission product barriers (e.g., loss of the RCS barrier).

Appendix 1 of NUREG-0654 does not contain example Emergency Action Levels (EALs) for each ECL, but rather Initiating Conditions (i.e., plant conditions that indicate that a radiological emergency, or events that could lead to a radiological emergency, has occurred).

NUREG-0654 states that the Initiating Conditions form the basis for establishment by a licensee of the specific plant instrumentation readings (as applicable) which, if exceeded, would initiate the emergency classification.

Thus, it is the specific instrument readings that would be the EALs. 2.3 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. Discussion:

EAL statements may utilize a variety of criteria including instrument readings and status indications; observable events; results of calculations and analyses; entry into particular procedures; and the occurrence of natural phenomena.

2.4 FISSION PRODUCT BARRIER THRESHOLD A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier. Discussion:

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.

The primary fission product barriers are: Fuel Clad Reactor Coolant System (RCS) Containment Upon determination that one or more fission product barrier thresholds have been exceeded, the combination of barrier loss and/or potential loss thresholds is compared to the fission product barrier IC/EAL criteria to determine the appropriate ECL. In some accident sequences, the ICs and EALs presented in the Abnormal Radiation Levels/ Radiological Effluent (R) Recognition Category will be exceeded at the same time, or shortly after, the loss of one or more fission product barriers.

This redundancy is intentional as the former ICs address radioactivity releases that result in certain offsite doses from whatever cause, including events that might not be fully encompassed by fission product barriers (e.g., spent fuel pool accidents, design containment leakage following a LOCA, etc.). 6 L 3 DESIGN OF THE DAEC EMERGENCY CLASSIFICATION SCHEME 3.1 ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS (ECLS) An effective emergency classification scheme must incorporate a realistic and accurate assessment of risk, both to plant workers and the public. There are obvious health and safety risks in underestimating the potential or actual threat from an event or condition; however, there are also risks in overestimating the threat as well (e.g., harm that may occur during an evacuation).

The DAEC emergency classification scheme attempts to strike an appropriate balance between reasonably anticipated event or condition consequences, potential accident trajectories, and risk avoidance or minimization.

There are a range of "non-emergency events" reported to the US Nuclear Regulatory Commission (NRC) staff in accordance with the requirements of 10 CFR 50.72. Guidance concerning these rep01iing requirements, and example events, are provided in NUREG-1022.

Certain events reportable under the provisions of 10 CFR 50.72 may also require the declaration of an emergency.

In order to align each Initiating Conditions (IC) with the appropriate ECL, it was necessary to determine the attributes of each ECL. The goal of this process is to answer the question, "What events or conditions should be placed under each ECL ?" The following sources provided information and context for the development ofECL attributes.

Assessments of the effects and consequences of different types of events and conditions DAEC abnormal and emergency operating procedure setpoints and transition criteria DAEC Technical Specification limits and controls Offsite Dose Assessment Manual (ODAM) radiological release limits Review of selected Updated Final Safety Analysis Report (UFSAR) accident analyses Environmental Protection Agency (EPA) Protective Action Guidelines (PAGs) NUREG 0654, Appendix 1, Emergency Action Level Guidelines for Nuclear Power Plants Industry Operating Experience Input from DAEC subject matter experts The following ECL attributes are used to aid in the development of ICs and Emergency Action Levels (EALs ). The attributes may be useful in briefing and training settings (e.g., helping an Emergency Director understand why a particular condition is classified as an Alert). 7 3.1.1 Notification of Unusual Event (NODE) A Notification of Unusual Event, as defined in section 2.1.1, includes but is not limited to an event or condition that involves: (A) A precursor to a more significant event or condition. (B) A minor loss of control of radioactive materials or the ability to control radiation levels within the plant. (C) A consequence otherwise significant enough to warrant notification to local, State and Federal authorities.

3.1.2 Alert An Alert, as defined in section 2.1.2, includes but is not limited to an event or condition that involves: (A)A loss or potential loss of either the fuel clad or Reactor Coolant System (RCS) fission product barrier. (B) An event or condition that significantly reduces the margin to a loss or potential loss of the fuel clad or RCS fission product barrier. (C) A significant loss of control of radioactive materials resulting in an inability to control radiation levels within the plant, or a release of radioactive materials to the environment that could result in doses greater than 1 % of an EPA PAG at or beyond the site boundary. (D)A HOSTILE ACTION occurring within the OWNER CONTROLLED AREA, including those directed at an Independent Spent Fuel Storage Installation (ISFSI). 3.1.3 Site Area Emergency (SAE) A Site Area Emergency, as defined in section 2.1.3, includes but is not limited to an event or condition that involves: (A)A loss or potential loss of any two fission product barriers -fuel clad, RCS and/or containment. (B) A precursor event or condition that may lead to the loss or potential loss of multiple fission product barriers within a relatively short period of time. Precursor events and conditions of this type include those that challenge the monitoring and/or control of multiple SAFETY SYSTEMS. (C) A release of radioactive materials to the environment that could result in doses greater than 10% of an EPA PA G at or beyond the site boundary. (D)A HOSTILE ACTION occurring within the plant PROTECTED AREA. 8 3.1.4 General Emergency (GE) A General Emergency, as defined in section 2.1.4, includes but is not limited to an event or condition that involves: (A) Loss of any two fission product barriers AND loss or potential loss of the third barrier -fuel clad, RCS and/or containment. (B) A precursor event or condition that, unmitigated, may lead to a loss of all three fission product barriers.

Precursor events and conditions of this type include those that lead directly to core damage and loss of containment integrity. ( C) A release of radioactive materials to the environment that could result in doses greater than an EPA PA G at or beyond the site boundary. (D)A HOSTILE ACTION resulting in the loss of key safety functions (reactivity control, core cooling/RPV water level or RCS heat removal) or damage to spent fuel. 3.1.5 Risk-Informed Insights Emergency preparedness is a defense-in-depth measure that is independent of the assessed risk from any particular accident sequence; however, the development of an effective emergency classification scheme can benefit from a review ofrisk-based assessment results. To that end, the development and assignment of certain ICs and EALs also considered insights from several site-specific probabilistic safety assessments.

Some generic insights from this review included:

1. Accident sequences involving a prolonged loss of all AC power are significant contributors to core damage frequency at many Boiling Water Reactors (BWRs). For this reason, a loss of all AC power for greater than 15 minutes, with the plant at or above Hot Shutdown, was assigned an ECL of Site Area Emergency.

Precursor events to a loss of all AC power were also included as an Unusual Event and an Alert. A station blackout coping analyses performed in response to 10 CFR 50.63 and Regulatory Guide 1.155, Station Blackout, may be used to determine a time-based criterion to demarcate between a Site Area Emergency and a General Emergency.

The time dimension is critical to a properly anticipatory emergency declaration since the goal is to maximize the time available for State and local officials to develop and implement offsite protective actions. 2. For severe core damage events, uncertainties exist in phenomena important to accident progressions leading to containment failure. Because of these uncertainties, predicting the status of containment integrity may be difficult under severe accident conditions.

This is why maintaining containment integrity alone following sequences leading to severe core damage is an insufficient basis for not escalating to a General Emergency.

3. PSAs indicated that leading contributors to latent fatalities were sequences involving a containment bypass, a large Loss of Coolant Accident (LOCA) with early containment failure, a Station Blackout lasting longer than the DAEC coping period, and a reactor coolant pump seal failure. The generic EAL methodology needs to be sufficiently rigorous to address these sequences in a timely fashion. 9 3.2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS The NEI 99-01 methodology makes use of symptom-based, barrier-based and based ICs and EALs. Each type is discussed below. Symptom-based ICs and EALs are parameters or conditions that are measurable over some range using plant instrumentation (e.g., core temperature, reactor coolant level, radiological effluent, etc.). When one or more of these parameters or conditions are normal, reactor operators will implement procedures to identify the probable cause(s) and take corrective action. Fission product barrier-based ICs and EALs are the subset of symptom-based EALs that refer specifically to the level of challenge to the principal barriers against the release of radioactive material from the reactor core to the environment.

These barriers are the fuel cladding, the reactor coolant system pressure boundary, and the containment.

The barrier-based ICs and EALs consider the level of challenge to each individual barrier -potentially lost and lost -and the total number of barriers under challenge.

Event-based I Cs and EALs define a variety of specific occurrences that have potential or actual safety significance.

These include the failure of an automatic reactor scram to shut down the reactor, natural phenomena (e.g., an earthquake), or man-made hazards such as a toxic gas release. 10 3.3 DAEC-SPECIFIC ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION The scheme's generic information is organized by Recognition Category in the following order. R -Abnormal Radiation Levels / Radiological Effluent -Section 6 C -Cold Shutdown I Refueling System Malfunction

-Section 7 E -Independent Spent Fuel Storage Installation (ISFSI) -Section 8 F -Fission Product Barrier -Section 9 H -Hazards and Other Conditions Affecting Plant Safety -Section 10 S -System Malfunction

-Section 11 Each Recognition Category section contains a matrix showing the ICs and their associated emergency classification levels. The following information and guidance is provided for each IC: ECL -the assigned emergency classification level for the IC. Initiating Condition

-provides a summary description of the emergency event or condition.

Operating Mode Applicability

-Lists the modes during which the IC and associated EAL(s) are applicable (i.e., are to be used to classify events or conditions).

Emergency Action Level(s)-Provides examples of reports and indications that are considered to meet the intent of the IC. For Recognition Category F, the fission product barrier thresholds are presented in tables and arranged by fission product barrier and the degree of barrier challenge (i.e., potential loss or loss). This presentation method shows the synergism among the thresholds, and supports accurate assessments.

Basis -Provides background information that explains the intent and application of the IC and EALs. In some cases, the basis also includes relevant source information and references.

11 3 .4 IC AND EAL MODE APPLICABILITY The DAEC emergency classification scheme was developed recognizing that the applicability of ICs and EALs will vary with plant mode. For example, some based ICs and EALs can be assessed only during the power operations, startup, or hot standby/shutdown modes of operation when all fission product barriers are in place, and plant instrumentation and SAFETY SYSTEMS are fully operational.

In the cold shutdown and refueling modes, different symptom-based ICs and EALs will come into play to reflect the opening of systems for routine maintenance, the unavailability of some SAFETY SYSTEM components and the use of alternate instrumentation.

The following table shows which Recognition Categories are applicable in each plant mode. The ICs and EALs for a given Recognition Category are applicable in the indicated modes. MODE APPLICABILITY MATRIX Recognition Category Mode R C E F H s Power Operations X X X X X Startup X X X X X Hot Shutdown X X X X X Cold Shutdown X X X X Refueling X X X X Defueled X X X X DAEC Operating Modes Power Operations (1): Startup (2): Hot Shutdown (3): Cold Shutdown ( 4): Refueling (5): Mode Switch in Run Mode Switch in Startup/Hot Standby or Refuel (with all vessel head closure bolts fully tensioned)

Mode Switch in Shutdown, Average Reactor Coolant Temperature

>212 °F (with all vessel head closure bolts fully tensioned)

Mode Switch in Shutdown, Average Reactor Coolant Temperature~

212 °F (with all vessel head closure bolts fully tensioned)

Mode Switch in Shutdown or Refuel (with one or more vessel head closure bolts less than fully tensioned) 12 4 DEVELOPMENT OF THE DAEC EMERGENCY CLASSIFICATION SCHEME 4.1 GENERAL DEVELOPMENT PROCESS The DAEC ICs and EALs were developed to be unambiguous and readily assessable.

The IC is the fundamental event or condition requiring a declaration.

The EAL(s) is the pre-determined threshold that defines when the IC is met. Useful acronyms and abbreviations associated with the DAEC emergency classification scheme are presented in Appendix A, Acronyms and Abbreviations.

Many words or terms used in the DAEC emergency classification scheme have specific definitions.

These words and terms are identified by being set in all capital letters (i.e., ALL CAPS). The definitions are presented in Appendix B, Definitions.

4.2 CRITICAL CHARACTERISTICS When crafting the scheme, DAEC ensured that certain critical characteristics have been met. These critical characteristics are listed below.

The ICs, EALs, Operating Mode Applicability criteria, Notes and Basis information are consistent with industry guidance; while the actual wording may be different, the classification intent is maintained.

With respect to Recognition Category F, DAEC includes a user-aid to facilitllte timely and accurate classification of fission product baiTier losses and/or potential losses. The user-aid logic is consistent with the classification logic presented in Section 9.

The I Cs, EALs, Operating Mode Applicability criteria, Notes and Basis information are technically complete and accurate (i.e., they contain the information necessary to make a correct classification).

EAL statements use objective criteria and observable values.

I Cs, EALs, Operating Mode Applicability and Note statements and formatting consider human factors and are user-friendly.

The scheme facilitates upgrading and downgrading of the emergency classification where necessary.

The scheme facilitates classification of multiple concurrent events or conditions.

13 4.3 INSTRUMENTATIONUSEDFOREALS DAEC incorporated instrumentation that is reliable and routinely maintained in accordance with site programs and procedures.

Alarms referenced in EAL statements are those that are the most operationally significant for the described event or condition.

EAL setpoints are within the calibrated range of the referenced instrumentation, and consider any automatic instrumentation functions that may impact accurate EAL assessment.

In addition, EAL setpoint values do not use terms such as "off-scale low" or "off-scale high" since that type of reading may not be readily differentiated from an instrument failure. 4.4 EAL/THRESHOLD REFERENCES TO AOP AND EOP SETPOINTS/CRITERIA Some of the criteria/values used in several EALs and fission product barrier thresholds are drawn from DAEC AOPs and EOPs. This approach is intended to maintain good alignment between operational diagnoses and emergency classification assessments.

Appropriate administrative controls are in place to ensure that a subsequent change to an AOP or EOP is screened to determine if an evaluation pursuant to 10 CFR 50.54( q) is required.

14 5 GUIDANCE ON USING THE DAEC EALS 5.1 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 referred to as Fission Product Barrier Thresholds; the thresholds serve the same function as an EAL. 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/or Nuclear Power Plants. All emergency classification assessments should 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, validation could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel.

The validation of indications should be completed in a manner that supports timely emergency declaration.

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.

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. The assessment of some EALs is based on the results of analyses that are necessary to asce11ain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.); the EAL and/or the associated basis discussion will identify the necessary analysis.

In these cases, the 15-minute declaration 15 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).

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.

This 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 into the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier. 5.2 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(s) must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, then the IC is considered met and the associated ECL is declared in accordance with plant procedures.

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.

5.3 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.

Additionally, there is no "additive" effect from multiple EALs meeting the same ECL. For example: If two Alert EALs are met, 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. 16 5.4 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. 5.5 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.

5.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING An ECL may be downgraded when the event or condition that meets the highest IC and EAL no longer exists, and other site-specific downgrading requirements are met. If downgrading the ECL is deemed appropriate, the new ECL would then be based on a lower applicable IC(s) and EAL(s). The ECL may also simply be terminated.

17 The following approach to downgrading or tenninating an ECL is recommended.

ECL Unusual Event Alert Site Area Emergency with no long-term plant damage Site Area Emergency with long-term plant damage General Emergency Action When Condition No Longer Exists Terminate the emergency in accordance with plant procedures.

Downgrade or terminate the emergency in accordance with plant procedures.

Terminate the emergency and enter recovery in accordance with plant procedures.

As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02. 5.7 CLASSIFICATION OF SHORT-LIVED EVENTS As discussed in Section 3 .2, event-based I Cs 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 a failure of the reactor protection system to automatically scram the reactor followed by a successful manual scram or an earthquake.

5.8 CLASSIFICATION OF TRANSIENT CONDITIONS Many of the I Cs and/or EALs contained in this document 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 where 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 perfonned in accordance with procedures.

18 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 auxiliary feedwater system fails to automatically start. Steam generator levels rapidly decrease and the plant enters an inadequate RCS heat removal condition (a potential loss of both the fuel clad and RCS barriers).

If an operator manually starts the auxiliary feedwater system in accordance with an EOP step and clears the inadequate RCS heat removal 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 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 where 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.

5.9 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 is applicable.

Specifically, the event should be reported to the NRC in accordance with 10 CFR 50.72 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.

5 .10 RETRACTION OF AN EMERGENCY DECLARATION Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022.

19 6 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS 20 RU1 ECL: Notification of Unusual Event Initiating Condition:

Release of gaseous or liquid rad i oactivity greater than 2 times the ODAM limits for 60 minutes or longer. Operating Mode Applicability:

All Emergency Action Levels: Notes:

The Emergency Director shou ld declare the event promptly upon determining that the applicable time ha s been exceeded, or w ill lik e l y be exceeded.

If an ongo in g r e l ease is detected an d the r e l ease start time i s unknown , assume that the release duration has exceeded the specified time limit.

If the effluent flow past an effl u ent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no l onger va lid for classification purposes.

RUI .1 Reading on ANY Tab l e R-1 effluent radiation monitor greater than column " NOUE" for 60 minutes or lon ger: RUl.2 RUl.3 Table R-1 -Effluent Monitor Classification Thresholds

V) ::l 0 (l) V) ro l!J Monitor Rea cto r Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8) Turbine Building ventilation rad monit or (Kaman 1/2) Offga s Stack rad monitor (Kaman 9/10) LLRPSF rad monitor (Kam an 12) GSW r a d monitor (RIS-4767) RHRSW & ESW rad monitor (RM-199 7) RHRSW & ESW Rupture Di sc rad monitor (RM-4268) NOUE 8.0E-04 uci/cc 8.0E-04 uci/cc 2.0E-01 uci/cc l.2E-03 uci/cc 1.SE+03 cps 8.4E+02 cps l.OE+03 cps Reading on ANY effluent.radiatio n monitor g reater than 2 times the alarm setpoint established by a current radioactivity dischar ge permit for 60 minutes or l onger. Sample ana l ysis for a gaseous or liquid release indi cates a conce ntra tion or release rate greater than 2 times the ODAM limits for 60 minute s or lon ger. 21 Definitions:

None Basis: -----------

Th i s IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radio lo gical 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 monitored, including those for which a radioactivity discharge permit is normally prepared.

DAEC incorporates design features intended to control the release of radioactive effluents to the environment.

Further, there are administrat i ve controls established to prevent unintentional releases , and to control and monitor intentional releases. The occurrence of an extended , uncontrolled radioactive re l ease to the env ironm ent i s indicative of degradation in these features and/or controls.

Radiological effluent EALs are also included to provide a basi s for classifyin g events and cond iti ons that cannot be readi l y or appropr i ate l y class ifi ed 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 readin gs assumes that a release path to the environment is estab li shed. If the effluent flow past an effluent monitor i s known to have stopped due to actions to iso l ate the re l ease path , then the effluent monitor reading is no longer valid for classification purposes.

Releases should n ot be prorated or averaged.

For example, a release exceeding 4 t im es release limits for 30 minutes does not meet the EAL. EAL RUl .1 -This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or l iquid effluent pathways. EAL RUl .2 -Th i s EAL addresses radioactiv it y releases that cause effl u ent radiation monitor readings to exceed 2 times the limit estab li s h ed b y a radioactivity discharge permit. This EAL will typically be associated wit h planned batch re l eases fro m non-continuou s release pathwa ys (e.g., radwaste, waste gas). EAL RU 1.3 -This EAL addresses uncontro ll ed gaseous or liquid relea ses that are detected by sa mple analysis or environmental surveys , particul arly on unmonitored pathways (e.g., spi ll s of radioactive liquids into storm drains , heat exchanger l eakage in river water systems, etc.). Esca lati on of the emergency classification l eve l wou ld be via IC RA 1. 22 ECL: Notificat ion of Un u s ual Eve nt Initiating Condition:

UNPLANNED l oss of water l eve l a bo ve irradi ated fue l. Operating Mode Applicability:

A ll Emergency Action Levels: RU2 RU2.l a. UNPLANNED water level d rop in the REFUELING PATHWAY as indicated b y ANY of the following:

Report to control room (v i sua l observat i on)

F uel pool l evel indication (LI-3413) l ess than 36 feet an d lo we rin g

WR GEMAC F l ood up indication (LI-4541) com in g on sca l e AND b. UNPLANNED rise in area radiation l eve l s as indicated by ANY of the following radiation monitors.

Spen t F uel Pool Area, R I-9178

North Refuel Floor, RI-9163

New F u e l Vault Area , RI-9153

Sout h Refuel Floor, RI-9164

NW Drywell Area Hi Range Rad Monitor , RIM-9184A

South Dr ywe ll Area H i Range Rad Monitor , RIM-9184B Definitions:

UNPLANNED:

A p arameter change or an eve nt that is not 1) the result of an intended evolut i on or 2) an expected plant re spo nse to a transient.

The cause of the parameter change or event may be known or unknown. REFUELING PATHWAY: T h e r eactor r efuel in g cavity , spent fue l pool and fuel transfer canal. 23 Basis: This IC addresses a decrease 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 will be primarily determined by indications from available l evel instrumentation.

Other sources of level indications may include reports from plant personnel ( e.g., from a refueling crew) or video camera observations.

A significant drop in the water level may also cause an increase 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 br i dge area radiation monitor reading may increase 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. During preparation for reactor cavity flood up prior to entry into refuel mode, reactor vessel leve l instrument LI-4541 (WR GEMAC , FLOODUP) on control room panel 1C04 is placed in service by I&C personnel connecting a compensating air signal after the reference leg is disconnected from the reactor head. Normal refuel water level is above the top of the span of this flood up level indicator.

A va l id indication (e.g., not due to loss of compensating air signal or other instrument channel failure) of reactor cavity level coming on span for this instrument is used at DAEC as an indicator of uncontrolled reactor cavity level decrease.

DAEC Technica l Specifications require a minimum of 36 feet of water in the spent fuel pool when moving irradiated fuel into the secondary containment.

During refueling, the gates between the reactor cavity and the refueling cavity are removed and the spent fuel pool level indicator LI-3413 is used to monitor refueling water level. Procedures require that a normal refueling water level be maintained at 37 feet 5 inches. A low level alarm actuates when spent fuel pool level drops below 37 feet 1 inch. Symptoms of inventory loss at DAEC include visual observation of decreasing water levels in reactor cavity or spent fuel storage pool , Reactor Building (RB) fuel storage pool radiation monitor or refueling area radiation monitor alarms, observation of a decreasing trend on the spent fuel pool water level indicator , and actuation of the spent fuel pool low water level alarm. To eliminate minor level perturbations from concern , DAEC uses LI-3413 indicated water level below 36 feet and lowering. [ncreased radiation levels can be detected by the local area radiation monitors surrounding the spent fuel pool and refueling cavity areas. Applicable area radiation monitors are those listed in AOP 981. 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 RA2. 24

RA1 ECL: Alert Initiating Condition:

Release of gaseous or liquid radioactivity resulting in offsite dose greate r than 10 mrem TEDE or 50 mrem thyroid CDE. Operating Mode Applicability:

All Emergency Action Levels: Notes:

The Emergency Director should declare the event 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 i s unknown , assume that the release duration has exceeded the specified time limit.

If the effluent flow past an effluent monitor is known to have stopped due to actions to i so late the release path, then the effluent monitor reading i s no l onger valid for classification purposes.

The pre-calculated effluent monitor values presented in EAL 1.1 should only be used for emergency classification assessments until the results from a dose assessment u sing actual meteorology are available.

RAl.1 RAl.2 RAl.3 RAl.4 Reading on ANY Table R-1 effluent radiation monitor greater than column "A lert" for 15 minutes or longer: Table R-1-Effluent Monitor Classification Thresholds

V, ::::, 0 QJ V, n, (.!J Monitor Alert Reactor Building ventilation rad monitor (Kaman 3/4 , 5/6 , 7 /8) Turbine Building ventilation rad monitor (Kaman 1/2) Offgas Sta c k rad monitor (K aman 9/10) LLR PS F rad monitor (Kaman 12) GSW rad monitor (RIS-4767) RHRSW & ESW rad monitor (RM-1997) RHRSW & ESW Rupture Disc rad monitor (RM-4268) 1.lE-02 uci/cc 1.4E-02 uci/cc 4.SE+Ol uci/cc 1.4E-02 uci/cc 1.7E+04 cps 1.2E+04 cps 1.8E+04 cps Do se assessment using actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond SITE BOUNDARY.

[Preferr ed] Analysis of a liquid effluent sample indicates a concentration or release rate that would result in do ses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for one hour of expos ure. Field survey results indicate EITHER of the fo llowin g at or beyond the SITE BOUNDARY:

Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer.

Analyses of field survey samples indicat e thyroid CDE g reater than 50 mrem for one hour of inhalation.

25 Definitions:

SITE BOUNDARY: That line beyond which the land is neither owned , nor leased , nor otherwise controlled by the licensee.

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).

This IC is modified by a note that EAL RAl .1 is only assessed for emergency classification until a qualified dose assessor is performing assessments using dose projection software incorporating actual meteorological data and current radiological conditions.

Radiological effluent EAL s 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 action s 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 RS 1. 26 RA2 ECL: Alert Initiating Condition:

Significant lowering of water level above , or damage to, irradiated fuel. Operating Mode Applicability:

All Emergency Action Levels: RA2.l RA2.2 RA2.3 Uncovery of irradiated fuel in the REFUELING PATHWAY. Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by Hi Rad alarm for ANY of the following ARMs:

Spent Fuel Pool Area, Rl-9178

North Refuel Floor , Rl-9163

New Fuel Vault Area , Rl-9153

South Refuel Floor , Rl-9164 OR Reading greater than 5 R/hr on ANY of the following radiation monitors (in Mode 5 only):

NW Drywell Area Hi Range Rad Monitor , RIM-9184A

South Drywell Area Hi Range Rad Monitor , RIM-9184B Lowering of spent fuel pool level to 25.17 feet. Definitions:

REFUELING PATHWAY -The reactor refueling cavity , spent fuel pool and fuel transfer canal. Basis: This IC addresses events that have caused IMMIN E NT or actual damage to an irradiated fuel assembly , or a significant lowering of water level within the spent fuel pool. These events present radiological safety challen g es 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. Expected radiation monitor alarm(s) during preplanned transfer of highly radioactive material through the affected areas are not considered valid alarms for the purpose of comparison to these EALs. 27 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-HUl. Escalation of the emergency would be based on either Recognition Category R or C !Cs. EALRA2.l This EAL escalates from RU2 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. 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. EALRA2.2 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.

An alarm on these 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).

Threshold values for the Drywell monitors are only applicable in Mode 5 since the calculated radiation levels from damage to irradiated fuel would be masked by the typical background levels on these monitors during plant operation, and mechanical damage to a fuel assembly in the vessel can only happen with the reactor head removed. EALRA2.3 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 RSI or RS2. 28 RA3 ECL: Alert Initiating Condition:

Radiation levels that impede access to areas necessary for normal plant operation.

Operating Mode Applicability:

All Emergency Action Levels: RA3.1 Dose rate greater than 15 mR/hr in ANY of the following areas:

Control Room (RM-9162)

Central Alarm Station (by survey) Definitions:

None 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 radiation levels and determine if another IC may be applicable.

Escalation of the emergency classification level would be via Recognition Category R , C or FI Cs. 29 RS1 ECL: Site Area Emergency Initiating Condition:

Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE. Operating Mode Applicability:

All Emergency Action Levels: Notes:

The Emergency Director should declare the event 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 the specified time limit.

If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the relea se path , then the effluent monitor reading is no longer valid for classification purposes.

The pre-calculated effluent monitor values presented in EAL 1.1 should only be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

RS 1.1 Reading on ANY Table R-1 effluent radiation monitor greater than column "SAE" for 15 minutes or longer: RSl.2 RSl.3 Ill ::::, 0 (lJ Ill n, l9 Monitor Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7 /8) Turbine Building ventilation rad monitor (Kaman 1/2) Offgas Stack rad monitor (Kaman 9/10) Dos e assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid C DE at or beyond the SITE BOUNDARY. [Preferred]

Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

Closed window dose rates grea ter than 100 mR/hr expected to continue for 60 minutes or longer.

Analyses of field survey samples indicate thyroid CDE g reater than 500 mrem for one hour of inhalation.

30 Definitions:

SITE BOUNDARY: That line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee. 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 (PA Gs). 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. This IC is modified by a note that EAL RS 1.1 is only assessed for emergency classification until a qualified dose assessor is performing assessments using dose projection software incorporating actual meteorological data and current radiological conditions.

However , if Kaman monitor readings are sustained for 15 minutes or longer and the required MIDAS dose assessments cannot be completed within this period, then the declaration can be made using Kaman readings PROVIDED the readings are not from an isolated flow path. 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. If Kaman readings are not valid , field survey results may be utilized to assess this IC using EAL RSI.3. Escalation of the emergency classification level would be via IC RG 1. 31 ECL: Site Area Emergency Initiating Condition:

Spent fue l pool level at 16.36 feet. Operating Mode Applicability:

All Emergency Action Levels: RS2.1 Lowering of spent fuel pool level to 16.36 feet. Definitions:

None Basis: RS2 This IC addresses a significant l oss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failure s 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 c l assification diversity. Escalation of the emergency class ifi cation level would be via IC RGl or RG2. 32 RG1 ECL: General Emergency Initiating Condition:

Release of gaseo u s radioactivity resulting in offsite dose greater than 1 , 000 mrem TEDE or 5,000 mrem thyroid CDE. Operating Mode Applicability:

All Emergency Action Level s: Notes:

The Emerge ncy Director shou ld declare the event promptly upon determining that the applicable time ha s been exceeded, or will lik ely be exceeded.

If an ongo in g release is detected and the release sta rt tim e is unknown , assume that the release duration has exceeded the spec ifi ed time limit.

If the efflue nt flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then t he effl uent monitor reading is no l onger valid for c la ssification purposes.

The pre-calculated efflue nt monitor values presented in EAL 1.1 s hould on l y be u sed for emergency classification assessments until the results from a dose assessment u s in g actual met eorology are available.

RG 1.1 Reading on ANY Tab l e R-1 effl u e nt radiation monitor greater than column " GE" for 15 minutes or lon ger: RGl.2 RGl.3 Mon i tor Reactor Building ventilation r ad monitor (K ama n 3/4, 5/6, 7 /8) :l Turbine Building vent il ation r ad mon it o r (Kam a n 1/2) V) "' l9 GE 1.1E+OO uci/cc 1.4E+OO uci/cc Dose assessment using actual meteorology indicates doses greater than 1 , 000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the SITE BOUNDARY.

[Preferred]

Field survey results indicate EITHER of the fo llowin g at or beyond the SITE BOUNDARY:

Closed window dose rates greater than 1 , 000 mR/hr expected to cont inu e for 60 minutes or longer.

Analyses of field s urve y samp le s indic ate thyroid CDE greater than 5 , 000 mrem for one hour of inh a l ation. 33 Definitions:

SITE BOUNDARY:

That line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

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. This IC is modified by a note that EAL RG 1.1 is only assessed for emergency classification until a qualified dose assessor is performing assessments using dose projection software incorporating actual meteorological data and current radiological conditions.

However, if Kaman monitor readings are sustained for 15 minutes or longer and the required MIDAS dose assessments cannot be completed within this period, then the declaration can be made using Kaman readings PROVIDED the readings are not from an isolated flow path. 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.

If Kaman readings are not valid, field survey results may be utilized to assess this IC using EAL RG 1.3. 34 *--_j l______ __ RG2 ECL: General Emergency Initiating Condition:

Spent fuel pool level cannot be restored to at least 16.36 feet for 60 minutes or longer. Operating Mode Applicability:

All Emergency Action Levels: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

RG2.1 Spent fuel pool level cannot be restored to at least 16.36 feet for 60 minutes or longer. Definitions:

None Basis: This IC 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.

35 7 COLD SHUTDOWN/

REFUELING SYSTEM MALFUNCTION ICS/EALS 36 ECL: Notification of Unusual Event Initiating Condition:

UNPLANNED loss ofRPV inventory for 15 minutes or longer. Operating Mode Applicability:

4, 5 Emergency Action Levels: CU1 Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

CUI.I CUl.2 UNPLANNED loss of reactor coolant results in RPV level less than a required lower limit for 15 minutes or longer. a. RPV level cannot be monitored.

AND b. UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool. Definitions:

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 RPV 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 RCS water inventory are carefully planned and controlled.

An UNPLANNED event that results in water level decreasing 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 CUI.I recognizes that the minimum required RPV 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. 37 EAL CUI .2 addresses a condition where all means to determine RPV level have been lost. If all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RSC inventory loss was occurring by observing sump and Suppression Pool level changes. The drywell floor and equipment drain sumps, reactor building equipment and floor drain sumps receive all liquid waste from floor and equipment drains inside the primary containment and reactor building.

A rise in Suppression Pool water level may be indicative of valve misalignment or leakage in systems that discharge to the Torus. Sump and Suppression Pool level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CAI or CA3. 38 CU2 ECL: Notification of Unusual Event Initiating Condition:

Loss of all but one AC power source to essential buses for 15 minutes or longer. Operating Mode Applicability:

4, 5, Defueled Emergency Action Levels: Note: The Emergency Director should declare the event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

CU2.l a. AC power capability to 1A3 and 1A4 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. Definitions:

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 systems are classified as safety-related.

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 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. An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.

A loss of all offsite power with a concurrent failure of aH but one emergency power source (e.g., an onsite diesel generator).

A loss of emergency power sources (e.g., onsite diesel generators) with a single train of essential buses being 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. 39 ECL: Notification of Unusual Event Initiating Condition:

UNPLANNED increase in RCS temperature.

Operating Mode Applicability:

4, 5 Emergency Action Levels: CU3 Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

CU3.l CU3.2 UNPLANNED increase in RCS temperature to greater than 2I2°F. Loss of ALL RCS temperature and RPV level indication for 15 minutes or longer. Definitions:

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. CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

For DAEC, this is considered to be Secondary Containment as required by Technical Specifications.

Basis: This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to determine RCS temperature and level, 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 CA3. 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 CU3.1 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 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.

40 EAL CU3.2 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.

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

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

41 ECL: Notification of Unusual Event Initiating Condition:

Loss of Vital DC power for 15 minutes or longer. Operating Mode Applicability:

4, 5 Emergency Action Levels: CU4 Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

CU4.l Indicated voltage is less than 105 VDC on BOTH Div 1 and Div 2 125 VDC buses for 15 minutes or longer. Definitions:

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 systems are classified as safety-related.

Basis: 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 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. 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 Bis 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 CAI or CA3, or an IC in Recognition Category R. 42 ECL: Notification of Unusual Event Initiating Condition:

Loss of all onsite or offsite communications capabilities.

Operating Mode Applicability:

4, 5, Defueled Emergency Action Levels: CU5.l Loss of ALL of the following onsite communication methods:

Plant Operations Radio System

In-Plant Phone System

Plant Paging System (Gaitronics)

CU5 CU5.2 Loss of ALL of the following offsite response organization communications methods:

DAEC All-Call phone

All telephone lines (PBX and commercial)

Cell Phones (including fixed cell phone system)

Control Room fixed satellite phone system

FTS Phone system CU5.3 Loss of ALL of the followingNRC communications methods:

FTS Phone system

All telephone lines (PBX and commercial)

Cell Phones (including fixed cell phone system)

Control Room fixed satellite phone system 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 offsite response organizations 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 CU5.l addresses a total loss of the communications methods used in supp011 of routine plant operations.

43 EAL CU5.2 addresses a total loss of the communications methods used to notify all offsite response organizations of an emergency declaration.

The offsite response organizations referred to here are the State of Iowa, Linn County, and Benton County. EAL CU 5 .3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

44 ECL: Alert Initiating Condition:

Loss ofRPV inventory.

Operating Mode Applicability:

4, 5 Emergency Action Levels: CA1 Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

CAl.1 CAl.2 Loss ofRPV inventory as indicated by level less than 119.5 inches. a. RPV level cannot be monitored for 15 minutes or longer AND b. UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool due to a loss ofRPV inventory.

Definitions:

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 CAl.1, a lowering of water level below 119.5 inches indicates that operator actions have not been successful in restoring and maintaining RPV 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 CA 1.1 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 EAL CAl.2, the inability to monitor RPV 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, the operators would need to determine that RCS inventory loss was occurring by observing sump and Suppression Pool level changes. The drywell floor and equipment drain sumps, reactor building equipment and floor drain sumps receive all liquid waste from floor and equipment drains inside the primary containment and reactor building.

A rise in Suppression Pool water level may be indicative of valve misalignment or leakage in systems that discharge to the Torus. Sump and Suppression Pool level increases must be 45 evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS 1 If the RPV inventory level continues to lower, then escalation to Site Area Emergency would be via IC CSL 46 ECL: Alert Initiating Condition:

Loss of all offsite and all onsite AC power to essential buses for 15 minutes or longer. Operating Mode Applicability:

4, 5, Defueled Emergency Action Levels: CA2 Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

CA2.l Loss of ALL offsite and ALL onsite AC Power to 1A3 and 1A4 buses for 15 minutes or longer. Definitions:

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 systems are classified as safety-related.

Basis: 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. When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore an emergency 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 CS 1 or RS 1. 47 ECL: Alert Initiating Condition:

Inability to maintain the plant in cold shutdown.

Operating Mode Applicability:

4, 5 Emergency Action Levels: CA3 Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

CA3.l CA3.2 UNPLANNED increase in RCS temperature to greater than 212°F for greater than the duration specified in Table C-2. Table C-2 RCS Heat-up Duration Thresholds RCS Integrity CONTAINMENT CLOSURE Heat-up Duration Status Intact Not applicable 60 minutes* Not intact 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.

UNPLANNED RCS pressure increase greater than 10 psig due to a loss of RCS cooling. Definitions:

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. CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

For DAEC, this is considered to be Secondary Containment as required by Technical Specifications.

Basis: This IC 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.

RCS integrity is 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). 48 The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact. The 20-minute criterion was included to allow time for operator action to address the temperature increase.

The RCS Heat-up Duration Thresholds table also addresses an increase 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 without a substantial degradation in plant safety. Finally, in the case where there is an increase in RCS temperature, the RCS is not intact, 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 CA3.2 provides a pressure-based indication of RCS heat-up. Escalation of the emergency classification level would be via IC CSl or RSl. 49 CA6 ECL: Alert Initiating Condition:

Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. Operating Mode Applicability:

4, 5 Emergency Action Levels: Notes: CA6.1

If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then this emergency classification is not warranted.

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.

a. AND b. The occurrence of ANY of the Table C-3 hazardous events: 1.

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 or Emergency Director Event damage has caused indications of degraded performance in one train of a SAFETY SYSTEM needed for the current operating mode. AND 2. EITHER of the following:

Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode, OR

The event has resulted in VISIBLE DAMAGE to the second train of a SAFETY SYSTEM needed for the current operating mode. 50 Definitions:

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.

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 may require a post-event inspection to determine if the attributes of an explosion are present. 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 systems are classified as safety-related.

VISIBLE DAMAGE: 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.

Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not VISIBLE DAMAGE. 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 perfonnance 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 train is from the same SAFETY SYSTEM that has indications of degraded performance for criteria CA6.l.b.1 of this EAL; commercial nuclear power plants are designed to be able to support single systerri issues without compromising public health and safety from radiological events. An event affecting a single-train SAFETY SYSTEM (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the one train) would not be classified under CA6 because the two-train impact criteria that underlie the EALs and Bases would not be met. If an event affects a single-train SAFETY SYSTEM, then the emergency classification should be made based on plant parameters/symptoms meeting the EALs for another IC. Depending upon the circumstances, classification may also occur based on Shift Manager/Emergency Director judgement.

Indications of degraded perfonnance 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. 51 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 RS I. 52 CS1 ECL: Site Area Emergency Initiating Condition:

Loss of RPV inventory affecting core decay heat removal capability.

Operating Mode Applicability:

4, 5 Emergency Action Levels: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

CSl.1 CSl.2 CSl.3 a. CONTAINMENT CLOSURE not established.

AND b. RPV level less than +64 inches a. CONTAINMENT CLOSURE established.

AND b. RPV level less than + 15 inches a. RPV level cannot be monitored for 30 minutes or longer. AND b. Core uncovery is indicated by EITHER of the following:

Drywell Monitor (9184A/B) reading greater than 5.0 R/hr

UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool of sufficient magnitude to indicate core uncovery Definitions:

CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

For DAEC, this is considered to be Secondary Containment as required by Technical Specifications.

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. 53 Basis: This IC addresses a significant and prolonged loss of RPV 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 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 reactor vessel levels of EALs CS 1.1.b and CS 1.2.b reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.

In the Cold Shutdown and Refueling Modes, LT/LI-4559, 4560, and 4561 (RX VESSEL NARROW RANGE LEVEL) instruments read up to 22" high due to hot calibrations.

LI-4541 (WR GEMAC, FLOODUP) should be used in these Modes for comparison to EAL thresholds since it is calibrated cold and reads accurately.

If normal means ofRPV level indication are not available due to plant evolutions, redundant means of RPV level indication will be normally installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted.

In EAL CS 1.3 .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 RPV 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 RPV. These EALs address 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 CGl or RGI. 54

*---

CG1 ECL: General Emergency Initiating Condition:

Loss ofRPV inventory affecting fuel clad integrity with containment challenged.

Operating Mode Applicability:

4, 5 Emergency Action Levels: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

CGl.1 CGl.2 a. RPV level less than + 15 inches for 3 0 minutes or longer. AND b. ANY indication from the Secondary Containment Challenge Table C-1. a. RPV level cannot be monitored for 30 minutes or longer. AND b. Core uncovery is indicated by EIHER of the following:

Drywell Monitor (9184A/B) reading greater than 5.0 R/hr.

UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool of sufficient magnitude to indicate core uncovery.

AND c. ANY indication from the Secondary Containment Ch_allenge Table C-1. Table C-1 Secondary Containment Challenge

CONTAINMENT CLOSURE not established*

Drywell Hydrogen or Torus Hydrogen greater than 6% AND Drywell Oxygen or Torus Oxygen greater than 5%

UNPLANNED increase in containment pressure

Secondary containment radiation monitors above max safe operating limits (MSOL) of EOP 3, Table 6

If CONTAINMENT CLOSURE is re-established prior to exceeding the 30 minute time limit, then declaration of a General Emergency is not required.

55 Definitions:

CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

For DAEC, this is considered to be Secondary Containment as required by Technical Specifications.

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 reactor vessel level above the top of active fuel with containment challenged.

This condition represents actual or IMMINENT substantial core degradation or melting with pote11tial for loss of containment integrity.

Releases can be reasonably expected to exceed EPA P AG exposure levels off site 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 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 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 bum (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 CG 1.2.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.

For EAL CGl.2.b, the calculated radiation level on the Drywell Monitors (9184A/B) is without the reactor head in place. Calculated in radiation levels with the reactor head in place are below the normal variation in background readings of these monitors.

56 The inability to monitor RPV 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 RPV. For the Containment Challenge Table, Secondary Containment max safe operating (MSOL) limits from EOP 3 are defined as the highest parameter value at which neither: (1) equipment necessary for the safe shutdown of the plant will fail nor (2) personnel access necessary for the safe shutdown of the plant will be precluded.

These EALs address 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.

57 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI} ICS/EALS 58 ECL: Notification of Unusual Event Initiating Condition:

Damage to a loaded cask CONFINEMENT BOUNDARY.

Operating Mode Applicability:

All Emergency Action Levels: E-HU1 E-HUl .1 Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by a radiation reading greater than the values shown on Table E-1 on the spent fuel cask. Table E-1 Cask Dose Rates 61BTDSC 3 feet from HSM Surface 800 mrem/hr Outside HSM Door-Centerline of DSC 200 mrem/hr End Shield Wall Exterior 40 mrem/hr Definition:

CONFINEMENT BOUNDARY:

The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry 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 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 technical specification multiple of "2 times", which is also used in Recognition Category RIC RUl, 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 I Cs HUl and HAl. 59 9 FISSION PRODUCT BARRIER ICS/EALS FA1 / L p L p 'L p RCS CTMT / 2/3-FS1 L p L p L p RCS CTMT FG1 60 FAlALERT ANY Loss OR ANY Potential Loss of EITHER the Fuel Clad OR RCS barrier. ... *:::: .. ,, ,. Fuel ([;:lad. Barrier ... ;:,.* Table F-1: DAEC EAL Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Barriers *-,i, FSl SITE AREA EMERGENCY Loss OR Potential Loss of ANY two barriers.

0 erating Mode A licability:

1, 2, 3 : *RCS Barrier :1 -: , .. _, ...*. FGlGENERALEMERGENCY

Contaitimenf Bartier. '" ! " LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS 1. RCS Activity 1. Primary Containment Conditions

1. Primary Containment Conditions A. Coolant activity Not Applicable A. Primary Not Applicable A. UNPLANNED A. Torus pressure greater than 3 00 containment rapid drop in greater than 5 3 µCi/gm dose pressure greater Drywell pressure ps1g equivalent I-131. than 2 psig due to following Drywell OR RCS leakage. pressure rise B. Drywell or Torus OR H2 cannot be B. Drywell pressure determined to be response not less than 6% and consistent with Drywell OR Torus LOCA conditions.

02 cannot be OR determined to be C. UNISOLABLE less than 5% direct downstream OR pathway to the C. HCL (Graph 4 of environment exists EOP 2) exceeded.

after primary containment isolation signal OR D. Intentional primary containment venting per EOPs 61 Fuel Clad Barrier RCS Barrier Containment Barrier ,, LOSS I POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS 2. RPV Water Level 2. RPV Water Level 2. RPV Water Level A. SAG entry is A. RPV water level A. RPV water level Not Applicable Not Applicable A. SAG entry is required cannot be restored cannot be restored required and maintained and maintained above + 15 inches above + 15 inches OR cannot be OR cannot be determined.

determined.

3. Not Applicable

3. RCS Leak Rate 3. Primary Containment Isolation Failure Not Applicable Not Applicable A. UNISOLABLE A. UNISOLABLE A. UNISOLABLE Not Applicable break in Main primary system primary system Steam, HPCI, leakage that leakage that Feedwater, results in results in RWCU, or RCIC exceeding the exceeding the as indicated by the Max Normal Max Safe failure of both Operating Limit Operating Limit isolation valves in (MNOL) of EOP (MSOL) of EOP ANY one line to 3, Table 6 for 3, Table 6 for close AND EITHER of the EITHER of the EITHER: following:

following:

HighMSL

Temperature

Temperature flow or steam OR OR tunnel

Radiation

Radiation Level temperature Level annunciators OR

Direct report of steam release OR B. Emergency RPV Depressurization required.

62 Fuel Clad Barrier RCS Barrier Containment Barrier LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS 4. Primary Containment Radiation

4. Primary Containment Radiation

4. Primary Containment Radiation A. Drywell Monitor Not Applicable A. Drywell Monitor Not Applicable Not Applicable A. Drywell Monitor (9184A/B)

(9184A/B)

(9184A/B) reading greater reading greater reading greater than 2000 R/hr. than 5 R/hr after than 5000 R/hr. OR reactor shutdown OR B. Torus Monitor B. Torus Monitor (9185A/B)

(9185A/B) reading greater reading greater than 200 R/hr than 500 R/hr 5. Other Indications

5. Other Indications

5. Other Indications A. Fuel damage Not Applicable Not Applicable Not Applicable Not Applicable A. Fuel damage assessment assessment indicates at least indicates at least 5% fuel clad 20% fuel clad damage. damage. 6. Emergency Director Judgment 6. Emergency Director Judgment 6. Emergency Director Judgment A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in the opinion of the the opinion of the the opinion of the the opinion of the the opinion of the the opinion of the Emergency Emergency Emergency Emergency Emergency Emergency Director that Director that Director that Director that Director that Director that indicates Loss of indicates Potential indicates Loss of indicates Potential indicates Loss of indicates Potential the Fuel Clad Loss of the Fuel the RCS Barrier. Loss of the RCS the Containment Loss of the Barrier. Clad Barrier. Barrier. Barrier. Containment Barrier. 63 Basis Information For DAEC EAL Fission Product Barrier Table F-1 DAEC FUEL CLAD BARRIER THRESHOLDS:

The Fuel Clad barrier consists of the zircalloy or stainless steel fuel bundle tubes that contain the fuel pellets. 1. RCS Activity Loss I.A 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 of2% 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 ofreactor 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 RCS Activity.

2. RPV Water Level Loss 2.A The Loss threshold represents any EOP requirement for entry into the Severe Accident Guidelines.

This is identified in the BWROG EPGs/SAGs when adequate core cooling cannot be assured. Potential Loss 2.A This water level co1Tesponds to the top of the active fuel and is used in the EOPs to indicate a challenge to core cooling. The RPV water level threshold is the same as RCS barrier Loss threshold 2.A. Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of the RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency.

64 DAEC FUEL CLAD BARRIER THRESHOLDS (cont.): This threshold is considered to be exceeded when, as specified in the EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization.

EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration ofRPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.

The term "cannot be restored and maintained above" means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation below the *limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.

In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL).

Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority.

For such events, ICs SA6 or SS6 will dictate the need for emergency classification.

Since the loss of ability to determine if adequate core cooling is being provided presents a significant challenge to the fuel clad barrier, a potential loss of the fuel clad barrier is specified.

3. Not Applicable (included for numbering consistency between barrier tables) 65 DAEC FUEL CLAD BARRIER THRESHOLDS (cont.): 4. Primary Containment Radiation Loss 4.A and Loss 4.B The Drywell and Torus radiation monitor readings correspond to an instantaneous release of all reactor coolant mass into the Drywell or Torus, 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 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 readings in this threshold are higher than that specified for RCS Barrier Loss threshold 4.A 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 emergency classification level to a Site Area Emergency.

There is no Potential Loss threshold associated with Primary Containment Radiation.

5. Other Indications Loss 5.A Results obtained from procedure PASAP 7.2, Fuel Damage Assessment, indicate at least 5% fuel clad damage. There is no Potential Loss threshold associated with Other Indications.

6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors that are to be used by the Emergency Director in detennining whether the Fuel Clad Barrier is lost. Potential Loss 6.A This threshold addresses any other factors that may 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.

66 DAEC RCS BARRIER THRESHOLDS:

The RCS Barrier is the reactor coolant system pressure boundary and includes the RPV and all reactor coolant system piping up to and including the isolation valves. 1. Primary Containment Conditions Loss I.A 2 psig is the drywell high pressure scram setpoint which indicates a LOCA by automatically initiating ECCS. There is no Potential Loss threshold associated with Primary Containment Pressure.

2. RPV Water Level Loss 2.A + 15 inches corresponds to the top of active fuel (T AF) and is used in the EOPs to indicate challenge to core cooling. The RPV water level threshold is the same as Fuel Clad barrier Potential Loss threshold 2.A. Thus, this threshold indicates a Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier and that appropriately escalates the emergency classification level to a Site Area Emergency.

This threshold is considered to be exceeded when, as specified in the EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack oflow pressure RPV injection sources preclude Emergency RPV depressurization EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this RCS barrier Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.

The term, "cannot be restored and maintained above," means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.

67 DAEC RCS BARRIER THRESHOLDS (cont.): In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL).

Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority.

For such events, ICs SAS or SSS will dictate the need for emergency classification.

There is no RCS Potential Loss threshold associated with RPV Water Level. 3. RCS Leak Rate Loss Threshold 3.A Large high-energy lines that rupture outside primary containment can discharge significant amounts of inventory and jeopardize the retaining capability of the RCS until they are isolated.

If it is determined that the ruptured line cannot be promptly isolated from the Control Room, the RCS barrier Loss threshold is met. Loss Threshold 3.B Emergency RPV Depressurization in accordance with the EOPs is indicative of a loss of the RCS barrier. If Emergency RPV Depressurization is performed, the plant operators are directed to open safety relief valves (SRVs) and keep them open. Even though the RCS is being vented into the suppression pool, a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary.

Potential Loss Threshold 3.A Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Nonna! Operating values in areas such as main steam line tunnel, RCIC, HPCI, etc., which indicate a direct path from the RCS to areas outside primary containment.

A Max Normal Operating Limit (MNOL) value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly.

The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification.

A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system. An UNISOLABLE leak which is indicated by MNOL values escalates to a Site Area Emergency when combined with Containment Barrier Loss threshold 3.A (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.

68 DAEC RCS BARRIER THRESHOLDS (cont.): 4. Primary Containment Radiation Loss 4.A The Drywell 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 4.A since it indicates a loss of the RCS Barrier only. There is no Potential Loss threshold associated with Primary Containment Radiation.

5. Other Indications There are no Loss or Potential Loss thresholds associated with Other Indications.

6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the RCS barrier is lost. Potential Loss 6.A 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.

69 DAEC CONTAINMENT BARRIER THRESHOLDS:

The Primary Containment Barrier includes the drywell, the wetwell, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.

1. Primary Containment Conditions Loss l .A and l .B Rapid UNPLANNED loss of drywell pressure (i.e., not attributable to drywell spray or condensation effects) following an initial pressure increase indicates a loss of drywell integrity.

Drywell pressure should increase as a result of mass and energy release into the primary containment from a LOCA. Thus, drywell pressure not increasing under these conditions indicates a loss of primary containment integrity.

These thresholds rely on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned.

The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition.

Loss l.C The use of the modifier "direct" in defining the release path discriminates against release paths through interfacing liquid systems or minor release pathways, such as instrument lines, not protected by the Primary Containment Isolation System (PCIS). 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. Following the leakage of RCS mass into primary containment and a rise in primary containment pressure, there may be minor radiological releases associated with allowable primary containment leakage through various penetrations or system components.

Minor releases may also occur if a primary containment isolation valve(s) fails to close but the primary containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of primary containment but should be evaluated using the Recognition Category R I Cs. Loss l.D EOPs may direct primary containment isolation valve logic(s) to be intentionally bypassed, even if offsite radioactivity release rate limits will be exceeded.

Under these conditions with a valid primary containment isolation signal, the containment should also be considered lost if primary containment venting is actually performed.

Intentional venting of primary containment for primary containment pressure or combustible gas control to the secondary containment and/or the environment is a Loss of the Containment.

Venting for primary containment pressure control when not in an accident situation (e.g., to control pressure below the drywell high pressure scram setpoint) does not meet the threshold condition.

70 DAEC CONTAINMENT BARRIER THRESHOLDS (cont.): L ~----Potential Loss l .A The threshold pressure is the Torus internal design pressure.

Structural acceptance testing demonstrates the capability of the primary containment to resist pressures greater than the internal design pressure.

A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier. Potential Loss l .B If hydrogen concentration reaches or exceeds the lower flammability limit, as defined in plant EOPs, in an oxygen rich environment, a potentially explosive mixture exists. If the combustible mixture ignites inside the primary containment, loss of the Containment barrier could occur. Potential Loss 1.C The Heat Capacity Limit (HCL) is the highest suppression pool temperature from which Emergency RPV Depressurization will not raise:

Suppression chamber temperature above the maximum temperature capability of the suppression chamber and equipment within the suppression chamber which may be required to operate when the RPV is pressurized, OR

Suppression chamber pressure above Primary Containment Pressure Limit A, while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent. The HCL is a function of RPV pressure, suppression pool temperature and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant and therefore, the inability to maintain plant parameters below the limit constitutes a potential loss of containment.

71 DAEC CONTAINMENT BARRIER THRESHOLDS (cont.): 2. RPV Water Level There is no Loss threshold associated with RPV Water Level. Potential Loss 2.A The Potential Loss threshold is identical to the Fuel Clad Loss RPV Water Level threshold 2.A. The Potential Loss requirement for Primary Containment Flooding indicates adequate core cooling cannot be restored and maintained and that core damage is possible.

BWR EPGs/SAGs specify the conditions that require primary containment flooding.

When primary containment flooding is required, the EPGs are exited and SA Gs are entered. Entry into SA Gs is a logical escalation in response to the inability to restore and maintain adequate core cooling. PRA studies indicate that the condition of this Potential Loss threshold could be a core melt sequence which, if not corrected, could lead to RPV failure and increased potential for primary containment failure. In conjunction with the RPV water level Loss thresholds in the Fuel Clad and RCS barrier columns, this threshold results in the declaration of a General Emergency.

3. Primary Containment Isolation Failure These thresholds address incomplete containment isolation that allows an UNISOLABLE direct release to the environment.

Loss 3.A The Max Safe Operating Limit (MSOL) for Temperature and Radiation Level are each the highest value of these parameters at which neither: (1) equipment necessary for the safe shutdown of the plant will fail, nor (2) personnel access necessary for the safe shutdown of the plant will be precluded.

EOPs utilize these temperatures and radiation levels to establish conditions under which RPV depressurization is required.

The temperatures and radiation levels should be confirmed to be caused by RCS leakage from a primary system. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system. In combination with RCS potential loss 3.A this threshold would result in a Site Area Emergency.

There is no Potential Loss threshold associated with RCS Leak Rate. 72 DAEC CONTAINMENT BARRIER THRESHOLDS (cont.): 4. Primary Containment Radiation There is no Loss threshold associated with Primary Containment Radiation.

Potential Loss 4.A The drywell radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the drywell, assuming that 20% of the fuel cladding has failed. The radiation monitor reading for the torus corresponds to an instantaneous release of all reactor coolant mass directly into the torus, 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 level to a General Emergency.

5. Other Indications There is no Loss threshold associated with Other Indications Potential Loss 5 .A Results obtained from procedure PASAP 7.2, Fuel Damage Assessment, indicate at least 25% fuel clad damage. PASAP 7.2 only shows whether fuel damage is greater than or less than 25%, thus this indication is not likely to be declared before containment barrier potential loss 4.A which indicates 20% fuel damage. However, this potential loss threshold adds an additional layer of diversity to the scheme. 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Containment barrier is lost. Potential Loss 6.A 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.

73

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74 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS 75 ECL: Notification of Unusual Event Initiating Condition:

Confirmed SECURITY CONDITION or threat. Operating Mode Applicability:

All Emergency Action Levels: HU1 HUl.l A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by DAEC Security Shift Supervision.

HUl.2 HUl.3 Notification of a credible security threat directed at DAEC. A validated notification from the NRC providing information of an aircraft threat. Definitions:

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. HOSTILE ACTION: An act toward DAEC 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 nuclear power plant. 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). 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 systems are classified as safety-related.

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 HAI, HSI and HGl. Timely and accurate communications between DAEC 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.

76 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].

EAL HUI .I references DAEC 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 IO CFR § 2.390 information.

EAL HUI .2 addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with Abnormal Operating Procedure (AOP) 914, Security Events. EAL HUI .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 Abnormal Operating Procedure (AOP) 914, Security Events. Emergency plans and implementing procedures are public documents; therefore, EALs do 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 is contained in the Security Plan. Escalation of the emergency classification level would be via IC HAI. 77 ECL: Notification of Unusual Event Initiating Condition:

Seismic event greater than OBE levels. Operating Mode Applicability:

All Emergency Action Levels: HU2 HU2.1 Seismic event greater than Operating Basis Earthquake (OBE) as indicated by receipt of the Amber Operating Basis Earthquake Light and the wailing seismic alarm on 1C35. Definitions:

DESIGN BASIS EARTHQUAKE (DBE): A DBE is vibratory ground motion for which certain (generally, safety-related) structures, systems, and components must be designed to remain functional.

OPERATING BASIS EARTHQUAKE (OBE): An OBE is vibratory ground motion for which those features of a nuclear power plant necessary for continued operation without undue risk to the health and safety of the public will remain functional.

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. 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.

OBE events are detected in accordance with AOP 901. The OBE is associated with a peak horizontal acceleration of+/- 0.06g. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SAS. 78 HU3 ECL: Notification of Unusual Event Initiating Condition:

Hazardous events Operating Mode Applicability:

All Emergency Action Levels: Note: EAL HU3.4 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.

HU3.l HU3.2 HU3.3 HU3.4 A tornado strike within the PROTECTED AREA. 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. Movement of personnel within the PROTECTED AREA is impeded due to an off site event involving hazardous materials ( e.g., an off site chemical spill or toxic gas release).

A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles.

Definitions:

PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. 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 systems are classified as safety-related.

Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. EAL HU3.l addresses a tornado striking (touching down) within the Protected Area. EAL HU3.2 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 HU3.3 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel.

within the PROTECTED AREA. 79 EAL HU3.4 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 HmTicane 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 R, F, S or C. 80 HU4 ECL: Notification of Unusual Event Initiating Condition:

FIRE potentially degrading the level of safety of the plant. Operating Mode Applicability:

All Emergency Action Levels: Notes: HU4.1 HU4.2 HU4.3 HU4.4

The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

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 alann AND b. The FIRE is located within ANY Table H-1 plant rooms or areas a. Receipt of a single fire alarm with no other indications of a FIRE. AND b. The FIRE is located within ANY Table H-1 plant rooms or areas AND c. The existence of a FIRE is not verified within 30-minutes of alarm receipt. A FIRE within the plant or ISFSI PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication.

A FIRE within the plant or ISFSI PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.

Table H-1 Fire Areas

1 G31 DG and Day Tank Rooms

1G21 DG and Day Tank Rooms

Battery Rooms

Essential Switchgear Rooms

Cable Spreading Room

TorusRoom

Intake Structnre

Pumphouse " Drywell

Torus

NE, NW, SE Comer Rooms

HPCIRoom

RCICRoom

RHR Valve Room

North CRD Area

South CRD Area

CSTs

Control Building

Remote Shutdown Panel 1 C388 Area

Panel 1C55/56 Area

SBGTRoom 81 Definitions:

FIRE: Combustion characterize.cl 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: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. 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. EALHU4.1 The intent of the IS-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 perfonned.

Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report. EALHU4.2 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 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.

EALHU4.3 In addition to a FIRE addressed by EAL HU4.1 or EAL HU4.2, a FIRE within the plant or ISFSI 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. 82 EALHU4.4 If a FIRE within the plant or ISFSI 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 Rand NFPA-805 Criterion 3 of Appendix A to 10 CFR 50 states in part 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." The Nuclear Safety Goal ("NSG") in NFPA 805, Section 1.3.1 states, "The nuclear safety goal is to provide reasonable assurance that a fire during any operational mode and plant configuration will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition." When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance because a safe shutdown success path, free of fire damage, must be available to meet the nuclear safety goals, objectives and performance criteria for a fire under any plant operational mode or configuration.

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 I-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). Even though DAEC has adopted the alternate approach provided by NFPA-805 in lieu of the deterministic requirements of Appendix R, the 30-minutes to verify a single alarm as used in EAL HU4.2 is considered a reasonable amount of time to determine if an actual FIRE exists without presenting a challenge to the nuclear safety performance criteria.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SAS. 83 HU6 ECL: Notification of Unusual Event Initiating Condition:

Other conditions exist which in the judgment of the Emergency Director warrant declaration of a NOUE. Operating Mode Applicability:

All Emergency Action Levels: HU6.l 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. Definitions:

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 systems are classified as safety-related.

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 NOUE. 84 HA1 ECL: Alert Initiating Condition:

HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes. Operating Mode Applicability:

All Emergency Action Levels: HAI.I HAl.2 A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the DAEC Security Shift Supervision.

A validated notification from NRC of an aircraft attack threat within 30 minutes of the site. Definitions:

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 DAEC 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 nuclear power plant. 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: The site property owned by or otherwise under the control of the licensee.

PROJECTILE:

An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. 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 DAEC 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).

85 The Alert declaration will also heighten the awareness of Offsite Response Organizations, allowing them to be better prepared should it be necessary to consider further actions. This IC 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. EAL HAI.I is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes any action directed against the ISFSI which is located outside the plant PROTECTED AREA. EAL HAI .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 offsite response organizations are in a heightened state of readiness.

This EAL is met when the threat-related information has been validated in accordance with Abnormal Operating Procedure (AOP) 914, Security Events. 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. Emergency plans and implementing procedures are public documents; therefore, EALs do 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 is contained in the Security Plan. Escalation of the emergency classification level would be via IC HS 1. 86 HAS ECL: Alert Initiating Condition:

Control Room evacuation resulting in transfer of plant control to alternate locations.

Operating Mode Applicability:

All Emergency Action Level: HAS.I An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel (1 C388). Definitions:

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.

Escalation of the emergency classification level would be via IC HS5. 87 HAG ECL: Alert Initiating Condition:

Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert. Operating Mode Applicability:

All Emergency Action Level: HA6.1 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. Definitions:

HOSTILE ACTION: An act toward DAEC 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 nuclear power plant. 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). HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. PROJECTILE:

An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. 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. 88 ECL: Site Area Emergency Initiating Condition:

HOSTILE ACTION within the PROTECTED AREA. Operating Mode Applicability:

All Emergency Action Levels: HS1 HSl.1 A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the DAEC Security Shift Supervision.

Definitions:

HOSTILE ACTION: An act toward DAEC 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 nuclear power plant. 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). HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. 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.

PROJECTILE:

An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. 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 DAEC 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].

89 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 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 does not apply to a HOSTILE ACTION directed at the ISFSI PROTECTED AREA which is located outside the plant PROTECTED AREA; such an attack should be assessed using IC HAI. 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 do 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 is contained in the Security Plan. Escalation of the emergency classification level would be via IC HG 1. 90 HS5 ECL: Site Area Emergency Initiating Condition:

Inability to control a key safety function from outside the Control Room. Operating Mode Applicability:

All Emergency Action Levels: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

HS5.l a. An event has resulted in plant control being transfen-ed from the Control Room to the Remote Shutdown Panel (1C388). AND b. Control of ANY of the following key safety functions is not reestablished within 20 minutes.

Reactivity control

RPV water level

RCS heat removal Definitions:

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 Shutdown Panel (1C388) is based on Emergency Director judgment.

The Emergency Director is expected to make a reasonable, informed judgment within 20 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s).

AOP 915, "Shutdown Outside Control Room" provides the following CAUTION -"For Control Room evacuation as the result of a fire, transfer of control at panels 1 C388, 1 C389, 1 C390, JC391, and JC392 is required to be completed within 20 minutes." Escalation of the emergency classification level would be via IC FGI or CGI. 91 HS6 ECL: Site Area Emergency Initiating Condition:

Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency.

Operating Mode Applicability:

All Emergency Action Level: HS6.l 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.

Definitions:

HOSTILE ACTION: An act toward DAEC 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 nuclear power plant. 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). HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. PROJECTILE:

An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. 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.

92 HG1 ECL: General Emergency Initiating Condition:

HOSTILE ACTION resulting in loss of physical control of the facility.

Operating Mode Applicability:

All Emergency Action Level: HGl.1 a. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the DAEC 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

RPV water level

RCS heat removal OR 2. Damage to spent fuel has occurred or is IMMINENT.

Definitions:

HOSTILE ACTION: An act toward DAEC 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 pati of a concerted attack on the nuclear power plant. 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). HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. 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. PROJECTILE:

An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. 93 Basis: This IC addresses an event in which a HOSTILE FORCE has taken physical control of the facility to the extent that the plant staff can no longer operate equipment necessary to maintain key safety functions.

It also addresses a HOSTILE ACTION leading to a loss of physical control that results in actual or IMMINENT damage to spent fuel due to 1) damage to a spent fuel pool cooling system ( e.g., pumps, heat exchangers, controls, etc.) or, 2) loss of spent fuel pool integrity such that sufficient water level cannot be maintained.

Timely and accurate communications between the DAEC 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].

Emergency plans and implementing procedures are public documents; therefore, EALs do 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 is contained in the Security Plan. 94 HG6 ECL: General Emergency Initiating Condition:

Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency.

Operating Mode Applicability:

All Emergency Action Level: HG6.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occmTed 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 off site for more than the immediate site area. Definitions:

HOSTILE ACTION: An act toward DAEC 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 nuclear power plant. Non-te1Torism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. 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. PROJECTILE:

An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. 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.

95 11 SYSTEM MALFUNCTION ICS/EALS 96 ECL: Notification of Unusual Event Initiating Condition:

Loss of ALL offsite AC power capability to essential buses for 15 minutes or longer. Operating Mode Applicability:

1, 2, 3 Emergency Action Level: SU1 Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

SUI.I Loss of ALL offsite AC power capability to 1A3 AND 1A4 buses for 15 minutes or longer. Definitions:

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*essential buses. This condition represents a potential reduction in the level of safety of the plant. The intent of this EAL is to declare a Notification of Unusual Event when offsite power has been lost and both of the emergency diesel generators have successfully started and energized their respective 4kv essential bus. For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the essential 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 SAl. 97 SU3 ECL: Notification of Unusual Event Initiating Condition:

UNPLANNED loss of Control Room indications for 15 minutes or longer. Operating Mode Applicability:

1, 2, 3 Emergency Action Level: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

SU3.1 An UNPLANNED event results in the inability to monitor one or more of the Table S-1 parameters from within the Control Room for 15 minutes or longer. -Table S-1>Safety System Parameters

'

Reactor power

RPV Water Level

RPV Pressure

Primary Containment Pressure

Suppression Pool Level

Suppression Pool Temperature Definitions:

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 systems are classified as safety-related.

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. 98 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, RPV level 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 RPV water 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 SA3. 99 SU4 ECL: Notification of Unusual Event Initiating Condition:

Reactor coolant activity greater than Technical Specification allowable limits. Operating Mode Applicability:

1, 2, 3 Emergency Action Levels: SU4.1 SU4.2 Pretreatment Offgas System (RM-4104)

Hi-Hi Radiation Alarm. Sample analysis indicates that reactor coolant specific activity is greater than 2.0 µCi/gm dose equivalent I-131 for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or longer. Definitions:

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. For EAL SU4.l, RM-4104 Hi-Hi Radiation Alarm has been chosen because it is operationally significant, is readily recognizable by the Control Room Operations Staff, and is set at a level corresponding to noble gas release rate, after 30-minute delay and decay of 1 Ci/sec. For EAL SU4.2, coolant samples exceeding the 2.0 µCi/gm dose equivalent I-13 lconcentration require prompt action by DAEC Technical Specifications and are representative of minor fuel cladding degradation.

Escalation of the emergency classification level would be via ICs FAl or the Recognition Category R ICs. 100 ECL: Notification of Unusual Event Initiating Condition:

RCS leakage for 15 minutes or longer. Operating Mode Applicability:

l, 2, 3 Emergency Action Levels: SUS Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

SUS.I SU5.2 SU5.3 RCS unidentified or pressure boundary leakage greater than 10 gpm for 15 minutes or longer. RCS identified leakage greater than 25 gpm for 15 minutes or longer. Leakage from the RCS to a location outside containment greater than 25 gpm for 15 minutes or longer. Definitions:

UNISOLABLE:

An open or breached system line that cannot be isolated, remotely or locally. Basis: 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. EAL SUS.I and EAL SU5.2 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).

EAL SU5.3 addresses a RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These EALs thus apply to leakage into the containment, a secondary-side system) or a location outside of containment.

The leak rate values for each EAL were selected because they are usually observable with nonnal Control Room indications.

Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation).

EAL SUS.I 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.

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 R or F. 101 ECL: Notification of Unusual Event Initiating Condition:

Automatic or manual scram fails to shutdown the reactor. Operating Mode Applicability:

1, 2 Emergency Action Levels: SU6 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.

SU6.l SU6.2 a. An automatic scram did not shutdown the reactor. AND b. ANY of the following manual actions taken at 1C05 are successful in lowering reactor power below 5% power

Manual Scram Pushbuttons

Mode Switch to Shutdown

Alternate Rod Insertion (ARI) a. A manual scram did not shutdown the reactor. AND b. EITHER of the following:

1. ANY of the following subsequent manual actions taken at 1 COS are successful in lowering reactor power below 5% power

Manual Scram Pushbuttons

Mode Switch to Shutdown

Alternate Rod Inse1iion (ARI) OR 2. A subsequent automatic scram is successful in shutting down the reactor. Definitions:

None Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic scram 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. Following the failure on an automatic reactor scram, operators will promptly initiate manual actions at the reactor control console to shutdown the reactor (e.g., initiate a manual reactor scram quickly fall to a level within the capabilities of the plant's decay heat removal systems. 102 If an initial manual reactor scram is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control console to shutdown the reactor (e.g., initiate a manual reactor scram using a different switch). Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram 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 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 scram). 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 console".

Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. The plant response to the failure of an automatic or manual reactor scram 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 F Al. Absent the plant conditions needed to meet either IC SA6 or FAl, an Unusual Event declaration is appropriate for this event. The reactor should be considered shutdown when it is producing less heat than the maximum decay heat load for which the SAFETY SYSTEMS are designed (typically 3 to 5% power). Should a reactor scram signal be generated as a result of plant work ( e.g., RPS setpoint testing), the following classification guidance should be applied.

If the signal causes a plant transient that should have included an automatic reactor scram and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.

If the signal does not cause a plant transient and the scram failure is determined through other means (e.g., assessment oftest results), then this IC and the EALs are not applicable and no classification is warranted.

103 ECL: Notification of Unusual Event Initiating Condition:

Loss of ALL onsite or offsite communications capabilities.

Operating Mode Applicability:

1, 2, 3 . Emergency Action Levels: SU7.l Loss of ALL of the following onsite communication methods:

Plant Operations Radio System

In-Plant Phone System

Plant Paging System (Gaitronics)

SU7 SU7.2 Loss of ALL of the following offsite response organization communications methods: SU7.3 Basis:

DAEC All-Call phone

All telephone lines (PBX and commercial)

Cell Phones (including fixed cell phone system)

Control Room fixed satellite phone system

FTS Phone system Loss of ALL of the following NRC communications methods:

FTS Phone system

All telephone lines (PBX and commercial)

Cell Phones (including fixed cell phone system)

Control Room fixed satellite phone system 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 offsite response organizations 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 SU7 .1 addresses a total loss of the communications methods used in support of routine plant operations.

EAL SU7 .2 addresses a total loss of the communications methods used to notify all offsite resp'onse organizations of an emergency declaration.

The offsite response organizations referred to here are the State of Iowa, Linn County, and Benton County. EAL SU7 .3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

104 SA1 ECL: Alert Initiating Condition:

Loss of ALL but one AC power source to essential buses for 15 minutes or longer. Operating Mode Applicability:

1, 2, 3 Emergency Action Level: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

SAl.1 a. AC power capability to 1A3 and 1A4 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. Definitions:

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 systems are classified as safety-related.

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 SUl. An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.

A loss *of all offsite power with a concurrent failure of all but one emergency power source ( e.g., an onsite diesel generator).

A loss of emergency power sources (e.g., onsite diesel generators) with a single train of essential buses being 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 SSL 105 SA3 ECL: Alert Initiating Condition:

UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.

Operating Mode Applicability:

1, 2, 3 Emergency Action Level: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

SA3.1 a. An UNPLANNED event results in the inability to monitor one or more Table S-1 parameters from within the Control Room for 15 minutes or longer.

Reactor power

RPV Water Level

RPV Pressure

Primary Containment Pressure

Suppression Pool Level

Suppression Pool Temperature AND b. ANY of the Table S-2 transient events are in progress.

T 1ble 5:0.2 'Sig'nificaht Transients:, -',.,"' ,, f.._ ' -_;.., "~;,,

Automatic or manual runback greater than 25% thermal reactor power

Electrical load rejection greater than 0; 25% full electrical load C, :' 11

Reactor scram ECCS actuation R

I'

Thermal power oscillations greater ti: than 10% 106 Definitions:

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 systems are classified as safety-related.

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, RPV level 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 RPV water 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 ICs FSl or IC RSl. 107 SAG ECL: Alert Initiating Condition:

Automatic or manual scram fails to shutdown the reactor, and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor. Operating Mode Applicability:

1, 2 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.

Emergency Action Level: SA6.l a. An automatic or manual scram did not shutdown the reactor. AND b. ALL of the following manual actions taken at 1C05 are not successful in lowering reactor power below 5% power

Manual Scram Pushbuttons

Mode Switch to Shutdown

Alternate Rod Insertion (ARI) Definitions:

None Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram 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 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 scram. 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 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. 108 The plant response to the failure of an automatic or manual reactor scram 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 shutdown the reactor is prolonged enough to cause a challenge to the RPV water level 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 FSl. Absent the plant conditions needed to meet either IC SS6 or FSl, 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.

The reactor should be considered shutdown when it is producing less heat than the maximum decay heat load for which the SAFETY SYSTEMS are designed (typically 3 to 5% power). 109 SAS ECL: Alert Initiating Condition:

Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. Operating Mode Applicability:

1, 2, 3 Emergency Action Level: Notes: SA8.l

If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then this emergency classification is not warranted.

If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of the SAFETY SYSTEM, then this emergency classification is not warranted.

a. AND b. The occurrence of ANY of the Table S-3 hazardous events: Ir > T;>ble S-:}-Ha~ardo;,s Events~--1 1 !-----**-**----*---** . **---*--*-**-. --1.

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 or Emergency Director Event damage has caused indications of degraded performance in one train of a SAFETY SYSTEM needed for the current operating mode. AND 2. EITHER of the following:

Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode, OR

The event has resulted in VISIBLE DAMAGE to the second train of a SAFETY SYSTEM needed for the current operating mode. 110 Definitions:

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 may 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.

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 systems are classified as safety-related.

VISIBLE DAMAGE: 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.

Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not VISIBLE DAMAGE. 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 perfonnance 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 train is from the same SAFETY SYSTEM that has indications of degraded perfonnance for criteria SA8.l.b.l of this EAL; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events. An event affecting a single-train SAFETY SYSTEM (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the one train) would not be classified under SA8 because the two-train impact criteria that underlie the EALs and Bases would not be met. If an event affects a single-train SAFETY SYSTEM, then the emergency classification should be made based on plant parameters/symptoms meeting the EALs for another IC. Depending upon the circumstances, classification may also occur based on Shift Manager/Emergency Director judgement.

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. 111 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 FSl or RSl. 112 SS1 ECL: Site Area Emergency Initiating Condition:

Loss of ALL offsite and ALL onsite AC power to essential buses for 15 minutes or longer. Operating Mode Applicability:

1, 2, 3 Emergency Action Level: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

SSl.1 Loss of ALL offsite and ALL onsite AC power to 1A3 and 1A4 buses for 15 minutes or longer. Definitions:

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 systems are classified as safety-related.

Basis: 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. 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 RGl, FGl or SGl. 113 ECL: Site Area Emergency Initiating Condition:

Loss of ALL Vital DC power for 15 minutes or longer. Operating Mode Applicability:

1, 2, 3 Emergency Action Level: SS2 Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

SS2.l Indicated voltage is less than 105 VDC on BOTH Div 1 and Div 2 125 VDC buses for 15 minutes or longer. Definitions:

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 systems are classified as safety-related.

Basis: 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 I Cs RG 1, FG 1 or SG2. 114 556 ECL: Site Area Emergency Initiating Condition:

Inability to shutdown the reactor causing a challenge to RPV water level or RCS heat removal. Operating Mode Applicability:

1, 2 Emergency Action Levels: SS6.1 a. An automatic or manual scram did not shutdown the reactor. AND b. ALL of the following manual actions taken at 1 COS are not successful in lowering reactor power below 5% power:

Manual Scram Pushbuttons

Mode Switch to Shutdown

Alternate Rod Insertion (ARI) AND c. EITHER of the following conditions exist:

RPV level cannot be restored and maintained above -25 inches. OR

HCL (Graph 4 of EOP 2) exceeded.

Definitions:

None Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram 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. The reactor should be considered shutdown when it is producing less heat than the maximum decay heat load for which the SAFETY SYSTEMS are designed (typically 3 to 5% power). Escalation of the emergency classification level would be via IC RGI or FGl. 115 SG1 ECL: General Emergency Initiating Condition:

Prolonged loss of ALL offsite and ALL onsite AC power to essential buses. Operating Mode Applicability:

1, 2, 3 Emergency ActionLevel:

Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

SOI.I a. Loss of ALL offsite and ALL onsite AC power to 1A3 and 1A4 buses. AND b. EITHER of the following:

Restoration of at least one AC essential bus in less than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is not likely. OR

RPV level cannot be restored and maintained above -25 inches. Definitions:

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 systems are classified as safety-related.

Basis: This IC addresses a prolonged loss of all power sources to AC essential 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. 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 FG 1. This will allow additional time for implementation of off site 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 essential bus by the end of the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.

The estimate for restoring at least one essential 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. 116 SG2 ECL: General Emergency Initiating Condition:

Loss of ALL AC and Vital DC power sources for 15 minutes or longer. Operating Mode Applicability:

1, 2, 3 Emergency Action Level: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

SG2.1 a. Loss of ALL offsite and ALL onsite AC power to 1A3 and 1A4 buses for 15 minutes or longer. AND b. Indicated voltage is less than 105 VDC on BOTH Div 1 and Div 2 125 VDC buses for 15 minutes or longer. Definitions:

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 systems are classified as safety-related.

Basis: This IC addresses a concurrent and prolonged loss of both AC and Vital DC power. 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. A loss of Vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and 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. 117 APPENDIX A -ACRONYMS AND ABBREVIATIONS AC ......................................................................................................................

Alternating Current AOP .................................................................................................

Abnormal Operating Procedure ATWS ...................................................................................

Anticipated Transient Without Scram BWR .............................................................................................................

Boiling Water Reactor CDE ......................................................................................................

Committed Dose Equivalent CFR ......................................................................................................

Code of Federal Regulations CNMT ...........................................................................................................................

Containment DC ..............................................................................................................................

Direct Current EAL ...........................................................................................................

Emergency Action Level ECCS ............................................................................................

Emergency Core Cooling System ECL ................................................................................................

Emergency Classification Level EOF ..................................................................................................

Emergency Operations Facility EOP ...............................................................................................

Emergency Operating Procedure EPA .............................................................................................

Environmental Protection Agency EPG ...............................................................................................

Emergency Procedure Guideline FEMA .............................................................................

Federal Emergency Management Agency GE ......................................................................................................................

General Emergency HCL ...................................................................................................................

Heat Capacity Limit HPCI ..............................................................................................

High Pressure Coolant Injection IC ........................................................................................................................

lnitiating Condition ID .............................................................................................................................

Inside Diameter ISFSI ...........................................................................

Independent Spent Fuel Storage Installation Keff ....................................................................................

Effective Neutron Multiplication Factor LCO ...............................................................................................

Limiting Condition of Operation LOCA ........................................................................................................

Loss of Coolant Accident mR, mRem, mrem, mREM ............................................................

milli-Roentgen Equivalent Man MW ....................................................................................................................................

Megawatt NEI .............................................................................................................

Nuclear Energy Institute NRC ..............................................................................................

Nuclear Regulatory Commission NORAD .................................................................

North American Aerospace Defense Command NOUE ..............................................................................................

Notification Of Unusual Event NUMARC 1 ***************************************************************

Nuclear Management and Resources Council OBE .......................................................................................................

Operating Basis Earthquake OCA .............................................................................................................

Owner Controlled Area ODAM .........................................................................................

Offsite Dose Assessment Manual PA ..............................................................................................................................

Protected Area PAG .......................................................................................................

Protective Action Guideline PRA/PSA ....................................

Probabilistic Risk Assessment

/ Probabilistic Safety Assessment PWR ........................................................................................................

Pressurized Water Reactor PSIG .................................................................................................

Pounds per Square Inch Gauge R .........................................................................................................................................

Roentgen RCIC ...............................................................................................

Reactor Core Isolation Cooling RCS .............................................................................................................

Reactor Coolant System Rem, rem, REM ......................................................................................

Roentgen Equivalent Man 1 NUMARC was a predecessor organization of the Nuclear Energy Institute (NEI). A-1 RPS .........................................................................................................

Reactor Protection System RPV .............................................................................................................

Reactor Pressure Vessel RWCU ..........................................................................................................

Reactor Water Cleanup SCBA .... ......................................................

................

.. .........

Self-Contained Breathing Apparatus SPDS ............................................................................................

Safety Parameter Display System TEDE .............................................................................................

Total Effective Dose Equivalent TAF .....................................................................................................................

Top of Active Fuel TSC ..........................................................................................................

Techn1cal Support Center UFSAR .................................................................................

Updated Final Safety Analysis Report A-2 APPENDIX B -DEFINITIONS The following definitions are taken from Title 10, Code of Federal Regulations, and related regulatory guidance documents.

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 P AG exposure levels. 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. Notification of Unusual Event (NOUE): 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 off site response or monitoring are expected qnless further degradation of SAFETY SYSTEMS occurs. 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.

The following are key terms necessary for overall understanding the DAEC emergency classification scheme. 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 ofnames 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) B-1 Fission Product Barrier Threshold:

A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier. 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.

Selected terms used in Initiating Condition and Emergency Action Level statements 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. CONFINEMENT BOUNDARY:

The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. This corresponds to the pressure boundary for the Dry Shielded Canister (DSC) shell (including the inner bottom cover plate) base metal and associated confinement boundary welds. CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

For DAEC, this is considered to be Secondary Containment as required by Technical Specifications.

DESIGN BASIS EARTHQUAKE (DBE): A DBE is vibratory ground motion for which certain (generally, safety-related) structures, systems, and components must be designed to remain functional.

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 may 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.

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 nuclear power plant 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 nuclear power plant. 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.

B-2 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. 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.

OPERATING BASIS EARTHQUAKE (OBE): An OBE is vibratory ground motion for which those features of a nuclear power plant necessary for continued operation without undue risk to the health and safety of the public will remain functional.

OWNER CONTROLLED AREA: This term is typically taken to mean the site property owned by or otherwise under the control of the licensee.

PROJECTILE:

An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. REFUELING PATHWAY: Includes all the cavities, tubes, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel. 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 systems are classified as safety-related.

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 BOUNDARY:

That line beyond which the land is neither owned, nor leased, nor otherwise controlled by the Company. UFSAR Figure 1.2-1 identifies the DAEC SITE BOUNDARY.

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. VISIBLE DAMAGE: 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.

Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not VISIBLE DAMAGE. B-3 ATTACHMENT 3 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELATING TO LICENSE AMENDMENT REQUEST TSCR-166 UPDATED DEVIATIONS AND DIFFERENCES MATRIX 100 pages follow UPDATED DAEC DEVIATIONS AND DIFFERENCES MATRIX TABLE OF CONTENTS

GENERAL COMMENT

S ................................................................................................................................

1 ABNORMAL RAD LEVELS/ RADIOACTIVE EFFLUENT ICS/EALS ...................................................................

5 COLD SHUTDOWN/

REFUELING SYSTEM MALFUNCTION ICS/EALS ........................................................

20 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS ....................................................

36 FISSION PRODUCT BARRIER ICS/EALS .......................................................................................................

38 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS ..............................................

47 SYSTEM MALFUNCTION ICS/EALS .............................................................................................................

63 APPENDIX A-ACRONYMS AND ABBREVIATIONS

....................................................................................

84 APPENDIX B -DEFINITIONS

.......................................................................................................................

89 APPENDIX C -PERMANENTLY DEFUELED ICS/EALS ..................................................................................

98 UPDATED DAEC DEVIATIONS AND DIFFERENCES MATRIX

GENERAL COMMENT

S Page 1 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 DAEC Change Justification Validatidn

'. # .; ,* GLOBAL#l References to NEI 99-01 Replaced with DAEC Difference Convert generic guidance to DAEC specific.

None GL0BAL#2 Effective date Replaced with TBD, 2018 Difference Convert generic guidance to DAEC specific.

None GL0BAL#3 Defined terms in Appendix B; Defined terms in Appendix B; Difference All defined terms in Appendix B used in the Title Case Upper Case document are in upper case (CAPs} to None indicate that the terms are defined. GL0BAL#4 PWR specific references PWR references removed Difference DAEC is a BWR None GLOBAL#S Recognition Category A-Recognition Category R-Difference DAEC implemented the optional Abnormal Radiation Abnormal Radiation designation of "R" for radiological related Levels/Radiological Effluent Levels/Radiological Effluent items to maintain continuity with previous None category and Emergency Action category and Emergency Action practice at DAEC. Levels; AU, AA, AS, and AG Levels; RU, RA, RS, and RG GL0BAL#6 Permanently Defueled Section Deleted references to Difference Not Applicable to DAEC None Permanently Defueled Station GL0BAL#7 Acknowledgments, Notice and Deleted Difference Not Applicable to DAEC None Executive Summary GL0BAL#8 Parameters or indications listed Some parameters or indications Difference Tables or bullets were created to present in EALs listed in EALs were placed in DAEC-specific information in a manner None tables or bulletized lists. familiar to and desired by scheme users. GL0BAL#9 Site specific information or "Site specific information or Difference Compliance with intent of the guidance.

indication statements indications" were replaced with None DAEC-specific information or indications where applicable.

GLOBAL#lO Operating Mode Applicability lists Operating Mode Applicability lists Difference Mode numbers used for consistency with mode names (i.e., Power mode numbers (i.e., 1, 2, etc.} DAEC procedures and training.

None Operation, Startup} GLOBAL#ll Developer's Notes Developer's Notes deleted Difference Developer's notes are not reflected in the implementation of the EALs. None GL0BAL#12 Example EAL statement "Example" deleted from Difference In adopting the EAL, the "example" status statement is no longer applicable.

None GL0BAL#13 The following terms: "all, any, or, Consistently capitalized and Difference Capitalized and balded conditional terms in either" are sometimes capitalized balded the following terms: "ALL, !Cs and EALs for consistency based on user None and/or balded in !Cs and EALs ANY, OR, EITHER" in !Cs and EALs. feedback.

GL0BAL#14 Defined terms are only listed in Defined terms are also listed as in Difference Aid to the user to present all needed APPENDIX B -DEFINITIONS separate section of each IC/EAL information within the same section of the None where the terms are used. Basis document.

GLOBAL#lS Term "emergency buses" Replaced with "essential buses" Difference Changed to reflect DAEC nomenclature None 2 DAEC DEVIATIONS AND DIFFERENCES MATRIX , , Section NEI 99.:01 Rev. 6 DAEC Change Justification Validation " ', ', # " \ '. COVER PAGE Development of Emergency Duane Arnold Emergency Action Difference Changes made to adapt the generic NEI None Action Levels for Non-Passive Level Technical Bases Document guidance to a DAEC-specific document Reactors Introduction Acknowledgments, Notice and Deleted Difference Not Applicable to DAEC None Executive Summary TOC 1. Regulatory Background

1. Basis for Emergency Action Difference Title change None Levels TOC 1.1 Operating Reactors 1.1 Regulatory Background Difference Title change None TOC 1.2 Permanently Defueled Station Deleted section Difference Not Applicable to DAEC None TOC 1.3 Independent Spent Fuel 1.2 Independent Spent Fuel Difference Re-numbered None Storage Installation (ISFSI) Storage Installation (ISFSI) TOC 1.4 NRC Order EA-12-051 1.3 NRC Order EA-12-051 Difference Re-numbered None TOC 1.5 Applicability of Advance and Deleted section Difference Not Applicable to DAEC None Small Modular Reactor Designs TOC 3.Design of the NEI 99-01 3. Design of the DAEC Emergency Difference Title Change None Emergency Classification Scheme Classification Scheme TOC 3.3 NSSS Design Differences Deleted section Difference Changes made to adapt the generic NEI None guidance to a DAEC-specific document TOC 3.4 Organization and Changed to 3.3 DAEC 3.4 Difference Changes made to adapt the generic NEI None Presentation of Generic Organization and Presentation of guidance to a DAEC-specific document Information Generic Information TOC 4.0 Site-Specific Scheme 4.0 DAEC Scheme Development Difference Title change None Development TOC 4.4; 4.5; 4.6; 4.8 Deleted sections Difference Changes made to adapt the generic NEI None guidance to a DAEC-specific document TOC 4.7 Developer and User Feedback None TOC Appendix (-Permanently Deleted section Difference Changes made to adapt the generic NEI None Defueled Station ICs/EALs guidance to a DAEC-specific document 1.1 Regulatory Background Regulatory Background Difference Changes made to adapt the generic NE! None guidance to a DAEC-specific document and removed developer information 1.2 Permanently Defueled Station Section deleted Difference Not Applicable to DAEC None 1.3 1.3 Independent Spent Fuel 1.2 Independent Spent Fuel Difference Re-numbered section. None Storage Installation (ISFSI) Storage Installation (ISFSI) 3 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 DAEC Change Justification Validation

" .. ., ,-,, " '* 1.4 1.4 NRC Order EA-12-051 1.3 NRC Order EA-12-051 Difference Re-numbered and removed wording to add None these readings (DAEC installation completed).

1.5 Applicability to Advanced and Section deleted Difference Not Applicable to DAEC None Small Modular Reactor Designs 2 KEY TERMINOLOGY USED IN NEI KEY TERMINOLOGY USED IN Difference Minor changes to reflect DAEC-specific None 99-01 DAEC EAL SCHEME implementation.

3 DESIGN OF THE NEI 99-01 DESIGN OF THE DAEC Difference Changes made to adapt the generic NEI None EMERGENCY CLASSIFICATION EMERGENCY CLASSIFICATION guidance to a DAEC-specific document SCHEME SCHEME 3.1 Assignment of Emergency Assignment of Emergency Difference Changes made to adapt the generic NEI None Classification Levels (ECLs) Classification Levels (ECLs) guidance to a DAEC-specific document, removed references to PWRs, and removed developer information.

3.2 Types of Initiating Conditions and Types of Initiating Conditions and Verbatim None Emergency Action Levels Emergency Action Levels 3.3 Text referring to NSSS design Deleted Difference Guidance is now DAEC specific None differences for various types or plants; Developer guidance 3.4 Organization and Presentation of DAEC-Specific Organization and Difference Renumbered to 3.3, made DAEC-specific, None Generic Information Presentation of Generic and deleted developer information Information 3.5 Mode of Applicability Matrix; Deleted "Permanently Defueled" Difference Renumbered to 3.4, removed PWR Vl Typical BWR Operating Modes section of matrix; replaced information, removed permanently Typical BWR Operating Modes defueled, and inserted DAEC Operating with DAEC-specific Operating Modes to comply with the document intent. Modes 4 Site Specific Scheme Development of the DAEC Difference Upda.ted to reflect DAEC specific scheme None Development Guidance Emergency Classification Scheme development process. 5 GUIDANCE ON MAKING GUIDANCE ON USING THE DAEC Difference Guidance is now DAEC specific None EMERGENCY CLASSIFICATIONS EALS 6-11 Recognition Category IC/EAL removed Difference Matrixes were intended for use by EAL None Matrixes developers.

Inclusion in licensee scheme is not desired. 4 DAEC DEVIATIONS AND DIFFERENCES MATRIX ABNORMAL RAD LEVELS/ RADIOACTIVE EFFLUENT ICS/EALS 5 DAEC DEVIATIONS AND DIFFERENCES MATRIX Recognition Category:

AUl RUl Difference Global Comment #5 None Initiating Condition:

Release of Release of gaseous or liquid Difference Global Comment #9 None gaseous or liquid radioactivity radioactivity greater than 2 times greater than 2 times the {site-the ODAM limits for 60 minutes .-i ::::> specific effluent release or longer. <t controlling document}

limits for 60 minutes or longer. Operating Mode of Applicability:

Operating Mode of Applicability:

Verbatim None All All 6 DAEC DEVIATIONS AND DIFFERENCES MATRIX I : DA.EC *"-,.}

Cf:tange 1

Justification*

* *;,:* I V~lidatio9

-I (1) Reading on ANY effluent (1) Reading on ANY Table R-1 Difference See Global Comments #8, 9, 12, & 13. V2 radiation monitor greater effluent radiation monitor than 2 times the (site-greater than column "NOUE" Reworded EAL statement to remove specific effluent release for 60 minutes or longer: operator confusion as to whether they controlling document) needed to multiply the values of the limits for 60 minutes or [inserted Table R-1 of DAEC-following table by 2 or if the value provided longer: (site-specific specific radiation monitors already was 2X. Wording now matches monitor list and threshold and threshold values] wording of RSl and RGl allowing for easier values corresponding to 2 operator progression through the EALs. times the controlling document limits) (2) Reading on ANY effluent (2) Reading on ANY effluent Difference Global Comment #13 None radiation monitor greater radiation monitor greater than 2 times the alarm than 2 times the alarm -. ...: setpoint established by a setpoint established by a = 0 current radioactivity current radioactivity '-' discharge permit for 60 discharge permit for 60 'l"'"'I minutes or longer. minutes or longer. (3) Sample analysis for a (3) Sample analysis for a gaseous Difference Global Comment #9 None gaseous or liquid release or liquid release indicates a indicates a concentration concentration or release rate or release rate greater greater than 2 times the than 2 times the (site-ODAM limits for 60 minutes specific effluent release or longer. controlling document) limits for 60 minutes or longer. Intent and meaning of the EALs are not altered. 7 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section . ! .. NEI 99-01 Rev. 6 ,, . " '~.' .. 1 . DAEC Change I . Ju~tification I Valida~ion

. I Recognition Category:

AU2 RU2 Difference Global Comment #5 & 14 None Initiating Condition:

UNPLANNED UNPLANNED loss of water level Verbatim None loss of water level above above irradiated fuel. irradiated fuel. Operating Mode of Applicability:

Operating Mode of Applicability:

Verbatim None All All (1) a. UNPLANNED water level (1) a UNPLANNED water level Difference Global Comment #9, 12 & 13 V3 drop in the REFUELING drop in the REFUELING PATHWAY as indicated by PATHWAY as indicated by ANY of the following:

ANY of the following:

N (site-specific level

Report to control ::::) <C indications).

room (visual observation)

Fuel pool level indication (Ll-3413) less than 36 feet and lowering

WR GEMAC Floodup indication (Ll-4541) coming on scale AND AND 8 DAEC DEVIATIONS AND DIFFERENCES MATRIX **Section

' Justification V~lidatio1V#

b. UNPLANNED increase in b. UNPLANNED rise in area Difference Global Comments #9 & 13 V4 area radiation levels as radiation levels as indicated by ANY of the indicated by ANY of the following radiation following radiation monitors.

monitors. (site-specific list of area

Spent Fuel Pool Are~ radiation monitors)

Rl-9178

North Refuel Floor, RI-9163

New Fuel Vault Area, -Rl-9153 ...; South Refuel Floor, RI-C

0 9164 N NW Drywell Area Hi ::::, * <( Range Rad Monitor, RIM-9184A

South Drywell Area Hi Range Rad Monitor, RIM-91848 Intent and meaning ofthe EALs are not altered. 9 DAEC DEVIATIONS AND DIFFERENCES MATRIX section J ' .'"NEl,99~01 Rev.* G * '** DAEC :-' ',, Change .. ' ...

Justifica~ion , Validatioh

., I' .. ' *::. .. Recognition Category:

AAl RAl Difference Global Comment #5 & 14 None Initiating condition:

Release of Release of gaseous or liquid Verbatim None gaseous or liquid radioactivity radioactivity resulting in offsite resulting in offsite dose greater dose greater than 10 mrem TEDE than 10 mrem TEDE or 50 mrem or 50 mrem thyroid CDE. thyroid CDE. Operating Mode of Applicability:

Operating Mode of Applicability:

Verbatim None All All (1) Reading on ANY of the (1) Reading on ANY Table R-1 Difference Global Comment #8, 9, 12 & 13 vs following radiation radiation monitor greater monitors greater than the than column "Alert" for 15 reading shown for 15 minutes or longer: minutes or longer: (site-specific monitor list [inserted Table R-1 of DAEC-and threshold values) specific radiation monitors and threshold values] .-1 (2) Dose assessment using (2) Dose assessment using actual Difference Global Comment #9 None <C <C actual meteorology meteorology indicates doses Added bracketed

'Preferred' to reinforce indicates doses greater greaterthan 10 mrem TEDE the 4th Note of the IC than 10 mrem TEDE or 50 or 50 mrem thyroid CDE at or mrem thyroid CDE at or beyond the SITE BOUNDARY.

beyond (site-specific dose [Preferred]

receptor point). (3) Analysis of a liquid (3) Analysis of a liquid effluent Difference Global Comment #9 effluent sample indicates sample indicates a a concentration or release concentration or release rate rate that would result in that would result in doses doses greater than 10 greater than 10 mrem TEDE mrem TEDE or 50 mrem or 50 mrem thyroid CDE at or thyroid CDE at or beyond beyond the SITE BOUNDARY (site-specific dose for one hour of exposure.

receptor point) for one hour of exposure.

10 DAEC DEVIATIONS AND DIFFERENCES MATRIX I ' ' DAEC Change

I' .. ' Justification (4) Field survey results (4) Field survey results indicate Difference Global Comment #9 None indicate EITHER of the EITHER of the following at or following at or beyond beyond the SITE BOUNDARY: (site-specific dose

Closed window dose receptor point): rates greater than 10 -* Closed window dose mR/hr expected to .... rates greater than 10 continue for 60 minutes s: 0 mR/hr expected to or longer. .-4 continue for 60 Analyses of field survey

minutes or longer. samples indicate thyroid

Analyses offield survey COE greater than SO samples indicate mrem for one hour of thyroid CDE greater inhalation.

than SO mrem for one hour of inhalation.

Intent and meaning of the EALs are not altered. 11

-DAEC DEVIATIONS AND DIFFERENCES MATRIX DAEC I" Change , I *' Justification I V~lidation

I Recognition Category:

AA2 RA2 Difference Global Comment #5 & 14 None Initiating Condition:

Significant Significant lowering of water Verbatim None lowering of water level above, or level above, or damage to, damage to, irradiated fuel. irradiated fuel. Operating Mode of Applicability:

Operating Mode of Applicability:

Verbatim None All All (1) Uncovery of irradiated fuel (1) Uncovery of irradiated fuel in Verbatim None in the REFUELING the REFUELING PATHWAY. PATHWAY. (2) Damage to irradiated fuel (2) Damage to irradiated fuel Difference Global Comment #8, 9, 12 & 13 V6 resulting in a release of resulting in a release of radioactivity from the fuel radioactivity from the fuel as as indicated by ANY of the indicated by Hi Rad alarm for following radiation ANY of the following ARMs: monitors:

Spent Fuel Pool Area, RI-9178 (site-specific listing of radiation

North Refuel Floor, Rl-9163 N monitors, and the associated

New Fuel Vault Area, RI-readings, setpoints and/or 9153 alarms)

South Refuel Floor, Rl-9164 OR Threshold values for the Drywell monitors Reading greater than 5 R/hr are only applicable in Mode 5 since the on ANY of the following calculated radiation levels from damage to radiation monitors (in Mode irradiated fuel would be masked by the 5 only): typical background levels on these

NW Drywell Area Hi Range monitors during plant operation, and Rad Monitor, RIM-9184A mechanical damage to a fuel assembly in

South Drywell Area Hi the vessel can only happen with the reactor Range Rad Monitor, RIM-head removed (Mode 5). 91848 (3) Lowering of spent fuel pool (3) Lowering of spent Difference Global Comment #9 V7 level to (site-specific Level fuel pool level to 2 value). [See Developer 25.17 feet Intent and meaning of the EALs are not Notes altered. 12 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section "NEI 99-01 Rev. 6 DAEC ~hange Justification i Validation

I Recognition Category:

AA3 RA3 Difference Global Comment #5 & 14 None Initiating Condition:

Radiation Radiation levels that impede Difference Reworded IC to reflect non-applicability of None levels that impede access to access to areas necessary for EAL #2. equipment necessary for normal normal plant operation.

plant operations, cooldown or shutdown.

Operating Mode Applicability:

All Operating Mode Applicability:

All Verbatim None (1) Dose rate greater than 15 (1) Dose rate greater than 15 Difference Global Comment #9, 12 & 13 None mR/hr in ANY of the mR/hr in ANY of the following areas: following areas:

Control Room

Control Room ARM (RM-* Central Alarm Station 9162} * (other site-specific

Central Alarm Station (by areas/rooms) survey) rtl (2) An UNPLANNED event Not used at DAEC Difference EALs RA3 and HAS are not applicable to V8 results in radiation levels DAEC because an evaluation has shown that prohibit or impede that there are no rooms or areas that access to any ofthe contain equipment which require a following plant rooms or manual/local action as specified in areas: operating procedures used for normal plant operation, cooldown and shutdown.

All (site-specific list of plant rooms areas outside the Control Room that or areas with entry-related mode contain equipment necessary for normal applicability identified) plant operation, cooldown and shutdown do not require physical access to operate. Intent and meaning of the EALs are not altered. 13 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 DAEC j . Change j .. Justification I Validation

I Recognition Category:

ASl RSl Difference Global Comment #5 & 14 None Initiating Condition:

Release of Release of gaseous radioactivity Verbatim None gaseous radioactivity resulting in resulting in offsite dose greater offsite dose greater than 100 than 100 mrem TEDE or 500 mrem TEDE or 500 mrem thyroid mrem thyroid CDE. CDE. Operating Mode Applicability:

All Operating Mode Applicability:

All Verbatim None (1) Reading on ANY of the (1} Reading on ANY Table R-1 Difference Global Comment #8, 9, 12 & 13 V9 following radiation effluent radiation monitor monitors greater than the greater than column "SAE" reading shown for 15 for 15 minutes or longer. minutes or longer: .-I (site-specific monitor list and [inserted Table R-1 of DAEC-V) <( threshold values) specific radiation monitors and threshold values] (2) Dose assessment using (2) Dose assessment using actual Difference Global Comment #3 & 9 None actual meteorology meteorology indicates doses Added bracketed

'Preferred' to reinforce indicates doses greater greater than 100 mrem TEDE the 4th Note of the IC than 100 mrem TEDE or or 500 mrem thyroid CDE at 500 mrem thyroid CDE at or beyond the SITE or beyond (site-specific BOUNDARY.

dose receptor point). [Preferred]

14 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 DAEC Change Justification Valjdation

(3) Field survey results (3) Field survey results indicate Difference Global Comment #3, 9, & 13 None indicate EITHER of the EITHER of the following at or following at or beyond beyond the SITE BOUNDARY: (site-specific dose receptor point):

Closed window dose

Closed window dose rates greater than 100 rates greater than 100 mR/hr expected to mR/hr expected to continue for 60 minutes continue for 60 or longer. minutes or longer.

Analyses of field survey

Analyses of field survey samples indicate thyroid samples indicate CDE greater than 500 thyroid CDE greater -mrem for one hour of than 500 mrem for one C: inhalation.

hour of inhalation.

0 .... V) <( Intent and meaning of the EALs are not altered. 15 L ____ -

DAEC DEVIATIONS AND DIFFERENCES MATRIX Secfion "NH99.:.01 Rev~ 6 DAEC: . Change Justification Validation

. Recognition Category:

AS2 RS2 Difference Global Comment #5 None Initiating Condition:

Spent fuel Spent fuel pool level at 16.36 feet Difference Global Comment #9 VlO pool level at (site-specific Level 3 description).

Operating Mode Applicability:

All Operating Mode Applicability:

All Verbatim None N (1) Lowering of spent fuel pool (1) Lowering of spent fuel pool Difference Global Comment #9 & 12 VlO Ill <C level to (site-specific Level level to 16.36 feet 3 value). Intent and meaning ofthe EALs are not altered. 16 DAEC DEVIATIONS AND DIFFERENCES MATRIX 1:* .* Section I

NEI 99-01.Re\r; . 6

I DAEC .. , Change I ... , Justification*

1 vaUdatip'n

1 Recognition Category:

AGl RGl Difference Global Comment #5 & 14 None Initiating Condition:

Release of Release of gaseous radioactivity Verbatim None gaseous radioactivity resulting in resulting in offsite dose greater offsite dose greater than 1,000 than 1,000 mrem TEDE or 5,000 mrem TEDE or 5,000 mrem mrem thyroid CDE. thyroid CDE. Operating Mode Applicability:

All Operating Mode Applicability:

All Verbatim None (1) Reading on ANY of the (1) Reading on ANY Table R-1 Difference Global Comment #8, 9, 12 & 13 V9 following radiation effluent radiation monitor monitors greater than the greater than column "GE" for .... reading shown for 15 15 minutes or longer . minutes or longer: <C (site-specific monitor list and [inserted Table R-1 of DAEC-threshold values) specific radiation monitors and threshold values] (2) Dose assessment using (2) Dose assessment using actual Difference Global Comment #3 & 9 None actual meteorology meteorology indicates doses Added bracketed

'Preferred' to reinforce indicates doses greater greater than 1,000 mrem the 4th Note of the IC than 1,000 mrem TEDE or TEDE or 5,000 mrem thyroid 5,000 mrem thyroid CDE at CDE at or beyond the SITE or beyond (site-specific BOUNDARY.

[Preferred]

dose receptor point). 17 --~

DAEC DEVIATIONS AND DIFFERENCES MATRIX DAEC * , . Change , . Justification I Validati(;)n

I (3) Field survey results (3) Field survey results indicate Difference Global Comment #3 & 9 None indicate EITHER of the EITHER of the following at or following at or beyond beyond the SITE BOUNDARY: (site-specific dose receptor

Closed window dose point): rates greater than 1,000

Closed window dose mR/hr expected to rates greater than 1,000 continue for 60 minutes mR/hr expected to or longer. continue for 60 minutes

Analyses of field survey or longer. samples indicate thyroid -* Analyses of field survey CDE greater than 5,000 ... samples indicate thyroid mrem for one hour of C 0 CDE greater than 5,000 inhalation.

.-l mrem for one hour of C, <C inhalation.

Intent and meaning of the EALs are not altered. 18 DAEC DEVIATIONS AND DIFFERENCES MATRIX I*: Section *f :.:. NEI 99~01 Rev: 6 Ju~tification J Validatio.n

'# I .. . ' Recognition Category:

AG2 RG2 Difference Global Comment #5 None Initiating Condition:

Spent fuel Spent fuel pool level cannot be Difference Global Comment #9 VlO pool level cannot be restored to restored to at least 16.36 feet for at least (site-specific Level 3 60 minutes or longer. description) for 60 minutes or longer. Operating Mode Applicability:

All Operating Mode Applicability:

All Verbatim None N (1) Spent fuel pool level cannot (1) Spent fuel pool level cannot Difference Global Comment #9 & 12 VlO <( be restored to at least (site-be restored to at least 16.36 specific Level 3 value) for 60 feet for 60 minutes or longer. minutes or longer. Intent and meaning of the EALs are not altered. 19 DAEC DEVIATIONS AND DIFFERENCES MATRIX COLD SHUTDOWN/

REFUELING SYSTEM MALFUNCTION ICS/EALS 20 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section .NEI 99-01 Rev: 6 DAEC Change Justification j Validation

[ Recognition Category:

CUl CUl Verbatim Global Comment #11, 14 None Initiating Condition:

UNPLANNED UNPLANNED loss of RPV Difference Global Comment #4 None loss of (reactor vessel/RCS

[PWR] inventory for 15 minutes or or RPV [BWR]) inventory for 15 longer minutes or longer. Operating Mode Applicability:

Operating Mode Applicability:

4, Difference Global Comment #10 None Cold Shutdown, Refueling 5 (1) UNPLANNED loss of (1) UNPLANNED loss of reactor Difference Global Comment #4 & 12 None reactor coolant results in coolant results in RPV level (reactor vessel/RCS

[PWR] less than a required lower or RPV [BWR]) level less limit for 15 minutes or than a required lower limit longer. .-1 for 15 minutes or longer. ::> u (2) a. (Reactor vessel/RCS

[PWR] (2) a. RPV level cannot be Difference Global Comment #4 None or RPV [BWR]) level cannot monitored.

be monitored.

AND AND b. UNPLANNED increase in b. UNPLANNED level rise in Difference Global Comment #9 None (site-specific sump and/or Drywell/Reactor Building tank) levels. Equipment or Floor Drain sump, or Suppression Pool. Intent and meaning of the EALs are not altered. Recognition Category:

CU2 CU2 Verbatim Global Comment #11, 14 None Initiating Condition:

Loss of all Loss of all but one AC power Difference Global comment #15 None but one AC power source to source to essential buses for 15 emergency buses for 15 minutes minutes or longer. or longer. N Operating Mode Applicability:

Operating Mode Applicability:

4, Difference Global Comment #10 None ::> u Cold Shutdown, Refueling, 5, Defueled Defueled (1) a. AC power capability to (1) a. AC power capability to Difference Global Comment #9, 12, & 13 Vll (site-specific emergency 1A3 and 1A4 buses is buses) is reduced to a reduced to a single power 21 single power source for 15 source for 15 minutes or minutes or longer. longer. AND AND b. Any additional single b. Any additional single power source failure will power source failure will result in loss of all AC result in loss of ALL AC power to SAFETY SYSTEMS. power to SAFETY SYSTEMS. Intent and meaning of the EALs are not altered. 22 DAEC DEVIATIONS AND DIFFERENCES MATRIX . Section NEI 99-01 Rev. 6 DAEC -I Change . j iustification I Validation

I Recognition Category:-CU3 CU3 Verbatim Global Comment #11, 14 None Initiating Condition:

UNPLANNED UNPLANNED increase in RCS Verbatim None increase in RCS temperature.

temperature.

Operating Mode Applicability:

4, Difference Global Comment #10 None Cold Shutdown, Refueling 5 (1) UNPLANNED increase in (1) UNPLANNED increase in RCS Difference Global Comment #9 & 12 Vl RCS temperature to temperature to greater than greater than (site-specific 212°F Technical Specification cold shutdown temperature limit). (2) Loss of ALL RCS (2) Loss of ALL RCS temperature Difference Global Comment #4 & 13 None temperature and (reactor and RPV level indication for m vessel/RCS

[PWR] or RPV 15 minutes or longer ::::, [BWR]) level indication for u 15 minutes or longer. Intent and meaning of the EALs are not altered. 23 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section *. ** NEI 99-01 Rev: 6 DAEC I Change Justification I Validation

I Recognition Category:

CU4 CU4 Verbatim Global Comment #11, 14 None Initiating Condition:

Loss of Vital Loss of Vital DC power for 15 Verbatim None DC power for 15 minutes or minutes or longer. longer. Operating Mode Applicability:

Operating Mode Applicability:

4, Difference Global Comment #10 None Cold Shutdown, Refueling 5 ,::I' (1) Indicated voltage is less (1) Indicated voltage is less than Difference Global Comment #9, 12, 13 V12 ::::, than (site-specific bus 105 VDC on BOTH Div 1 and u voltage value) on required Div 2 125 VDC buses for 15 Vital DC buses for 15 minutes or longer minutes or longer. Intent and meaning of the EALs are not altered. 24 DAEC DEVIATIONS AND DIFFERENCES MATRIX Settion NEI 99-Ql Rev. 6

DAEC , .. Change . Jµstification I Validation

I Recognition Category:

CU5 CU5 Verbatim None Initiating Condition:

Loss of all Loss of all onsite or offsite Verbatim None onsite or offsite communications communications capabilities.

capabilities.

Operating Mode Applicability:

4, Difference Global Comment #10 None Cold Shutdown, Refueling, 5, Defueled Defueled (1) Loss of ALL of the following (1) Loss of ALL of the following Difference Global Comment #9, 12 & 13 V13 onsite communication onsite communication methods: methods: (site-specific list of

Plant Operations Radio communications methods) System

In-Plant Phone System

Plant Paging System in :::::, (Gaitronics) u (2) Loss of ALL of the following (2) Loss of ALL of the following Difference Global Comment #9 & 13 V13 ORO communications offsite response organization V14 methods: communications methods: (site-specific list of

DAEC All-Call phone communications methods)

All telephone lines (PBX and commercial)

Cell Phones (including fixed cell phone system)

Control Room fixed satellite phone system

FTS Phone system 25 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 DAEC

I change Ju~tification I Validation

I (3) Loss of ALL of the following (3) Loss of ALL of the following Difference Global Comment #9, 12 & 13 V13 NRC communications NRC communications methods: methods: (site-specific list of

FTS Phone system -communications methods)

All telephone lines (PBX +: C and commercial) 0 Cell Phones (including in * :::> fixed cell phone system) u

Control Room fixed satellite phone system Intent and meaning of the EALs are not altered. 26 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section < .NEI 99-01. Rev. *6 DAEC Change Justification Validation

Recognition Category:

CAl CAl Verbatim Global Comment #11, 14 None Initiating Condition:

Loss of Loss of RPV inventory.

Difference Global Comment #4 None (reactor vessel/RCS

[PWR] or RPV [BWR]) inventory.

Operating Mode Applicability:

4, Difference Global Comment #10 None Cold Shutdown, Refueling 5 (1) Loss of (reactor vessel/RCS (1) Loss of RPV inventory as Difference Global Comment #4, 9 & 12 V15 [PWR] or RPV [BWR]) indicated by level less than inventory as indicated by 119.5 inches level less than (site-specific level). (2) a. (Reactor vessel/RCS

[PWR] (2) a. RPV level cannot be Difference Global Comment #4 None or RPV [BWR]) level cannot monitored for 15 minutes .... be monitored for 15 or longer <C u minutes or longer AND AND Difference Global Comment #4, 9 & 13 None b. UNPLANNED increase in b. UNPLANNED level rise in (site-specific sump and/or Drywell/Reactor Building tank) levels due to a loss of Equipment or Floor Drain (reactor vessel/RCS

[PWR] sump, or Suppression Pool or RPV [BWR]) inventory.

due to a loss of RPV inventory.

Intent and meaning ofthe EALs are not altered. 27 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev: *6 DAEC Change Justification

! Validation

! Recognition Category:

CA2 CA2 Verbatim Global Comment #11, 14 None Initiating Condition:

Loss of all Loss of all offsite and all onsite Difference Global Comment #15 None offsite and all onsite AC power to AC power to essential buses for emergency buses for 15 minutes 15 minutes or longer. or longer. Operating Mode Applicability:

Operating Mode Applicability:

4, Difference Global Comment #10 None N <( Cold Shutdown, Refueling, 5, Defueled u Defueled (1) Loss of ALL offsite and ALL (1) Loss of ALL offsite and ALL Difference Global Comment #9, 12 & 13 V12 onsite AC Power to (site-onsite AC Power to 1A3 and specific emergency buses) 1A4 for 15 minutes or longer. for 15 minutes or longer. Intent and meaning of the EALs are not altered. 28

DAEC DEVIATIONS AND DIFFERENCES MATRIX Recognition Category:

CA3 CA3 Verbatim Global Comment #11, 14 None Initiating Condition:

Inability to Inability to maintain the plant in Verbatim None maintain the plant in cold cold shutdown.

shutdown.

Operating Mode Applicability:

4, Difference Global Comment #10 None Cold Shutdown, Refueling 5 (1) UNPLANNED increase in (1) UNPLANNED increase in RCS Difference Global Comment #9 & 12 Vl RCS temperature to temperature to greater than greater than (site-specific 212°F for greater than the Technical Specification duration specified in Table C-cold shutdown 2. temperature limit) for greater than the duration specified in the following table. Table: RCS Heat-up Duration Threi h~lrlr .c C-2 RCS Heat-up Duration Th Difference Global Comment #4 None RCS Status Containment sea -up C'I) Closure Status Duration . Containment Closure Changed "RCS Status" to "RCS Integrity" to <( u Intact (but not at *-** -a, ... Status match current site nomenclature reduced inventory Not applicable i:;n m;~ .. +nr* [PWR]) Intact Not Applicable Established N%iffirncJtes*

Established Not intact (or at reduced inventory

[PWR]) Not Established 0 minutes Not Established

If an RCS heat removal system is in ope ratio

If a&i RCS heat removal system is in operatior l wy:fam t JMi'R1cs emperature is being reduced, frame and RCS temperature is being reduce , tntpjfJB1i?t applicable.

(2) UNPLANNED RCS pressure (2) UNPLANNED RCS pressure Difference Global Comment #4 & 9 V16 increase greater than (site-increase greater than 10 psig Added "due to a loss of RCS cooling" to specific pressure reading).

due to a loss of RCS cooling. clarify the intent of the EAL (This EAL does not apply during water-solid plant conditions.

[PWR]) Intent and meaning of the EALs are not altered. 29 Recognition Category:

CA6 Initiating Condition:

Hazardous event affecting a SAFE1Y SYSTEM needed for the current operating mode. Operating Mode Applicability:

Cold Shutdown, Refueling (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 DAEC DEVIATIONS AND DIFFERENCES MATRIX CA6 Hazardous event affecting a SAFE1Y SYSTEM needed for the current operating mode. Operating Mode Applicability:

4, 5 (1) a. The occurrence of ANY of the Table C-3 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 or Emergency Director 30 Verbatim Global Comment #11, 14 Verbatim Difference Global Comment #10 Difference Global Comment #9, 12 & 13 None None None None DAEC DEVIATIONS AND DIFFERENCES MATRIX AND AND b. EITHER of the following:

b. 1. Event damage has Deviation Adopted the revised EAL wording provided V17 1. Event damage has caused indications of in approved EAL FAQ 2016-02. caused indications of degraded degraded performance performance in one in at least one train of a train of a SAFETY SAFETY SYSTEM needed SYSTEM needed for for the current the current operating operating mode. mode. AND OR 2. EITHER of the Deviation Adopted the revised EAL wording provided V17 1. The event has caused following:

in approved EAL FAQ 2016-02. VISIBLE DAMAGE to a

Event damage has SAFETY SYSTEM caused indications Difference Added the following clarification to the V18 component or structure of degraded Basis from EALFAQ 2018-04: -needed for the current performance to a An event affecting a single-train SAFETY ...; operating mode. second train of the SYSTEM (i.e., there are indications of s:: 0 SAFETY SYSTEM degraded performance and/or VISIBLE U) needed for the DAMAGE affecting the one train) would not <( u current operating be classified under SA8 because the two-mode, or train impact criteria that underlie the EALs

The event has and Bases would not be met. If an event resulted in VISIBLE affects a single-train SAFETY SYSTEM, then DAMAGE to the the emergency classification should be second train of a made based on plant SAFETY SYSTEM parameters/symptoms meeting the EALs needed for the for another IC. Depending upon the current operating circumstances, classification may also occur mode. based on Shift Manager/Emergency Director judgement.

Intent and meaning of the EALs are not altered. 31 DAEC DEVIATIONS AND DIFFERENCES MATRIX Recognition Category:

CSl CSl Verbatim Global Comment #11, 14 None Initiating Condition:

Loss of Loss of reactor vessel/RCS Difference Global Comment #4 None (reactor vessel/RCS

[PWR] or RPV inventory affecting core decay [BWR]) inventory affecting core heat removal capability.

decay heat removal capability.

Operating Mode Applicability:

4, Difference Global Comment #10 None Cold Shutdown, Refueling 5 (1) a. CONTAINMENT CLOSURE (1) a. CONTAINMENT CLOSURE Difference Global Comment #9 & 12 V19 not established.

not established.

.... AND AND II) b. (Reactor vessel/RCS

[PWR] b. RPV level less than +64 u or RPV [BWR]) level less inches than (site-specific level). (2) a. CONTAINMENT CLOSURE (2) a. CONTAINMENT CLOSURE Difference Global Comment #4 & 9 V19 established.

established.

AND AND b. (Reactor vessel/RCS

[PWR] b. RPV level less than +15 or RPV [BWR]) level less inches than (site-specific level). 32 I I L -.... s:: 0 .... V) u (3) a. b. (Reactor vessel/RCS

[PWR] or RPV [BWR]) level cannot be monitored for 30 minutes or longer. AND 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)

DAEC DEVIATIONS AND DIFFERENCES MATRIX (3) a. RPV level cannot be Difference Global Comment #4 None monitored for 30 minutes or longer. AND b. Core uncovery is indicated Difference Global Comment #9 &13 V6 by ANY of the following:

Drywell Monitor (9184A/B) reading greater than 5.0 R/hr

UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool of sufficient magnitude to indicate core uncovery Intent and meaning of the EALs are not altered. 33 DAEC DEVIATIONS AND DIFFERENCES MATRIX Recognition Category:

CGl CGl Verbatim Global Comment #11, 14 None Initiating Condition:

Loss of Loss of reactor vessel/RCS Difference Global Comment #4 None (reactor vessel/RCS

[PWR] or RPV inventory affecting fuel clad [BWR]) inventory affecting fuel integrity with containment clad integrity with containment challenged.

challenged.

Operating Mode Applicability:

4, Difference Global Comment #10 None Cold Shutdown, Refueling 5 (1) a. (Reactor vessel/RCS

[PWR] (1) a. RPV level less than +15 Difference Global Comment #4, 9, 12 & 13 V19 or RPV [BWR]) level less inches for 30 minutes or ..-1 than (site-specific level) for longer. u 30 minutes or longer. AND AND b. ANY indication from the b. ANY indication from the Containment Challenge Containment Challenge Table (see below). Table (see below). (2) a. (Reactor vessel/RCS

[PWR] (2) a. RPV level cannot be Difference Global Comment #4 None or RPV [BWR]) level cannot monitored for 30 minutes be monitored for 30 or longer. minutes or longer. 34

DAEC DEVIATIONS AND DIFFERENCES MATRIX AND AND Difference Global Comment #8, 9 & 13 V6 b. Core uncovery is indicated

b. Core uncovery is indicated by ANY of the following:

by ANY of the following:

{Site-specific radiation

Drywell Monitor monitor) reading greater (9184A/B) reading than (site-specific value) greater than 5.0 R/hr

Erratic source range

Erratic source range monitor indication monitor indication

[PWR]

UNPLANNED level rise

UNPLANNED increase in in Drywell/Reactor (site-specific sump Building Equipment or and/or tank) levels of Floor Drain sump, or sufficient magnitude to Suppression Pool of indicate core uncovery sufficient magnitude to AND indicate core uncovery AND C. ANY indication from the C. ANY indication from the Difference Global Comment #9 None Containment Challenge Secondary Containment Table (see below). Challenge Table C-1. Containment Challenge Table Table C-1 Containment Challenge n Difference Global Comment #9 V20 ONTAINMENT CLOSURE not established*

CONTAINMENT CLOSURE not established V21 xplosive mixture} exists inside containment

Drywell Hydrogen or Torus Hydrogen gre AND Drywell Oxygen or Torus Oxygen gn NPLANNED increase in containment pressure ~condary containment radiation monitor reading

UNPLANNED increase in containment pre ite specific value} [BWR]

Secondary containment radiation monito safe operating limits (MSOL} of EOP 3, Ta *If CONTAINMENT CLOSURE is

If CONTAINMENT CLOSURE is re-established prior to exceeding Verbatim re-established prior to exceeding the 30-minute time limit, then the 30-minute time limit, then declaration of a General Intent and meaning of the EALs are not declaration of a General Emergency is not required.

altered. Emergency is not required.

35 DAEC DEVIATIONS AND DIFFERENCES MATRIX INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS 36 DAEC DEVIATIONS AND DIFFERENCES MATRIX Sectipn NEI 99-01 Rev. 6 j* DAEC Change Justification j v~li_dation

j Recognition Category:

E-HUl E-HUl Verbatim None Initiating Condition:

Damage to a Damage to a loaded cask Verbatim None loaded cask CONFINEMENT CONFINEMENT BOUNDARY.

BOUNDARY.

Operating Mode Applicability:

All Operating Mode Applicability:

All Verbatim None (l} Damage to a loaded cask (1) Damage to a loaded cask Difference Global Comment #8, 9, 12 & 14 V22 CONFINEMENT BOUNDARY CONFINEMENT BOUNDARY as indicated by an on-as indicated by a radiation contact radiation reading reading greater than the greater than (2 times the values shown on Table E-1 on site-specific cask specific the spent fuel cask. technical specification

...... allowable radiation level) Table E-1 Cask Dose :::> J: on the surface of the spent Rates I LU fuel cask. 6IBTDSC 800 mrem/hr 3 feet from HSM 200 mrem/hr Surface Outside HSMDoor-40 Centerline mrem/hr of DSC Intent and meaning of the EALs are not altered. 37 DAEC DEVIATIONS AND DIFFERENCES MATRIX FISSION PRODUCT BARRIER ICS/EALS The following section is configured in a manner that is different from the Fission Product Barrier Tables in the DAEC EAL Technical Bases Document.

Where the Technical Bases Document evaluates all three fission product barriers simultaneously for a specific sub-category, this matrix presents each fission product barrier individually for all sub-categories.

The significance of this presentation is that where the fission product barrier table in the Technical Bases Document moves vertically through the categories, this matrix moves horizontally.

38 DAEC DEVIATIONS AND DIFFERENCES MATRIX Fission Product Barrier Emergency Classifications ,NEI 99-01 Rev. 6 DAEC Change Justification Validation

Table 9-F-1: Recognition Category "F" Initiating Condition Matrix Alert Site Area General Emergency Emergency Any Loss or Loss or Loss of any two any Potential Potential barriers and Loss of either Loss of any Loss or Deleted per developer note. Mode the Fuel Clad two barriers.

Potential Loss Difference applicability carried over onto Table 9-F None Deleted EAL listing. or RCS barrier. of the third barrier. Global Comment #11 Op Modes: Op Modes: Op Modes: Power Power Power Operation, Hot Operation, Operation, Hot Standby, Hot Standby, Standby, Startup, Hot Startup, Hot Startup, Hot Shutdown Shutdown Shutdown Table 9-F-2: BWR EAL Fission Product Barrier Table F-1: DAEC EAL Fission Product Renumbered and re-labeled due to Table Thresholds for LOSS or POTENTIAL LOSS of Barrier Table Thresholds for LOSS or Difference deletion ofTables 9-F-1 & 3. None Barriers POTENTIAL LOSS of Barriers Added Global Comment #9 Table 9-F-3: PWR EAL Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Deleted Difference Global Comment #4 None Barriers Basis lnfqrmation For BWR EAL Fission Product Deleted Developer Notes Difference Transform generic NEI 99-01 guidance Barrier Table 9-F Developer Notes. into DAEC-specific application.

None Figure 9-F-4: PWR Containment Integrity or Deleted Difference Global Comment #4 None Bypass Example 39

Ssub~tafegory, ;i..-.Rts Activity 2. RPV Water

J~v:.~i: 4. Primary Containment R~cliati~ri'ti*

' .. . *, .. ,' .' ' DAEC DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of Fuel Clad Barrier [\JEI 99-01 Rev.* 6 , .'

DAEC . "'~ Loss A. (Site-specific

- indications that reactor coolant activity is greater than 300 µCi/gm dose equivalent 1-131). Potential Loss Not Applicable Loss A. Coolant activity greater than 300 µCi/gm dose equivalent 1-131 Potential Loss Not Applicable

Change . J_ustification . Difference General Comment #9 A. Primary containment A. RPV water level A. SAG entry is required.

A. RPV water level cannot Difference EPFAQ 2015-004 flooding required.

cannot be restored and maintained above (site-specific RPV water level corresponding to the top of active fuel) or cannot be determined.

Not Applicable A. Primary containment radiation monitor reading greater than (site-specific value). A. (site-specific as applicable)

Not Applicable Not Applicable A. (site-specific as applicable)

Not Applicable A. Drywell Monitor (9184A/B) reading greater than 2000 R/hr. OR B. Torus Monitor (9185A/B) reading greater than 200 R/hr A. Fuel damage assessment indicates at least 5% fuel clad damage. 40 be restored and maintained V15 above +15 inches OR cannot be determined.

Not Applicable Not Applicable Not Applicable General Comment #9, 13 Verbatim None Difference V23 Global Comment #9 Difference Global Comment #9 Core damage assessment procedure.

DAEC DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of Fuel Clad Barrier Table 9-F NEI 99-01 Rev. 6 DAEC Change Justification

.. *Sub-Category Loss Potential Loss Loss Potential Loss 6. Emergency A. ANY condition in the A. ANY condition in A. ANY condition in A. ANY condition in the Verbatim None Di.rector

... * .,: ' opinion of the the opinion of the the opinion of the opinion of the Emergency Judgment:

Emergency Director Emergency Director Emergency Director Director that indicates that indicates Loss of that indicates that indicates Loss Potential Loss of the Fuel ., the Fuel Clad Barrier. Potential Loss of of the Fuel Clad Clad Barrier. ,, ,, the Fuel Clad Barrier. Barrier. 41 Table 9-F .i: Primary :C~nt~in*rnent, Pressure Renamed to ... '): .. f>rim~iy:'.' . Containment

~Qnditjoiis

' :2:RPV Water L~vel DAEC DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of RCS Barrier . ,:-:. NEI 99-01 Rev. 6 Loss Potential Loss A. Primary containment Not Applicable pressure greater than (site-specific value) due to RCS leakage. A. RPV water level Not Applicable cannot be restored and maintained above (site-specific RPV water level corresponding to the top of active fuel) or cannot be determined.

A. UNISOLABLE break in . , ANY of the following:

* (site-specific systems with potential for energy line breaks) OR B. Emergency RPV Depressurization.

A. OR UNISOLABLE primary system leakage that results in exceeding EITHER of the following:

1. Max Normal Operating Temperature

2. Max Normal Operating Area Radiation Level. DAEC Loss Potential Loss A. Primary Not Applicable containment pressure greater than 2 psig due to RCS leakage. A. RPV water level Not Applicable cannot be restored and maintained above +15 inches OR cannot be determined.

A. UNISOLABLE break in Main Steam, HPCI, Feedwater, RWCU, or RCIC as indicated by the failure of both isolation valves in ANY one line to close AND EITHER:

High MSL flow or steam tunnel temperature annunciators OR

Direct report of steam release OR B. Emergency RPV Depressu rization required.

42 A . UNISOLABLE primary system leakage that results in exceeding the Max Normal Operating Limit (MNOL) of EOP 3, Table 6 for EITHER of the following:

Temperature OR

Radiation Level Change .. . . Justification Difference V24 Global Comment #9 Difference V19 Global Comment #9, 13 Difference V25 Global Comment #9 Added site-specific indication of an unisolable steam line break which includes failure of both isolation valves to LOSS 3.A.

DAEC DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of RCS Barrier 4. Primary A. Primary Not Applicable A. Drywell Monitor Not Applicable Difference Global Comment #9 Containment containment radiation (9184A/B) reading V23 Radiation.

monitor reading greater than 5 R/hr greater than (site-after reactor ' specific value). shutdown 5. Other ,, .. A. (site-specific as A. (site-specific as Not Applicable Not Applicable Difference Global Comment #9 indications applicable) applicable)

6. Emergency*

A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in the Verbatim None Director the opinion of the the opinion of the the opinion of the opinion of the J.u~gment.

Emergency Director Emergency Emergency Emergency Director that indicates Loss Director that Director that that indicates Potential of the RCS Barrier. indicates Potential indicates Loss of Loss of the RCS Barrier. Loss of the RCS the RCS Barrier. '\. Barrier. 43 DAEC DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of Containment Barrier Table 9-F-2 NEI 99~01 Rev. 6 .* DAEC Gtiange J ustific;ation

Sub:category Loss Potential Loss Loss Potential Loss 1. Primary A. UNPLANNED rapid drop A. Primary containment A. UNPLANNED rapid A. Torus pressure Difference Global Comment #9 .CQntainmen~.

in primary containment pressure greater than drop in Drywell greater than 53 psig V20 Coi:aditions.

pressure following (site-specific value) pressure following OR V26 primary containment OR Drywell pressure rise B. Drywell or Torus H2 V27 .. pressure rise B. (site-specific OR ... . . ' cannot be OR explosive mixture) exists B. Drywell pressure determined to be Primary Containment B. Primary containment inside primary response not less than 6% and Isolation Failure Loss 3.A pressure response not containment consistent with LOCA Drywell OR Torus and 3.B moved to sub-consistent with LOCA OR conditions.

02 cannot be category 1 "Primary conditions.

C. HCTL exceeded.

OR determined to be Containment Conditions" .. C. UNISOLABLE direct less than 5% as Losses 1.C and 1.D to ,. downstream pathway OR consolidate concepts into . . .. .. to the environment single sub-category exists after primary C. HCL (Graph 4 of containment isolation EOP 2) .* exceeded.

",' .. signal .. OR .. D. Intentional primary .. containment venting per EOPs 2. RPY.Water Not Applicable A. Primary Not Applicable A. SAG entry is Difference EPFAQ 2015-004 Levei* " containment required.

flooding required.

44 1able 9-F-2 . 3: P~imary * .

Containment ls9lati~n

:-,;t,:_ :-Failure 4.: Primary. Containment Rai:li~tion

.'i:!;}.

DAEC DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of Containment Barrier NEI 99-01 Rev. 6 DAEC Loss A. UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal OR B. Intentional primary containment venting per EOPs OR C. UNISOLABLE primary system leakage that results in exceeding EITHER of the following:

1. Max Safe Operating Temperature.

OR 2. Max Safe Operating Area Radiation Level. Not Applicable Potential Loss Not Applicable Loss A. UNISOLABLE primary system leakage that results in exceeding the Max Safe Operating Limit (MSOL) of EDP 3, Table 6 for EITHER of the following:

Temperature OR

Radiation Level A. Primary containment Not Applicable radiation monitor reading greater than (site-specific value). 45 Potential Loss Not Applicable A. Drywell Monitor (9184A/B) reading greater than 5000 R/hr. OR B. Torus Monitor (9185A/B) reading greater than 500 R/hr Difference . Justification

.. Global Comment #9 V28 Primary Containment Isolation Failure Loss 3.A and 3.B moved to category 1 "Primary Containment Conditions" as Losses 1.C and 1.D to consolidate concepts into single sub-category Difference Global Comment #9 V23

, Table 9-F-2 * 't 5. Other Indications.: J' '". < '6;~~mer~en~y . Dfr'E!Ctor

,. *, ' JiJ(Jgment , 'e; "2i~~' ;',':!)ti' Loss A. (site-specific as applicable)

A. ANY condition in the opinion of the Emergency Director that indicates Loss of* the Containment Barrier. DAEC DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of Containment Barrier Potential Loss A. (site-specific as applicable)

Loss Not Applicable Potential Loss A. Fuel damage assessment indicates at least 20% fuel clad damage. Change. Difference B. ANY condition in the C. ANY condition in the D. ANY condition in the Verbatim opinion of the Emergency Director that indicates Potential Loss of the Containment Barrier. opinion of the Emergency Director that indicates Loss of the Containment Barrier. 46 opinion of the Emergency Director that indicates Potential Loss of the Containment Barrier. Justification . : Global Comment #9 Core damage assessment procedure.

None

~-

DAEC DEVIATIONS AND DIFFERENCES MATRIX HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS 47 DAEC DEVIATIONS AND DIFFERENCES MATRIX I* Section***

r .

NEI 99.:01.Rev.

6 Justification , I Validation,#

I Recognition Category:

HUl HUl Verbatim Global Comment #11, 14 None Initiating Condition:

Confirmed Confirmed SECURITY CONDITION Verbatim None SECURITY CONDITION or threat. or threat. Operating Mode Applicability:

All Operating Mode Applicability:

All Verbatim None (1) A SECURITY CONDITION (1) A SECURITY CONDITION that Difference Global Comment #9 & 12 None that does not involve a does not involve a HOSTILE HOSTILE ACTION as ACTION as reported by DAEC reported by the (site-Security Shift Supervision.

specific security shift supervision).

.-1 (2) Notification of a credible (2) Notification of a credible Difference Global Comment #9 None :::, security threat directed at security threat directed at :::c the site. DAEC. (3) A validated notification (3) A validated notification.from Verbatim None None from the NRC providing the NRC providing information of an aircraft information of an aircraft threat. threat. Intent and meaning of the EALs are not altered. 48 DAEC DEVIATIONS AND DIFFERENCES MATRIX j' S'edion Just!fication j Vali_datio'i:t

.#, j Recognition Category:

HU2 HU2 Verbatim Global Comment #11, 14 None Initiating Condition:

Seismic Seismic event greater than OBE Verbatim None event greater than OBE levels. levels. Operating Mode Applicability:

All Operating Mode Applicability:

All Verbatim None (1) Seismic event greater than (1) Seismic event greater than Difference Global Comment #9 & 12 V29 Operating Basis Operating Basis Earthquake Earthquake (OBE) as (OBE) as indicated by receipt N indicated by: of the Amber Operating

> (site-specific indication that a Basis Earthquake Light and :c seismic event met or exceeded the wailing seismic alarm on OBE limits) 1C35. Intent and meaning of the EALs are not altered. 49 DAEC DEVIATIONS AND DIFFERENCES MATRIX j; . Section *

f * .. NEI 99:...01 Rev:. 6 I ' .. ' ' DAEC * 'r* 'Change Justification

! . vii(lidation

**, Recognition Category:

HU3 HU3 Verbatim Global Comment #11, 14 None Initiating Condition:

Hazardous Hazardous event. Verbatim None event. Operating Mode Applicability:

All Operating Mode Applicability:

All Verbatim None (1) A tornado strike within the (1) A tornado strike within the Verbatim Global Comment #12 None PROTECTED AREA. PROTECTED AREA. (2) Internal room or area (2) Internal room or area Verbatim None flooding of a magnitude flooding of a magnitude sufficient to require sufficient to require manual manual or automatic or automatic electrical electrical isolation of a isolation of a SAFETY SYSTEM SAFETY SYSTEM component needed for the component needed for the current operating mode. current operating mode. (3) Movement of personnel (3) Movement of personnel Verbatim None within the PROTECTED within the PROTECTED AREA AREA is impeded due to an is impeded due to an offsite ("/) offsite event involving event involving hazardous

, hazardous materials (e.g., materials (e.g., an offsite :::c: an offsite chemical spill or chemical spill or toxic gas toxic gas release).

release).

(4) A hazardous event that (4) A hazardous event that Verbatim None results in on-site results in on-site conditions conditions sufficient to sufficient to prohibit the plant prohibit the plant staff staff from accessing the site from accessing the site via via personal vehicles.

personal vehicles.

(5) (Site-specific list of natural Difference Global Comment #9 or technological hazard events) Intent and meaning of the EALs are not altered. 50 DAEC DEVIATIONS AND DIFFERENCES MATRIX . NEI 99:-01 Rev: 6 J

DAEC

J* Change

J Justification

I Validation

*I Recognition Category:

HU4 HU4 Verbatim Global Comment #11, 14 None Initiating Condition:

FIRE FIRE potentially degrading the Verbatim None potentially degrading the level of level of safety of the plant. safety of the plant. Operating Mode Applicability:

All Operating Mode Applicability:

All Verbatim None (1) a. A FIRE is NOT extinguished (1) a. A FIRE is NOT extinguished Difference Global Comment #12 & 13 None within 15-minutes of ANY within 15-minutes of ANY of the following FIRE of the following FIRE detection indications:

detection indications:

Report from the field

Report from the field (i.e., visual observation) (i.e., visual

Receipt of multiple observation) (more than 1) fire

Receipt of multiple alarms or indications (more than 1) fire

Field verification of a alarms or indications single fire alarm

Field verification of a AND single fire alarm o::t AND :::, b. The FIRE is located within b. The FIRE is located within Difference Global Comment #8, 9, & 13 J: None ANY of the following plant ANY Table H-1 plant rooms or areas: rooms or areas. (site-specific list of plant rooms Table H-1 Fire Areas or areas)

1G31 DG and Day Tank Rooms

1G21 DG and Day Tank Rooms

Battery Rooms

Essential Switchgear Rooms

Cable Spreading Room

Torus Room

Intake Structure

Pumphouse

Drywell

Torus

NE, NW, SE Corner Rooms

HPCI Room

RCIC Room

RHR Valve Room

North CRD Area

South CRD Area

CSTs

Control Building

Remote Shutdown Panel 1C388 Area

Panel lCSS/56 Area

SBGTRoom 51 DAEC DEVIATIONS AND DIFFERENCES MATRIX

Section j * ; NEI 99;01 Rev. *G .

DAEC:' .

I Change I '* Justification I \/alidatiOJ1

1 (2) a. Receipt of a single fire (2) a. Receipt of a single fire Difference Global Comment #8, 9 & 13 None alarm (Le., no other alarm with no other indications of a FIRE). indications of a FIRE. AND AND b. The FIRE is located within b. The FIRE is located within ANY of the following plant ANY Table H-1 plant rooms rooms or areas: or areas. (site-specific list of plant rooms or areas) AND AND c. The existence of a FIRE is c. The existence of a FIRE is Verbatim N/A None not verified within 30-not verified within 30-minutes of alarm receipt. minutes of alarm receipt. (3) A FIRE within the plant or (3) A FIRE within the plant or Difference Global Comment #9 None -ISFSI [for plants with an ISFSI PROTECTED AREA not ....; C 0 ISFSI outside the plant extinguished within 60 "" Protected Area] minutes of the initial ::::, PROTECTED AREA not report, alarm or indication.

c extinguished within 60-minutes of the initial report, alarm or indication.

(4) A FIRE within the plant or (4) A FIRE within the plant or Difference Global Comment #9 None ISFSI [for plants with an ISFSI PROTECTED AREA ISFSI outside the plant that requires firefighting Protected Area] support by an offsite fire PROTECTED AREA that response agency to requires firefighting extinguish.

support by an offsite fire response agency to extinguish.

Basis revised to include NFPA-805 in the discussion of Appendix R basis for the EAL thresholds.

Intent and meaning of the EALs are not altered. 52 DAEC DEVIATIONS AND DIFFERENCES MATRIX 1 ,

Section . l*

1,fEI 99.:01 Re~.'* 6 . . \ ' DAEC ' I Change I . ;Justification

',,> J Validatidn

j Recognition Category:

HU7 HU7 Verbatim Global Comment #11, 14 None Initiating Condition:

Other Other conditions exist which in Difference NOUE versus (NO)UE, DAEC uses the full None conditions exist which in the the judgment of the Emergency NOUE term judgment of the Emergency Director warrant declaration of a Director warrant declaration of a NOUE. (NO) UE. Operating Mode Applicability:

All Operating Mode Applicability:

All Verbatim None (1) Other conditions exist (1) Other conditions exist Verbatim Global Comment #3, 12, 14 None which i.n the judgment of which in the judgment of the Emergency Director the Emergency Director indicate that events are in indicate that events are in r,.. progress or have occurred progress or have occurred ::::, which indicate a potential which indicate a potential

c degradation of the level of degradation of the level of safety of the plant or safety of the plant or indicate a security threat indicate a security threat to facility protection has to facility protection has been initiated.

No releases been initiated.

No releases of radioactive material of radioactive material requiring offsite response requiring offsite response or monitoring are or monitoring are expected unless further expected unless further degradation of safety degradation of SAFE1Y systems occurs. SYSTEMS occurs. 53 I DAEC DEVIATIONS AND DIFFERENCES MATRIX , . . Section

1 ** * :*,._*NEI 99-01 Rev: *s

Justification

I va.lidatiory

1. 1 Recognition Category:

HAl HAl Verbatim Global Comment #11, 14 None Initiating Condition:

HOSTILE HOSTILE ACTION within the Verbatim None ACTION within the OWNER OWNER CONTROLLED AREA or CONTROLLED AREA or airborne airborne attack threat within 30 attack threat within 30 minutes. minutes. Operating Mode Applicability:

All Operating Mode Applicability:

All Verbatim None (1) A HOSTILE ACTION is (1) A HOSTILE ACTION is Difference Global Comment #9, 12, 14 None occurring or has occurred occurring or has occurred .-I within the OWNER within the OWNER <( CONTROLLED AREA as CONTROLLED AREA as :::c reported by the (site-reported by the DAEC specific security shift Security Shift Supervision.

supervision).

(2) A validated notification (2) A validated notification Verbatim from NRC of an aircraft from NRC of an aircraft attack threat within 30 attack threat within 30 minutes of the site. minutes of the site. Intent and meaning of the EALs are not altered. 54 DAEC DEVIATIONS AND DIFFERENCES MATRIX j: Section. * 'f' *1 NEI 99~01*Re(it

6. j , DAEC. ,j Chan~e j ** Justification I Validatiph#'I Recognition Category:

HAS Not used at DAEC Difference EALs RA3 and HAS are not applicable to V8 DAEC because an evaluation has shown that there are no 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.

All areas outside the Control Room that contain equipment necessary for normal plant operation, cooldown and shutdown do not require physical access to operate. Initiating Condition:

Gaseous Not used at DAEC Difference None release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.

Operating Mode Applicability:

All Not used at DAEC Difference None (1) a. Release of a toxic, Not used at DAEC Difference None corrosive, asphyxiant or flammable gas into any of the following plant rooms 11'1 or areas: <( :::c (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. 55 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 DAEC

Change Justification Validation

Recognition Category:

HA6 HAS Difference Renumbered to align with other similar ICs None Initiating Condition:

Control Control Room evacuation Verbatim None Room evacuation resulting in resulting in transfer of plant transfer of plant control to control to alternate locations.

alternate locations.

Operating Mode Applicability:

All Operating Mode Applicability:

All Verbatim None ID (1) An event has resulted in (1) An event has resulted in plant Difference Global Comment #9 & 12 V30 <( :::c plant control being control being transferred transferred from the from the Control Room to the Control Room to (site-Remote Shutdown Panel specific remote shutdown (1C388}. panels and local control stations).

Intent and meaning of the EALs are not altered. 56 DAEC DEVIATIONS AND DIFFERENCES MATRIX * ,: NEI .99:.iOl Rev:: 6

I DAEC 'I Change I*

Justification . I Validation

I Recognition Category:

HA7 HAG Difference Global Comment #11, 14 None Renumbered to align with other similar ICs Initiating Condition:

Other Other conditions exist which in Verbatim None conditions exist which in the the judgment of the Emergency judgment of the Emergency Director warrant declaration of Director warrant declaration of an Alert. an Alert. Operating Mode Applicability:

All Operating Mode Applicability:

All Verbatim None (1) Other conditions exist (1) Other conditions exist which, Verbatim Global Comment #12 None which, in the judgment of in the judgment of the the Emergency Director, Emergency Director, indicate indicate that events are in that events are in progress or progress or have occurred have occurred which involve which involve an actual or an actual or potential " potential substantial substantial degradation of c::t degradation of the level of the level of safety of the ::i::: safety of the plant or a plant or a security event that security event that involves probable life involves probable life threatening risk to site threatening risk to site personnel or damage to site personnel or damage to equipment because of site equipment because of HOSTILE ACTION. Any HOSTILE ACTION. Any releases are expected to be releases are expected to limited to small fractions of be limited to small the EPA Protective Action fractions of the EPA Guideline exposure levels. Protective Action Guideline exposure levels. Intent and meaning of the EALs are not altered 57 DAEC DEVIATIONS AND DIFFERENCES MATRIX NEI 99;;01 Rev~ *6 DAEG I*

Change

I Justification I ValidatiQn

I Recognition Category:

HSl HSl Verbatim Global Comment #11, 14 None Initiating Condition:

HOSTILE HOSTILE ACTION within the Verbatim None ACTION within the PROTECTED PROTECTED AREA. AREA. Operating Mode Applicability:

All Operating Mode Applicability:

All Verbatim None (1) A HOSTILE ACTION is (1) A HOSTILE ACTION is Difference Global Comment #9 & 12 None occurring or has occurred occurring or has occurred II) within the PROTECTED within the PROTECTED AREA :::c AREA as reported by the as reported by the DAEC (site-specific security shift Security Shift Supervision.

supervision).

Intent and meaning of the EALs are not altered. 58 DAEC DEVIATIONS AND DIFFERENCES MATRIX NEI 99.:01 Rev.' 6 :I : DAEC" I .: Chang~ I : . Justification I Validation

I Recognition Category:

HS6 HSS Difference Renumbered to align with other similar ICs None Initiating Condition:

Inability to Inability to control a key safety Verbatim None control a key safety function function from outside the Control from outside the Control Room. Room. Operating Mode Applicability:

All Operating Mode Applicability:

All Verbatim None Note: The Emergency Director Note: The Emergency Director Global Comment #9 V30 should declare the Site Area should declare the Site Area Emergency promptly upon Emergency promptly upon determining that (site specific determining that 20 minutes has number of) minutes has been been exceeded, or will likely be exceeded, or will likely be exceeded.

exceeded.

(1) a. An event has resulted in (1) a. An event has resulted in Difference Global Comment #9, 12 None plant control being plant control being transferred from the transferred from the ID Control Room to (site-Control Room to the V) :I: specific remote shutdown Remote Shutdown Panel panels and local control (1C388). stations).

AND AND Difference Global Comment #4, 9 V30. b. Control of ANY of the b. Control of ANY of the following key safety following key safety functions is not functions is not reestablished within (site-reestablished within 20 specific number of minutes. minutes).

Reactivity control

Reactivity control

RPV water level

Core cooling [PWR] /

RCS heat removal RPV water level [BWR]

RCS heat removal Intent and meaning ofthe EALs are not altered. 59 DAEC DEVIATIONS AND DIFFERENCES MATRIX I ..

Section 1 *

Jcistification I Validation

I Recognition Category:

HS7 Recognition Category:

HS6 Difference Global Comment #11, 14 None Renumbered to align with other similar ICs Initiating Condition:

Other Initiating Condition:

Other Verbatim None conditions exist which in the conditions exist which in the judgment of the Emergency judgment of the Emergency Director warrant declaration of a Director warrant declaration of a Site Area Emergency.

Site Area Emergency.

Operating Mode Applicability:

All Operating Mode Applicability:

Verbatim None ALL (1) Other conditions exist (1) Other conditions exist which Verbatim Global Comment #12 None which in the judgment of in the judgment of the the Emergency Director Emergency Director indicate indicate that events are in that events are in progress or progress or have occurred have occurred which involve which involve actual or actual or likely major failures likely major failures of of plant functions needed for plant functions needed for protection of the public or .... protection of the public or HOSTILE ACTION that results V) :::c: HOSTILE ACTION that in intentional damage or results in intentional malicious acts, (1) toward site damage or malicious acts, personnel or equipment that (1) toward site personnel could lead to the likely failure or equipment that could of or, (2) that prevent lead to the likely failure of effective access to equipment or, (2) that prevent needed for the protection of effective access to the public. Any releases are equipment needed for the not expected to result in protection of the public. exposure levels which exceed Any releases are not EPA Protective Action expected to result in Guideline exposure levels exposure levels which beyond the site boundary.

exceed EPA Protective Action Guideline exposure Intent and meaning of the EALs are not levels beyond the site altered boundary.

60 DAEC DEVIATIONS AND DIFFERENCES MATRIX .DAEC . I*** Change

Jtistification

1

Va)idation n 1 Recognition Category:

HGl HGl Verbatim Global Comment #11, 14 None Initiating Condition:

HOSTILE HOSTILE ACTION resulting in loss Verbatim None ACTION resulting in loss of of physical control of the facility.

physical control of the facility.

Operating Mode Applicability:

All Operating Mode Applicability:

All Verbatim None (1) a. A HOSTILE ACTION is (1) a. A HOSTILE ACTION is Difference Global Comment #9, 12 None occurring or has occurred occurring or has occurred within the PROTECTED within the PROTECTED AREA as reported by the AREA as reported by the (site-specific security shift DAEC Security Shift supervision).

Supervision.

AND AND Difference Global Comment #4, 9 None b. EITHER of the following

b. EITHER of the following

.-i has occurred:

has occurred: :::c 1. ANY of the following

1. ANY of the following safety functions cannot safety functions cannot be controlled or be controlled or maintained.

maintained.

Reactivity control

Reactivity control

Core cooling [PWR] /

RPV water level RPV water level

RCS heat removal [BWR]

RCS heat removal OR OR Verbatim None 2. Damage to spent fuel 2. Damage to spent fuel has occurred or is has occurred or is IMMINENT.

IMMINENT.

Intent and meaning of the EALs are not altered. 61 DAEC DEVIATIONS AND DIFFERENCES MATRIX I* Section ,, h*: NEI 99-01 Rev. 6 . Justification

.1 Validation

1 Recognition Category:

HG7 HG6 Difference Global Comment #11, 14 None Renumbered to align with other similar !Cs Initiating Condition:

Other Other conditions exist which in Verbatim None conditions exist which in the the judgment of the Emergency judgment of the Emergency Director warrant declaration of a Director warrant declaration of a General Emergency.

General Emergency.

Operating Mode Applicability:

All Operating Mode Applicability:

All Verbatim None (1) Other conditions exist (1) Other conditions exist which Verbatim Global Comment #12 None which in the judgment of in the judgment of the the Emergency Director Emergency Director indicate indicate that events are in that events are in progress or progress or have occurred have occurred which involve which involve actual or actual or IMMINENT " IMMINENT substantial substantial core degradation C, core degradation or or melting with potential for :I: melting with potential for loss of containment integrity loss of containment or HOSTILE ACTION that integrity or HOSTILE resu Its in an actual loss of ACTION that results in an physical control of the actual loss of physical facility.

Releases can be control of the facility.

reasonably expected to Releases can be reasonably exceed EPA Protective Action expected to exceed EPA Guideline exposure levels Protective Action Guideline offsite for more than the exposure levels offsite for immediate site area. more than the immediate site area. Intent and meaning of the EALs are not altered 62 DAEC DEVIATIONS AND DIFFERENCES MATRIX SYSTEM MALFUNCTION ICS/EALS 63 DAEC DEVIATIONS AND DIFFERENCES MATRIX ** * *

I DAEC *1 * . Change I . Justification I

Validatio*1;1

I Recognition Category:

SUl SUl Verbatim None Initiating Condition:

Loss of all Loss of ALL offsite AC power Difference Global Comment #15 None offsite AC power capability to capability to essential buses for emergency buses for 15 minutes 15 minutes or longer. or longer. Operating Mode Applicability:

Operating Mode Applicability:

1, Difference Global Comment #10 None .... ::::, Power Operation, Startup, Hot 2, 3 V) Standby, Hot Shutdown (1) Loss of ALL offsite AC (1) loss of ALL offsite AC power Difference Global Comment #9 & 12 None power capability to (site-capability to 1A3 AND 1A4 specific emergency buses) buses for 15 minutes or for 15 minutes or longer. longer. Intent and meaning of the EALs are not altered. 64 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 . DAEC : _-, Change J~stification j Validation

j Recognition Category:

SU2 SU3 Difference Global Comment #11, 14 None Renumbered to align with other similar ICs Initiating Condition:

UNPLANNED UNPLANNED loss of Control Verbatim None loss of Control Room indications Room indications for 15 minutes for 15 minutes or longer. or longer. Operating Mode Applicability:

Operating Mode Applicability:

1, Difference Global Comment #10 None Power Operation, Startup, Hot 2, 3 Standby, Hot Shutdown (1) a. An UNPLANNED event (1) a. An UNPLANNED event Difference Global Comment #12 None results in the inability to results in the inability to monitor one or more of monitor one or more of the following parameters the Table S-1 parameters from within the Control from within the Control Room for 15 minutes or Room for 15 minutes or longer. longer. N Table S-1 Safety System Difference Global Comment #4, 9 None :::) V'l [BWR parameter

[PWR Parameters list] parameter list]

Reactor Power Reactor Power Reactor Power RPV Water Level RCS Level

RPV Water Level RPV Pressure RCS Pressure

RPV Pressure Primary In-Core/Core

Primary Containment Containment Exit Pressure Pressure Temperature Suppression Pool Level Suppression Pool Levels in at least

Level {site-specific

Suppression Pool number) two Temperature steam generators Suppression Pool Steam Temperature Generator Auxiliary or Emergency Feed Water Flow Intent and meaning of the EALs are not altered. 65

-~~-----------------~

DAEC DEVIATIONS AND DIFFERENCES MATRIX Section

1 NEI 99-01 Rev .. 6 DAEC j: Change Justification I Validation

j Recognition Category:

SU3 SU4 Verbatim Global Comment #11, 14R None Renumbered IC to align with other similar ICs Initiating Condition:

Reactor Reactor coolant activity greater Verbatim None coolant activity greater than than Technical Specification Technical Specification allowable allowable limits. limits. Operating Mode Applicability:

Operating Mode Applicability:

1, Difference Global Comment #10 None Power Operation, Startup, Hot 2,3 Standby, Hot Shutdown (1) (Site-specific radiation (1) Pretreatment Offgas System Difference Global Comment #9 & 12 None m :::::, monitor) reading greater (RM-4104)

Hi-Hi Radiation V) than (site-specific value). Alarm (2) Sample analysis indicates (2) Sample analysis Difference Global Comment #9 V31 that a reactor coolant indicates that reactor activity value is greater coolant specific than an allowable limit activity is greater specified in Technical than 2.0 µCi/gm dose Specifications.

equivalent 1-131 for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or longer. Intent and meaning of the EALs are not altered. 66 DAEC DEVIATIONS AND DIFFERENCES MATRIX

NEI 9~t-Ol Rev:* 6 . 'Justification 1 *va.lidatio,:1:#

I Recognition Category:

SU4 SUS Verbatim Global Comment #11, 14 None Renumbered to align with other similar I Cs Initiating Condition:

RCS leakage RCS leakage for 15 minutes or Verbatim None for 15 minutes or longer. longer. Operatirm Mode Applicability:

Operating Mode Applicability:

1, Difference Global Comment #10 None Power Operation, Startup, Hot 2, 3 Standby, Hot Shutdown (1) RCS unidentified or (1) RCS unidentified or pressure Difference Global Comment #9 & 12 V32 pressure boundary leakage boundary leakage greater greater than (site-specific than 10 gpm for 15 minutes value) for 15 minutes or or longer. longer. ,;I" (2) RCS identified leakage (2) RCS identified leakage greater Difference Global Comment #9 V32 ::::, V'I greater than (site-specific than 25 gpm for 15 minutes value) for 15 minutes or or longer. longer. (3) Leakage from the RCS to a (3) Leakage from the RCS to a Verbatim None location outside location outside containment containment greater than greater than 25 gpm for 15 25 gpm for 15 minutes or minutes or longer. longer. Intent and meaning of the EALs are not altered. 67 DAEC DEVIATIONS AND DIFFERENCES MATRIX j:\ .

Section . :j ' I . * * * . DAEC: * * . j ** Change ..

Justification , . Validation

I Recognition Category:

SUS SU6 Verbatim Global Comment #11, 14 None Renumbered to align with other similar ICs Initiating Condition:

Automatic or Automatic or manual scram fails Difference Global Comment #4 None manual (trip [PWR] / scram to shutdown the reactor. [BWR]) fails to shutdown the reactor. Operating Mode Applicability:

Operating Mode Applicability:

1, Difference Global Comment #10 V33 Power Operation 2 DAEC can be up to 12% power in STARTUP Mode, so Mode 2 applicability added (1) a. An automatic (trip [PWR] / (1) a. An automatic scram did Difference Global Comment #4 & 12 None scram [BWR]) did not not shutdown the reactor. r.n shutdown the reactor. V) AND AND Difference Global Comment #9 None b. A subsequent manual b. ANY of the following manual I action taken at the reactor actions taken at lCOS are control consoles is successful in lowering reactor successful in shutting down power below 5% power the reactor.

Manual Scram Pushbuttons

Mode Switch to Shutdown

Alternate Rod Insertion (ARI) 68 DAEC DEVIATIONS AND DIFFERENCES MATRIX I' Section * *j 'NEI 99.::01 Rev.;: 6 . DAEC'. j Change I : j ustificati.on I Vc1li~ation

~

I (2) a. A manual trip ([PWR] / (2) a. A manual scram did not Difference Global Comment #4 None scram [BWR]) did not shutdown the reactor. shutdown the reactor. AND AND None b. EITHER of the following:

b. 1. EITHER of the following Difference Global Comment #9 1. A subsequent manual subsequent manual actions action taken at the taken at lCOS are successful reactor control consoles in lowering reactor power is successful in shutting below 5% power -down the reactor.

Manual Scram Pushbuttons

...i C

Mode Switch to Shutdown 0

Alternate Rod Insertion in :::> (ARI) V) OR OR Difference Global Comment #4 None 2. A subsequent automatic

2. A subsequent automatic (trip [PWR] / scram scram is successful in [BWR]) is successful in shutting down the reactor. shutting down the reactor. Intent and meaning of the EALs are not altered. 69 DAEC DEVIATIONS AND DIFFERENCES MATRIX I: * * *. Sec;tion * 'j':*** * * : *NEI 99-"0l Rev:* 6 .

I . DAEC. *I Chan~e . Justification I V~lidatiorr#

I Recognition Category:

SU6 SU7 Verbatim Global Comment #14 None Renumbered to align with other similar ICs Initiating Condition:

Loss of all Loss of ALL onsite or offsite Difference Global Comment #13 None onsite or offsite communications communications capabilities.

capabilities.

Operating Mode Applicability:

1, Difference Global Comment #10 None Power Operation, Startup, Hot 2, 3 Standby, Hot Shutdown (1) Loss of ALL of the following (1) Loss of ALL of the following Difference Global Comment #9, 12 & 13 V16 Onsite communication Onsite communication methods: methods: (site-specific list of

Plant Operations Radio communications methods) System

In-Plant Phone System ID

Plant Paging System :) (Gaitronics)

V) (2) Loss of ALL of the following (2) Loss of ALL of the following Difference Global Comment #9 & 13 V13, V14 ORO communications offsite response organization methods: communications methods: (site-specific list of

DAEC All-Call phone communications methods)

All telephone lines (PBX and commercial)

Cell Phones (including fixed cell phone system)

Control Room fixed satellite phone system

FTS Phone system 70 DAEC DEVIATIONS AND DIFFERENCES MATRIX :<, : Justification " (3) Loss of ALL of the following (4) Loss of ALL of the following Difference Global Comment #9 & 13 V13, V14 NRC communications NRC communications methods: methods: -(site-specific list of +'

FTS Phone system C: communications methods) All telephone lines (PBX 0

and commercial)

\D ::::>

Cell Phones (including V'l fixed cell phone system)

Control Room fixed satellite phone system Intent and meaning ofthe EALs are not altered. 71 DAEC DEVIATIONS AND DIFFERENCES MATRIX .J.ustification

! Validation

I Recognition Category:

SU7 Not Applicable Difference Global Comment #4 None This IC and EALs are only applicable to PWR plants. Initiating Condition:

Failure to isolate containment or loss of containment pressure control. [PWR] Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown (1) a. Failure of containment to isolate when required by an actuation signal. AND ..... b. ALL required penetrations

> are not closed within 15 Vl minutes of the actuation signal. (1) 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. 72 DAEC DEVIATIONS AND DIFFERENCES MATRIX I * .. ::s : DA.~c: l Change 'I

Justification I V~lidatiqn

I Recognition Category:

SAl SAl Verbatim Global Comment #11, 14 None Initiating Condition:

Loss of all Loss of ALL but one AC power Difference Global Comment #15 None but one AC power source to source to essential buses for 15 emergency buses for 15 minutes minutes or longer. or longer. Operating Mode Applicability:

Operating Mode Applicability:

1, Difference Global Comment #10 None Power Operation, Startup, Hot 2,3 Standby, Hot Shutdown (1) a. AC power capability to (1) a. AC power capability to Difference Global Comment #9, 12 None ....i (site-specific emergency 1A3 and 1A4 buses is <( V) buses) is reduced to a reduced to a single single power source for 15 power source for 15 minutes or longer. minutes or longer. AND AND Difference Global Comment #13 None b. Any additional single a. ANY additional single power source failure will power source failure will result in a loss of all AC result in a loss of ALL AC power to SAFETY SYSTEMS. power to SAFETY SYSTEMS. Intent and meaning of the EALs are not altered. 73 DAEC DEVIATIONS AND DIFFERENCES MATRIX .. NEI 99~01 Rev: .. 6 J *> DAEC. . J Change Justification I V~lidation:#

I Recognition Category:

SA2 SA3 Difference Global Comment #11, 14 None Renumbered to align with other similar ICs Initiating Condition:

UNPLANNED UNPLANNED loss of Control Verbatim None loss of Control Room indications Room indications for 15 minutes for 15 minutes or longer with a or longer with a significant significant transient in progress.

transient in progress.

Operating Mode Applicability:

1, Difference Global Comment #10 None Power Operation, Startup, Hot 2, 3 Standby, Hot Shutdown (1) a. An UNPLANNED event (1) a. An UNPLANNED event Verbatim Global Comment #12 None results in the inability to results in the inability to monitor one or more of monitor one or more Table the following parameters S-1 parameters from from within the Control within the Control Room Room for 15 minutes or for 15 minutes or longer longer. N [BWR [PWR parameter Table S-1 Safety System Difference Global Comment #4, 8 None c:( V) parameter list] list] Parameters Reactor Power Reactor Power Reactor Power

RPVWater RCS Level Level

RPV Water Level RPV Pressure RCS Pressure

RPV Pressure Primary In-Core/Core Exit

Primary Containment Containment Temperature Pressure Pressure Suppression Levels in at least

Suppression Pool Level Pool Level (site-specific

Suppression Pool number) steam Temperature generators Suppression Steam Generator Pool Auxiliary or Temperature Emergency Feed Water Flow AND AND 74 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rey. 6

DAEC Change I. Justification I Validation

I b. ANY of the following

b. Any of the Table 5-2 Difference Global Comment #4, 9 None transient events in transient events are in *-progress.

progress

Automatic or manual run back greater than Table S-2 Significant 25% thermal reactor Transients

-power

Automatic or manual ...: C:

Electrical load rejection run back greater than 25% 0 greater than 25% full thermal reactor power N C( electrical load

Electrical load rejection V)

Reactor scram [BWR] / greater than 25% full trip [PWR] electrical load

ECCS (SI) actuation

Reactor scram

Thermal power

ECCS actuation oscillations greater than

Thermal power oscillations 10% [BWR] greater than 10% Intent and meaning of the EALs are not altered. 75 DAEC DEVIATIONS AND DIFFERENCES MATRIX NEI 99.:01 Rev. 6 Jostification . I Validatici;n

I Recognition Category:

SAS SA6 Difference Global Comment #11, 14 None Renumbered to align with other similar I Cs Initiating Condition:

Automatic or Automatic or manual scram fails Difference Global Comment #4 & 9 None manual (trip [PWR] / scram to shutdown the reactor, and [BWR]) fails to shutdown the subsequent manual actions taken reactor, and subsequent manual at the reactor control consoles actions taken at the reactor are not successful in shutting control consoles are not down the reactor. successful in shutting down the reactor. Operating Mode Applicability:

Operating Mode Applicability:

1, Difference Global Comment #10 V33 Power Operation 2 DAEC can be up to 12% power in STARTUP Mode, so Mode 2 applicability added (1) a. An automatic or manual (1) a. An automatic or manual Difference Global Comment #4, 9 & 12 None (trip [PWR] / scram [BWR]) scram did not shutdown in did not shutdown the the reactor. <C V'I reactor. AND AND Difference Global Comment #9 None b. Manual actions taken at b. ALL of the following the reactor control manual actions taken at consoles are not successful lCOS are not successful in shutting down the in lowering reactor reactor. power below 5% power

Manual Scram Pushbuttons

Mode Switch to Shutdown

Alternate Rod Insertion (ARI) Intent and meaning of the EALs are not altered. L 76

~~--------

-DAEC DEVIATIONS AND DIFFERENCES MATRIX .'*1N,EI 99.;oa Rev, 6 ' . Justification I* Validation#

I Recognition Category:

SA9 SA8 Difference Global Comment #11, 14 None Renumbered to align with other similar ICs Initiating Condition:

Hazardous Hazardous event affecting a Verbatim None event affecting a SAFETY SYSTEM SAFETY SYSTEM needed for the needed for the current operating current operating mode. mode. Operating Mode Applicability:

Operating Mode Applicability:

1, Difference Global Comment #10 None Power Operation, Startup, Hot 2,3 Standby, Hot Shutdown (1) a. The occurrence of ANY of (1) a. The occurrence of ANY of Difference Global Comment #12 & 13 None the following hazardous the Table S-3 hazardous events: events:

Seismic event Table S-3 Hazardous Events Difference Global Comment #8 & 9 None en (earthquake)

Seismic event (earthquake)

<( V,

Internal or external

Internal or external flooding flooding event event

High winds or tornado

High winds or tornado strike strike

FIRE

FIRE

EXPLOSION

EXPLOSION

Other events with similar * (site-specific hazards) hazard characteristics as

Other events with determined by the Shift similar hazard Manager or Emergency characteristics as Director determined by the Shift Manager 77 DAEC DEVIATIONS AND DIFFERENCES MATRIX I' . ' Section . ,y . . :'NEI 99 .. 01 Rev: 6 : Justification I Validation

1 AND AND b. EITHER of the following:

b. 1. Event damage has Deviation Adopted the revised EAL structure and V17 1. Event damage has caused indications of wording provided in approved EAL FAQ caused indications of degraded 2016-02. degraded performance performance in one in at least one train of a train of a SAFETY SAFETY SYSTEM needed SYSTEM needed for for the current the current operating operating mode. mode. AND OR 2. EITHER of the following:

Deviation Adopted the revised EAL wording provided V17 2. The event has caused

Event damage has in approved EAL FAQ 2016-02 VISIBLE DAMAGE to a caused indications SAFETY SYSTEM of degraded component or structure performance to a Difference Added the following clarification to the V18 -needed for the current second train of the Basis from EALFAQ 2018-04: ....: operating mode. SAFETY SYSTEM An event affecting a single-train SAFETY C: 0 needed for the SYSTEM (i.e., there are indications of en current operating degraded performance and/or VISIBLE <( II') mode, or DAMAGE affecting the one train) would not

The event has be classified under SA8 because the two-resulted in VISIBLE train impact criteria that underlie the EALs DAMAGE to the and Bases would not be met. If an event second train of a affects a single-train SAFETY SYSTEM, then SAFETY SYSTEM the emergency classification should be needed for the made based on plant current operating parameters/symptoms meeting the EALs mode. for another IC. Depending upon the circumstances, classification may also occur based on Shift Manager/Emergency Director judgement.

Intent and meaning of the EALs are not altered. 78 L_ __

DAEC DEVIATIONS AND DIFFERENCES MATRIX Sed:fon

DJfEC " . Change . Justific.ation . Validation

. Recognition Category:

551 551 Verbatim Global Comment #11, 14 None Initiating Condition:

Loss of all Loss of ALL offsite and ALL onsite Difference Global Comment #13, 15 None offsite and all onsite AC power to AC power to essential buses for emergency buses for 15 minutes 15 minutes or longer. or longer. Operating Mode Applicability:

Operating Mode Applicability:

1, Difference Global Comment #10 None .... II) Power Operation, Startup, Hot 2, 3 II) Standby, Hot Shutdown (1) Loss of ALL offsite and ALL (1) Loss of ALL offsite and ALL Difference Global Comment #9, 12 & 13 None onsite AC power to (site-onsite AC power to 1A3 and specific emergency buses) 1A4 buses for 15 minutes or for 15 minutes or longer. longer. Intent and meaning of the EALs are not altered. 79 DAEC DEVIATIONS AND DIFFERENCES MATRIX * :NEI 99~.01 Rev. 6 I* DAEC *I* Change j \ Justificati.on I Validation

I Recognition Category:

SSS SS6 Difference Global Comment #11, 14 None Renumbered to align with other similar ICs Initiating Condition:

Inability to Inability to shutdown the reactor Difference Global Comment #4 None shutdown the reactor causing a causing a challenge to RPV water challenge to (core cooling [PWR] level or RCS heat removal. / RPV water level [BWR]) or RCS heat removal. Operating Mode Applicability:

Operating Mode Applicability:

1, Difference Global Comment #10 V33 Power Operation 2 DAEC can be up to 12% power in STARTUP Mode, so Mode 2 applicability added (1) a. An automatic or manual (1) a. An automatic or manual Difference Global Comment #4, 9 & 12 None (trip [PWR] / scram [BWR]) scram did not shutdown did not shutdown the the reactor. reactor. AND AND Verbatim None 11'1 b. All manual actions to b. All manual actions to Ill Ill shutdown the reactor have shutdown the reactor been unsuccessful.

have been unsuccessful.

AND AND Difference Global Comment #9 V34 c. EITHER of the following

c. EITHER of the following V27 conditions exist: conditions exist: * (Site-specific indication

RPV level cannot be of an inability to restored and maintained adequately remove heat above -25 inches. from the core) OR * (Site-specific indication

HCL (Graph 4 of EOP 2) of an inability to exceeded.

adequately remove heat from the RCS) Intent and meaning of the EALs are not altered. 80 DAEC DEVIATIONS AND DIFFERENCES MATRIX j: *

S~ction . *

j . ' ,; NEI 99;;01 Rev .. 6 DAEC :* J* Change -I'* Justification

J Validation

I Recognition Category:

558 552 Difference Global Comment #11, 14 None Renumbered to align with other similar ICs Initiating Condition:

Loss of all Loss of ALL Vital DC power for 15 Difference Global Comment #13 None Vital DC power for 15 minutes or minutes or longer. longer. Operating Mode Applicability:

Operating Mode Applicability:

1, Difference Global Comment #10 None Power Operation, Startup, Hot 2,3 00 Standby, Hot Shutdown V) V) (1) Indicated voltage is less (1) Indicated voltage is less Difference Global Comment #9 & 12 V12 than (site-specific bus than 105 VDC on BOTH voltage value) on ALL (site-Div 1 and Div 2 125 VDC specific Vital DC busses) for buses for 15 minutes or 15 minutes or longer. longer. Intent and meaning of the EALs are not altered. 81 DAEC DEVIATIONS AND DIFFERENCES MATRIX Se'ct:ion

.:'NEI 99 .. 01 Rev/ 6 *

DAEC ** Chang~ .Justification*

Validation#

Recognition Category:

SGl SGl Verbatim Global Comment #11, 14 None Initiating Condition:

Prolonged Prolonged loss of ALL offsite and Difference Global Comment #13, 15 None loss of all offsite and all onsite AC ALL onsite AC power to essential power to emergency buses. buses. Operating Mode Applicability:

Operating Mode Applicability:

1, Difference Global Comment #10 None Power Operation, Startup, Hot 2,3 Standby, Hot Shutdown (1) a. Loss of ALL offsite and ALL (1) a. Loss of ALL offsite and ALL Difference Global Comment #9 & 13 None onsite AC power to (site-onsite AC power to 1A3 specific emergency buses). and 1A4 buses .... C, AND AND Difference Global Comment #9 & 13 Ill b. EITHER of the following:

b. EITHER of the following:

Restoration of at least

Restoration of at least one AC emergency bus one AC essential bus in in less than (site-specific less than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is not hours) is not likely. likely. OR * (Site-specific indication

RPV level cannot be Difference Global Comment #9 V34 of an inability to restored and maintained adequately remove heat above from the core) -25 inches. Intent and meaning of the EALs are not altered. 82 DAEC DEVIATIONS AND DIFFERENCES MATRIX DAEc*. t

Change . Justification

.. Recognition Category:

SGS SG2 Difference Global Comment #11, 14 None Renumbered to align with other similar I Cs Initiating Condition:

Loss of all AC Loss of ALL AC and Vital DC Verbatim Global Comment #13 None and Vital DC power sources for power sources for 15 minutes or 15 minutes or longer. longer. Operating Mode Applicability:

Operating Mode Applicability:

1, Difference Global Comment #10 None Power Operation, Startup, Hot 2, 3 Standby, Hot Shutdown (1) a. Loss of ALL offsite and ALL (1) a. Loss of ALL offsite and ALL Difference Global Comment #9, 12, 13 None 00 onsite AC power to (site-onsite AC power to 1A3 l!1 V'l specific emergency buses} and 1A4 buses for 15 for 15 minutes or longer. minutes or longer. AND AND Difference Global Comment #9 & 13 V12 b. Indicated voltage is less b. Indicated voltage is less than (site-specific bus than 105 VDC on BOTH Div voltage value} on ALL (site-1 and Div 2 125 VDC buses specific Vital DC busses} for for 15 minutes or longer. 15 minutes or longer. Intent and meaning of the EALs are not altered. 83

DAEC DEVIATIONS AND DIFFERENCES MATRIX APPENDIX A -ACRONYMS AND ABBREVIATIONS 84 DAEC DEVIATIONS AND DIFFERENCES MATRIX )*. Section

j NEI 99-01 Rev. 6 DAEC Change Justification j Validation

J AC. ...... Alternating Current AC. ...... Alternating Current Verbatim N/A AOP ...... Abnormal Operating AOP ...... Abnormal Operating Verbatim N/A Procedure Procedure APRM ... Average Power Range Difference Not used N/A Meter ATWS ... Anticipated Transient ATWS ... Anticipated Transient Verbatim N/A Without Scram Without Scram B&W .... Babcock and Wilcox Difference Not used N/A BIIT ...... Boron Injection Initiating Difference Not used N/A Temperature VI BWR .... Boiling Water Reactor BWR .... Boiling Water Reactor Verbatim N/A z 0 CDE ...... Committed Dose CDE ...... Committed Dose Verbatim N/A > Equivalent Equivalent w CFR ...... Code of Federal CFR ...... Code of Federal Verbatim N/A c::: cc Regulations Regulations cc <C CTMT/CNMT

... Containment Difference Not used N/A C z CSF ...... Critical Safety Function Difference Not used N/A <C u, CSFST ... Critical Safety Function Difference Not used N/A 2: > Status Tree z DBA ...... Design Basis Accident Difference Not used N/A 0 c::: N/A u DC. ....... Direct Current DC. ....... Direct Current Verbatim <C I EAL. ...... Emergency Action Level EAL... .... Emergency Action Level Verbatim N/A <C ECCS .... Emergency Core Cooling ECCS .... Emergency Core Cooling Verbatim N/A X 15 System System z ECL... .... Emergency Classification ECL. ...... Emergency Classification Verbatim N/A w c.. c.. Level Level <C EOF ...... Emergency Operations EOF ...... Emergency Operations Verbatim N/A Facility Facility EOP ...... Emergency Operating EOP ...... Emergency Operating Verbatim N/A Procedure Procedure EPA ...... Environmental Protection EPA ...... Environmental Protection Verbatim N/A Agency Agency EPG ..... Emergency Procedure EPG ..... Emergency Procedure Verbatim N/A Guideline Guideline EPIP ..... Emergency Planning Difference Not used N/A Implementing Procedure 85 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section , , NEI 99-01 Rev. 6 DAEC

j Change 'Justification

' J Validation

j EPR ...... Evolutionary Power Difference Not used N/A Reactor EPRI ..... Electric Power Research Difference Not used N/A Institute ERG ..... Emergency Response Difference Not used N/A Guideline FEMA ... Federal Emergency FEMA ... Federal Emergency Verbatim N/A Management Agency Management Agency FSAR .... Final Safety Analysis Difference Not used N/A -Report ...; C: GE ........ General Emergency GE ........ General Emergency Verbatim N/A 0 .!:. HCTL .... Heat Capacity HCL. ... Heat Capacity Limit Difference Updated to reflect DAEC EOPs N/A Ill z Temperature Limit 0 HPCI ..... High Pressure Coolant HPCI ..... High Pressure Coolant Verbatim N/A > Injection Injection w HSI.. ...... Human System Interface Difference Not used N/A 0:: co ca IC ..........

lnitiating Condition IC ..........

lnitiating Condition Verbatim N/A <C C ID .........

lnside Diameter ID .........

lnside Diameter Verbatim N/A z IPEEE ... lndividual Plant Difference Not used N/A <C Ill Examination of External Events > (Generic Letter 88-20) z 0 ISFSl.. .. lndependent Spent Fuel ISFSl.. .. lndependent Spent Fuel Verbatim N/A 0:: u Storage Installation Storage Installation

<C I Keff ..... Effective Neutron Keff ..... Effective Neutron Verbatim N/A <C X Multiplication Factor Multiplication Factor c LCO ..... Limited Condition of LCO ..... Limited Condition of Verbatim N/A z w Operation Operation

0. 0. LOCA ... Loss of Coolant Accident LOCA ... Loss of Coolant Accident Verbatim N/A <C MCR .... Main Control Room Difference Not used N/A MSIV ... Main Steam Isolation Difference Not used N/A Valve MSL.. ... Main Stem Line Difference Not used N/A mR, mRem, mrem, mREM .... milli-mR, mRem, mrem, mREM .... milli-Verbatim N/A Roentgen Equivalent Man Roentgen Equivalent Man MW ..... Megawatt MW ..... Megawatt Verbatim N/A NEI ....... Nuclear Energy Institute NEI.. ..... Nuclear Energy Institute Verbatim N/A NPP ...... Nuclear Power Plant Difference Not used N/A 86 DAEC DEVIATIONS AND DIFFERENCES MATRIX j

Section NEI 99-01 Rev. 6. DAEC J ,

Change Justification J Validation.#

J NRC. .... Nuclear Regulatory NRC. .... Nuclear Regulatory Verbatim N/A Agency Agency NSSS .... Nuclear Steam Supply Difference Not used N/A System NORAD ... North American NORAD ... North American N/A Aerospace Defense Command Aerospace Defense Command (NO)UE ... (Notification of) Unusual NOUE ... Notification of Unusual Difference DAEC uses full NOUE terminology N/A Event Event -...; NUMARC. ... Nuclear Management NUMARC. ... Nuclear Management Verbatim N/A C: 0 and Resources Council and Resources Council Vl OBE ..... Operating Basis OBE ..... Operating Basis Verbatim N/A z 0 Earthquake Earthquake OCA ..... Owner Controlled Area OCA. .... Owner Controlled Area Verbatim N/A > ODCM/ODAM

.... Offsite Dose ODAM ... Offsite Dose Assessment Difference DAEC uses ODAM N/A w a:: cc Calculation (Assessment)

Manual Manual cc <t ORO ..... Offsite Response Difference Not used N/A C z Organization

<t PA .........

Protected Area PA. ........ Protected Area Verbatim N/A Vl 2 PACS .... Priority Information and Difference Not used N/A > z Control System 0 a:: PAG ...... Protective Action PAG ...... Protective Action Verbatim N/A u <t Guideline Guideline I <t PICS ..... Process Information and Difference Not used N/A X Control System c z PRA/PSA ... Probabilistic Risk PRA/PSA ... Probabilistic Risk Verbatim N/A w C. Assessment/Probabilistic Safety Assessment/Probabilistic Safety C. <t Assessment Assessment PWR .... Pressurized Water Reactor PWR .... Pressurized Water Reactor Verbatim N/A PS .........

Protection System Difference Not used N/A PSIG .... Pounds per Square Inch PSIG .... Pounds per Square Inch Verbatim N/A R ..........

Roentgen R ..........

Roentgen Verbatim N/A RCC. ... Reactor Control Console Difference Not used N/A RCIC. .. Reactor Core Isolation RCIC. .. Reactor Core Isolation Verbatim N/A Cooling Cooling 87

~~-

~~~~~~

DAEC DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6

j DAEC j Chai:ige

I Justificatiqn j Validation#

j RCS ..... Reactor Coolant System RCS ..... Reactor Coolant System Verbatim N/A Rem, rem, REM ... Roentgen Rem, rem, REM ... Roentgen Verbatim N/A Equivalent Man Equivalent Man RETS .... Radiological Effluent Difference Not used N/A Technical Specifications RPS ...... Reactor Protection System RPS ...... Reactor Protection System Verbatim N/A -RPV ...... Reactor Pressure Vessel RPV ...... Reactor Pressure Vessel Verbatim N/A ..l RVLIS ... Reactor Vessel Level Difference Not used N/A C: 0 Instrumentation System Vl RWCU ... Reactor Water Cleanup RWCU ... Reactor Water Cleanup Verbatim N/A z 0 SAR ....... Safety Analysis Report Difference Not used N/A SAS ........ Safety Automation Difference Not used N/A > w System a:: cc SBO ....... Station Blackout Difference Not used N/A cc <C SCBA ..... Self-Contained Breathing SCBA ..... Self-Contained Breathing Verbatim N/A C z Apparatus Apparatus

<C SG ..........

Steam Generator Difference Not used N/A Vl SI... ........ Safety Injection Difference Not used N/A > z SICS ...... Safety Information Difference Not used N/A 0 a:: Control System u <C SPDS ..... Safety Parameter Display SPDS ..... Safety Parameter Display Verbatim N/A I <C System System X c SRO ....... Senior Reactor Operator Difference Not used N/A z TEDE ..... Total Effective Dose TEDE ..... Total Effective Dose Verbatim N/A w Cl. Cl. Equivalent Equivalent

<C TOAF ..... Top of Active Fuel TAF ..... Top of Active Fuel Difference Updated to reflect DAEC EOPs N/A TSC.. ...... Technical Support TSC. ....... Technical Support Verbatim N/A System System -UFSAR .... Final Safety Analysis Difference Used in Section 3.1 N/A Report WOG ..... Westinghouse Owners Difference Not used N/A Group 88

DAEC DEVIATIONS AND DIFFERENCES MATRIX APPENDIX B -DEFINITIONS 89 DAEC DEVIATIONS AND DIFFERENCES MATRIX *NEI 99-01 Rev'! 6 QAEC.

1 Change '!'*:' ustific"tion 1 ValidatiQn

1 Alert: Events are in progress or have occurred Alert: Events are in progress or have occurred Verbatim None which involve an actual or potential which involve an actual or potential substantial degradation of the level of safety substantial degradation of the level of safety of the plant or a security event that involves of the plant or a security event that involves probable life threatening risk to site probable life threatening risk to site personnel or damage to site equipment personnel or damage to site equipment because of HOSTILE ACTION. Any releases are because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of expected to be limited to small fractions of the EPA PAG exposure levels. the EPA PAG exposure levels. V, General Emergency:

Events are in progress or General Emergency:

Events are in progress or Verbatim None z have occurred which involve actual or have occurred which involve actual or 0 E IMMINENT substantial core degradation or IMMINENT substantial core degradation or z u::: melting with potential for loss of melting with potential for loss of w containment integrity or HOSTILE ACTION C containment integrity or HOSTILE ACTION I that results in an actual loss of physical that results in an actual loss of physical co X control of the facility.

Releases can be control of the facility.

Releases can be 25 z reasonably expected to exceed EPA PAG reasonably expected to exceed EPA PAG w Q. exposure levels offsite for more than the exposure levels offsite for more than the Q. <t immediate site area. immediate site area. Notification of Unusual Event: Events are in Unusual Event: Events are in progress or have Difference See Global Comment #3 None progress or have occurred which indicate a occurred which indicate a potential potentia I degradation of the level of safety of degradation of the level of safety of the plant the plant or indicate a security threat to or indicate a security threat to facility facility protection has been initiated.

No protection has been initiated.

No releases of releases of radioactive material requiring radioactive material requiring offsite offsite response or monitoring are expected response or monitoring are expected unless unless further degradation of safety systems further degradation of SAFETY SYSTEMS occurs. occurs. 90 DAEC DEVIATIONS AND DIFFERENCES MATRIX .* : "NEI 99~01 Rev:* 6 "* J DAEC ... , I Change J . Justification I Validation:#

I Site Area Emergency:

Events are in progress Site Area Emergency:

Events are in progress Verbatim None or have occurred which involve actual or or have occurred which involve actual or likely major failures of plant functions needed likely major failures of plant functions needed for protection of the public or HOSTILE for protection of the public or HOSTILE ACTION that results in intentional damage or ACTION that results in intentional damage or malicious acts; 1) toward site personnel or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure equipment that could lead to the likely failure of or; 2) that prevent effective access to, of or; 2) that prevent effective access to, equipment needed for the protection of the equipment needed for the protection of the public. Any releases are not expected to public. Any releases are not expected to result in exposure levels which exceed EPA result in exposure levels which exceed EPA Ill PAG exposure levels beyond the site PAG exposure levels beyond the site 2 boundary.

boundary.

0 E Emergency Action Level (EAL): A pre-Emergency Action Level (EAL): A pre-Verbatim None 2 determined, site-specific, observable determined, site-specific, observable i:i: w threshold for an Initiating Condition that, threshold for an Initiating Condition that, C I when met or exceeded, places the plant in a when met or exceeded, places the plant in a cc X given emergency classification level. given emergency classification level. c 2 Emergency Classification Level (ECL): One of a Emergency Classification Level (ECL): One of a Verbatim None w C. set of names or titles established by the US set of names or titles established by the US C. c:( Nuclear Regulatory Commission (NRC) for Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions grouping off-normal events or conditions according to (1) potential or actual effects or according to (1) potential or actual effects or consequences, and (2) resulting onsite and consequences, and (2) resulting onsite and offsite response actions. The emergency offsite response actions. The emergency classification levels, in ascending order of classification levels, in ascending order of severity, are: severity, are:

Notification of Unusual Event (NOUE)

Notification of Unusual Event (NOUE)

Alert

Alert

Site Area Emergency (SAE)

Site Area Emergency (SAE)

General Emergency (GE)

General Emergency (GE) 91 DAEC DEVIATIONS AND DIFFERENCES MATRIX IC NEI 99-01 Rev. 6 DAEC .* j Change I, Justification j Validation

j Fission Product Barrier Threshold:

A pre-Fission Product Barrier Threshold:

A pre-Verbatim None determined, site-specific, observable determined, site-specific, observable threshold indicating the loss or potential loss threshold indicating the loss or potential loss of a fission product barrier. of a fission product barrier. Initiating Condition (IC}: An event or Initiating Condition (IC): An event or Verbatim None condition that aligns with the definition of condition that aligns with the definition of one of the four emergency classification one of the four emergency classification levels by virtue of the potential or actual levels by virtue of the potential or actual effects or consequences.

effects or consequences.

CONFINEMENT BOUNDARY: (Insert a site-CONFINEMENT BOUNDARY:

The barrier(s)

Difference Removed developer notes None V, specific definition for this term.) Developer between spent fuel and the environment and added site-specific z Note -The barrier(s) between spent fuel and once the spent fuel is processed for dry language.

0 i= the environment once the spent fuel is storage. This corresponds to the pressure z LL. processed for dry storage. boundary for the Dry Shielded Canister (DSC) w shell (including the inner bottom cover plate) C I base metal and associated confinement al X boundary welds. c z CONTAINMENT CLOSURE: (Insert a site-CONTAINMENT CLOSURE: Site specific Difference Removed developer notes None w C. specific definition for this term.) Developer procedurally defined actions taken to secure and added existing C. <( Note -The procedurally defined conditions containment and its associated structures, definition from present or actions taken to secure containment systems, and components as a functional EALs. (primary or secondary for BWR) and its barrier to fission product release under associated structures, systems, and existing plant conditions.

For DAEC, this is components as a functional barrier to fission considered to be Secondary Containment as product release under shutdown conditions.

required by Technical Specifications.

DESIGN BASIS EARTHQUAKE (DBE): A DBE is Difference Added term used in HU2 None vibratory ground motion for which certain versus use of footnotes (generally, safety-related) structures, systems, and components must be designed to remain functional.

92 DAEC DEVIATIONS AND DIFFERENCES MATRIX '.justification

' I Validati?r:t, # I EXPLOSION:

A rapid, violent and catastrophic EXPLOSION:

A rapid, violent, and catastrophic Verbatim None failure of a piece of equipment due to failure of a piece of equipment due to combustion, chemical reaction or combustion, chemical reaction, or overpressurization.

A release of steam (from overpressurization.

A release of steam (from high energy lines or components) or an high energy lines or components) or an electrical component failure (caused by short electrical component failure (caused by short circuits, grounding, arcing, etc.) should not circuits, grounding, arcing, etc.) should not automatically be considered an explosion.

automatically be considered an explosion.

Such events may require a post-event Such events may require a post-event V'l inspection to determine ifthe attributes of an inspection to determine if the attributes of an z explosion are present. explosion are present. 0 E z FAULTED: The term applied to a steam Difference Term not used for BWRs None u: w generator that has a steam leak on the Cl I secondary side of sufficient size to cause an co X unco~trolled drop in steam generator 25 z pressure or the steam generator to become w C. completely de pressurized.

Developer Note -C. <( This term is applicable to PWRs only. FIRE: Combustion characterized by heat and FIRE: Combustion characterized by heat and Verbatim None light. Sources of smoke such as slipping drive light. Sources of smoke such as slipping drive belts or overheated electrical equipment do belts or overheated electrical equipment do not constitute FIRES. Observation offlame is not constitute FIRES. Observation of flame is preferred but is NOT required if large preferred but is NOT required if large quantities of smoke and heat are observed.

quantities of smoke and heat are observed.

HOSTAGE: A person(s) held as leverage HOSTAGE: A person(s) held as leverage Verbatim None against the station to ensure that demands against the station to ensure that demands will be met by the station. will be met by the station. 93 DAEC DEVIATIONS AND DIFFERENCES MATRIX IC I NEI 99-01 Rev:; 6 DAEC Change Justification I Validation

I HOSTILE ACTION: An act toward a NPP or its HOSTILE ACTION: An act toward a nuclear Difference Spelled out 'NPP' in 2 None personnel that includes the use of violent power plant or its personnel that includes the places force to destroy equipment, take HOSTAGES, use of violent force to destroy equipment, and/or intimidate the licensee to achieve an take HOSTAGES, and/or intimidate the end. This includes attack by air, land, or water licensee to achieve an end. This includes using guns, explosives, PROJECTILEs, vehicles, attack by air, land, or water using guns, or other devices used to deliver destructive explosives, PROJECTILEs, vehicles, or other force. Other acts that satisfy the overall devices used to deliver destructive force. intent may be included.

HOSTILE ACTION Other acts that satisfy the overall intent may should not be construed to include acts of be included.

HOSTILE ACTION should not be civil disobedience or felonious acts that are construed to include acts of civil not part of a concerted attack on the NPP. disobedience or felonious acts that are not Non-terrorism-based EALs should be used to part of a concerted attack on the nuclear address such activities (i.e., this may include power plant. Non-terrorism-based EALs violent acts between individuals in the owner should be used to address such activities (i.e., controlled area). this may include violent acts between individuals in the owner controlled area). HOSTILE FORCE: One or more individuals who HOSTILE FORCE: One or more individuals who Verbatim None are engaged in a determined assault, overtly are engaged in a determined assault, overtly or by stealth and deception, equipped with or by stealth and deception, equipped with suitable weapons capable of killing, maiming, suitable weapons capable of killing, maiming, or causing destruction.

or causing destruction.

IMMINENT:

The trajectory of events or IMMINENT:

The trajectory of events or Verbatim None conditions is such that an EAL will be met conditions is such that an EAL will be met within a relatively short period of time within a relatively short period of time regardless of mitigation or corrective actions. regardless of mitigation or corrective actions. INDEPENDENT SPENT FUEL STORAGE INDEPENDENT SPENT FUEL STORAGE Verbatim None INSTALLATION (ISFSI): A complex that is INSTALLATION (ISFSI): A complex that is designed and constructed for the interim designed and constructed for the interim storage of spent nuclear fuel and other storage of spent nuclear fuel and other radioactive materials associated with spent radioactive materials associated with spent fuel storage. fuel storage. 94 DAEC DEVIATIONS AND DIFFERENCES MATRIX ,,.* IC '°'I : tNEI 99 7 01 Rev; 6 *oAEC .. ** *.

j: Change j: * .Justification.

I Validation:#

I NORMAL LEVELS: As applied to radiological Difference Term not used in this EAL None IC/EALs, the highest reading in the past scheme twenty-four hours excluding the current peak value. OPERATING BASIS EARTHQUAKE (OBE}: An Difference Added term used in HU2 None OBE is vibratory ground motion for which versus use of footnotes those features of a nuclear power plant necessary for continued operation without undue risk to the health and safety ofthe public will remain functional.

OWNER CONTROLLED AREA: (Insert a site-OWNER CONTROLLED AREA: The site Difference Definition from developer None specific definition for this term.) Developer property owned by or otherwise under the notes used. Developer 11'1 Note -This term is typically taken to mean control of the licensee.

Notes deleted. z 0 the site property owned by, or otherwise j:: 2 under the control of, the licensee.

In some u: cases, it may be appropriate for a licensee to w Cl define a smaller area with a perimeter closer I cc to the plant Protected Area perimeter (e.g., a X c site with a large OCA where some portions of z the boundary may be a significant distance w 0. 0. from the Protected Area). In these cases, <C developers should consider using the boundary defined by the Restricted or Secured Owner Controlled Area (ROCA/SOCA).

The area and boundary selected for scheme use must be consistent with the description of the same area and boundary contained in the Security Plan. PROJECTILE:

An object directed toward a NPP PROJECTILE:

An object directed toward a Difference Spelled out 'NPP' None that could cause concern for its continued nuclear power plant that could cause concern operability, reliability, or personnel safety. for its continued operability, reliability, or personnel safety. 95 DAEC DEVIATIONS AND DIFFERENCES MATRIX IC NEI 99-01 Rev. 6

1 DAEC Change Justification J .Validation#

J PROTECTED AREA: (Insert a site-specific PROTECTED AREA: The area under Difference Definition from developer None definition for this term.) Developer Note -continuous access monitoring and control, notes used. Developer This term is typically taken to mean the area and armed protection as described in the site Notes deleted. under continuous access monitoring and Security Plan. control, and armed protection as described in the site Security Plan. REFUELING PATHWAY: (Insert a site-specific REFUELING PATHWAY: The reactor refueling Difference DAEC-specific definition None definition for this term.) Developer Note -cavity, spent fuel pool, and fuel transfer supplied.

Developer This description should include all the canal. Notes deleted. cavities, tubes, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel. RUPTURE(D):

The condition of a steam Difference Not used None generator in which primary-to-secondary V, leakage is of sufficient magnitude to require a z safety injection.

Developer Note -This term 0 j::: is applicable to PWRs only. 2 U::: w SAFETY SYSTEM: A system required for safe SAFETY SYSTEM: A system required for safe C Difference Removed developer notes None I r::c plant operation, cooling down the plant plant operation, cooling down the plant and clarified last sentence.

X and/or placing it in the cold shutdown and/or placing it in the cold shutdown 2i z condition, including the ECCS. These are condition, including the ECCS. These systems w C. typically systems classified as safety-related.

are classified as safety-related.

C. <t Developer Note -This term may be modified to include the attributes of "safety-related" in accordance with 10 CFR 50.2 or other site-specific terminology, if desired. SECURITY CONDITION:

Any Security Event as SECURITY CONDITION:

Any Security Event as Verbatim None listed in the approved security contingency listed in the approved security contingency plan that constitutes a threat/compromise to plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or site security, threat/risk to site personnel, or a potential degradation to the level of safety a potential degradation to the level of safety 96 DAEC DEVIATIONS AND DIFFERENCES MATRIX NEI 99-01 Rev .. 6 DAEC Change Justification I Validation

I of the plant. A SECURITY CONDITION does of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION. not involve a HOSTILE ACTION. SITE BOUNDARY:

That line beyond which the Difference Defined term from ODCM None land is neither owned, nor leased, nor needed for several EALs otherwise controlled by the Company. UFSAR Figure 1.2-1 identifies the DAEC SITE BOUNDARY.

UNISOLABLE:

An open or breached system UNISOLABLE:

An open or breached system Verbatim None line that cannot be isolated, remotely or line that cannot be isolated, remotely or locally. locally. UNPLANNED:

A parameter change or an UNPLANNED:

A parameter change or an Verbatim N/A event that is not 1) the result of an intended event that is not 1) the result of an intended evolution or 2) an expected plant response to evolution or 2) an expected plant response to a transient.

The cause of the parameter a transient.

The cause of the parameter V) change or event may be known or unknown. change or event may be known or unknown. z 0 j::: VISIBLE DAMAGE: Damage to a component or VISIBLE DAMAGE: Damage to a component or Deviation Updated to reflect V17 z u: structure that is readily observable without structure that is readily observable without wording and guidance of w C measurements, testing, or analysis.

The visual measurements, testing, or analysis.

The visual approved EAL FAQ 2016-I cc impact of the damage is sufficient to cause impact of the damage is sufficient to cause 02. The updated wording X concern regarding the operability or concern regarding the operability or clarifies damage c z reliability of the affected component or reliability of the affected component or assessment meriting an w c.. structure.

structure.

Damage resulting from an ALERT declaration as used c.. <C equipment failure and limited to the failed in ICs using this definition component (i.e., the failure did not cause (CA6 and SA9). damage to a structure or any other equipment) is not VISIBLE DAMAGE. 97 DAEC DEVIATIONS AND DIFFERENCES MATRIX APPENDIX C -Permanently Defueled ICs/EALs DAEC DEVIATIONS AND DIFFERENCES MATRIX * * .. NEI 99..:p1 Rev *. 6 : . . Justification

\ . ., I Validation

I Appendix C -Permanently Not used at DAEC Difference Not applicable to DAEC None > Defueled ICs/EALs 'P C QJ !I § <C E w .........

QJ "' Cl. I "C u .!!! >< QJ *-::s "C -C QJ QJ C C. C. <C 99 ATTACHMENT 4 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELATING TO LICENSE AMENDMENT REQUEST TSCR-166 UPDATED SUPPORTING TECHNICAL INFORMATION 248 pages follow MODE TITLE 1 Power Operation 2 Startup 3 Hot Shutdown (a) 4 Cold Shutdown (a) 5 Refueling<b) Table 1.1-1 (page 1 of 1) MODES REACTOR MODE SWITCH POSITION Run Refuel (a) or Startup/Hot Standby Shutdown Shutdown Shutdown or Refuel (a) All reactor vessel head closure bolts fully tensioned.

Definitions 1.1 AVERAGE REACTOR COOLANT TEMPERATURE

(°F) NA NA > 212 s 212 NA (b) One or more reactor vessel head closure bolts less than fully tensioned. DAEC 1.1-8 Amendment 223 Developme nt of EAL Threshold values f rom NEE-323-CALC-003 C al c ula t e d v alu e s ar e pr ovided i n Calc-00 3 as shown b el ow. Values for the RU1 Gaseous EALs were de t ermined and are shoVvn below. Table 1 -GDseous E ffluent Setpoin t s Locat i o n Detecto r Offgas Stack K a ma n 10 Turbine B uil d i ng Ve n t Kaman 2 Reactor Bu il ding Vent Ka man4 Reactor Bu il ding Vent Kaman6 Reactor Bu il ding Vent Ka ma n a LLR P S F Bu il d i ng Vent Kaman 12 RU 1 Th r es h o l d (µC i/cc) 1.97 E-01 7.74 E-04 6.0 0E-04 9.GO E-04 9.60 E-04 1.1 9 E-03 Va l ues for th e Liqu i d Effluen t RU 1 EALs we re d etermtned and a re shown below. Table 2 -Uquid Effluent Se/points Locat i on . Equ i pment R U 1 U n usua.l ID Event Leve l C S GSW RE-47 67 1.53 E+o3 RHRSW/ESW RE-1 997 8.42E+0 2 RHRSW D il ution L i ne'* RE-4 268 1.06 E+03 The values are rounded for ease of operator use and to provide a step-wise progression t hrough the emergenc y class i fication levels. The resulting values used in the DAEC RU1 .1 EA L are show n in the NOUE column below: I Table R-1 -Effl u ent M o nit o r Class i fication Thresholds Monitor GE SAE Alert NOUE Reactor Building ven t i l at i on rad mon i tor l.lE+OO u c i/cc 1. lE-01 u ci/c c 1.lE-02 uci/cc 8.0E-04 uci/c c (K aman 3/4, 5/6, 7/8) Vl Turbine Building vent il ation rad mon i tor 1.4E+OO uci/cc l.4E-Ol uci/cc 1.4E-02 u c i/c c 8.0E-04 uci/cc ::, (Kaman 1/2) 0 (I/ Vl Offgas Stack rad mon i tor (ti 4.SE+03 u c i/cc 4.SE+02 uci/c c 2.0E-01 uci/cc l9 (Kaman 9/10) 4.SE+Ol u c i/cc LLRPSF rad monitor l.4E-Ol u c i/cc l.4E-02 uci/cc 1.2E-03 ucj/cc (Kaman 12) --GSW rad mon i tor l.7E+04 cps 1.SE+03 cps (RIS-4767) ---"U ::, RHRSW & ESW rad mon i tor 1.2E+04 cps 8.4E+02 cps O" (RM-1997) --::J RHRSW & ESW Rupture D i sc rad mon it or -1.8E+04 cps 1.0E+03 cps (RM-4268) -

Development of EAL Threshold values from NEE-323-CALC-004 Calculated values are provided in Calc-004 as shown below. fable 2 -Recommende d RA 1 Uqr.,id EALs Rad Monitor Equip. Modes 1,2 ,3 Modes4 , 5 cps cps GSW RE-4767 2.32E+4 1.04E+4 RHRSW/ESW R E-1997 1.60E+4 7.20E+3 RH R SW Dilution Li ne RE-4268 2.42E-i4 1.09E+4 The following table of threshold values was developed for use in the DAEC EAL scheme by averaging the separate Mode 1-3 and Mode 4-5 thresholds from Calc-004, and then rounding the average values for ease of EAL evaluator use , as well as to provide a step-wise progression through the emergency classification. Monitor GE SAE Alert GSW r ad mon i to r (R I S-4767) --2.0E+o4cps ::, RHRSW & ESW rad mon i to r (RM-1 997) --l.OE+o4 c ps 0-:::; RHRSW & ESW Ruptu r e D i sc r ad mon i tor (RM-4268)

--2.0E+04 cps Development of EAL Threshold values from NEE-323-CALC-002 Due to elevated background radiation levels on these monitors during plant operation (10-12 R/hr), the calculated threshold value was rounded to 5 (minimum serviceable threshold value accounting for scale of monitor) for ease of use by the EAL evaluator , and the " in Mode 5 only" caveat is added to the EAL usage. The resultant EALs are: RA2.2 Reading greater than 5 R/hr on ANY of the following radiation monitors (in Mode 5 only):

NW Drywell Area Hi Range Rad Monitor , RIM-9184A

South Drywell Area Hi Range Rad Monitor , RIM-9184B CS1/CG1 Core uncovery is indicated by ANY of the following:

Drywell Monitor (9184A/B) reading greater than 5.0 R/hr DAEC EOP BASES DOCUMENT EOP 3 -SECONDARY CONTAINMENT CONTROL GUIDELINE DISCUSSION SF/L-3 D SF/L-2 D Spent Fuel Pool level cannot be maintained above 37 ft 1 in. Main t ain Spent Fuel Pool level above 36 ft . ...-If necessary , use alternate or external makeup sources (SEP 312) . ...-Use only systems not required for adequate core cooling. BASES-EOP 3 Rev. 13 Page 27 of 29 If spent fuel pool level cannot be restored and maintained above the low level alarm setpoint, an alternate control band is established above the higher of the spent fuel pool level LCO (36 ft.) or the Minimum Safe Operating Spent Fuel Pool Level (25.17 ft.). If necessary , normal spent fuel pool makeup may be augmented by one or more of the alternate and external sources listed in SEP 312. The Minimum Safe Operating Spent Fuel Pool Level is generically defined to be the lowest water level providing adequate radiation shielding to (1) protect personnel performing local operations required by the EOPs and (2) allow unrestricted access to the main control room. At the DAEC, the Minimum Safe Operating Spent Fuel Pool Level is defined consistent with NEI 12-02 Level 2 , described as the level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck. The corresponding spent fuel pool level at the DAEC is defined to be 25.17 ft., approximately 10 ft. above the to of the fuel racl<s.

Local Operations for Operating and Normal Shutdown/Cooldown Procedure Step Action If action not performed , does Building Elevation Room Mode Section this prevent shutdown or and Step coo Id own? IPOl 3 , Between 50% and 60% Reactor Power No. The Feed Pumps and N/A N/A N/A N/A Section 5 , shutdown one Condensate and Reactor Condensate Pumps can be tripped step (9) Feed Pump per 01 644 unless otherwise from the Control Room if directed by CRS. necessary , and HPCI and/or RCIC can be used to maintain RPV Level. IPOl 3 , When turbine load is lowered to No. 2 nd Stage Reheat can be le ft in N/A N/A N/A N/A Section 5 , approximately 200 MWe , r emove the 1 E-service and the turbine can be s tep (10) 18A[B] 2 nd St a ge Reheat System from tripped i f nec e ssary. serv i ce in accordance with 01 646 , Extraction Steam. IPOl4 , Secure condensate demineralizers as No. Condensate Demineralizers N/A N/A N/A N/A Section 3 directed by 01 639 , Section 5.1. w i ll automatically go into the " hold" step (10) mode as power and flow are low e red. IPOl 4 , Commence prima ry containment purge No. This is only necessary i f a N/A N/A N/A N/A Section 3 per 01 573. Drywell entry i s ant i cipated. st e p (11) IPOl 4 , At th e refueling bridge , ver i fy that the Main No. Control rod insertion will no t be N/A N/A N/A N/A S e ction 3 Disconnect is closed and that the inhibited.

step (13) SYSTEM START pushbutton has been depressed. I P Ol4 , Prior to disconnecting the generator from No. Aux Boiler is not required t o N/A N/A N/A N/A S e ction 3 the grid , perform the following: (a) If accomplish shutdown. step (14) need e d , start up the Auxiliary Boiler pe r 01 727. IPOl4 , Following Turbine T ri p: (a) Verify that No. These systems can be left in N/A N/A N/A N/A S ection 3 Reactor Coolant Chlor i de and serv i ce if necessary.

step (22) Conductivity analyses have be e n perf o rmed. (b) Operate the Turbine Lube Oi l and Turning Gear System per 01 693.3. (c) Shut down the generator per 01 698. (d) Shut down the turbine per 01 693.1.

Procedure Step Action If action not performed, does Building Elevation Room Mode Section this prevent shutdown or and Step cooldown?

IPOl 4 , Shut down the following genera t or support No. These systems can be left in N/A N/A N/A N/A Section 3 systems , as des i red: Isolated Phase Bus service i f necessary. step (24) Cooling -01 698 , Stato r Water Cooling -01 697 , H 2 Seal Oil -01 695.1, H 2 and CO2 Gas -01 695.2 IPOl4 , Secure hydrogen , oxygen and/or air No. The Hydrogen Water N/A N/A N/A N/A Section 3 injection pe r 01 563 , Hydrogen Water Chemistry System will secure itself step (26) Chemis t ry. i f left i n service. IPOl 4 , As directed by the CRS , perform the No. The MSIVs can be closed if N/A N/A N/A N/A Section 3 following steps as necessa r y to l i m i t necessary t o limi t plant cooldown step (27) reactor vessel depressur i zation following rate. the reactor scram: (b) Start 1 P32 Mechan i cal Vacuum Pump per 01 691. (c) Secure the SJAEs and Offgas per 01 69 1 and 01672. IPOl4 , For the remainde r of this section use the (a) No. The MSIVs can be N/A N/A N/A N/A Section 4 following methods as necessary to closed if necessary to limit step (6) cooldown and depressur i ze the reacto r plant cooldown rate. vessel to maintain a controlled cooldown (b) No -opera t ed from t he rate less than the TS Li mit of 100°F in any Control Room 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per i od. (a) Use the Main Turbine (c) No -Operated from the Bypass Valve to control cooldown per 01 Control Room 693.1 Section 4.5 if available, (b) If (d) No. The MSIVs can be desired cooldown w i th RCIC per 01 150 closed if necessary to limit (preferred method i f MSIVs are closed), plant cooldown rate. (c) If desired cooldown with HPCI per 01 (e) No. The MSIVs can be 152 (RCIC may become inadequate as closed if necessary to limi t pressure lowers) (d) Control steam flow plant cooldown rate. from the reactor vessel to the ma i n condenser through steam seals and steam dra i ns , (e) Secure steam seals per 01 692 as required to l i mit cooldown after the turbine is on the jack and vacuum is broken.

Procedure Step Action If action not performed , does Building Elevation Room Mode Section this prevent shutdown or and Step cooldown?

IP014 , As plant cooldown cont i nues perform the No. The MSIVs can be closed if N/A N/A N/A N/A Section 4 following: (NA i f MSIVs are closed) (a) necessary to limit plant cooldown step (7) Control steam seal pressure 3 to 4 ps i g rate. using M0-1169 , MAIN STEAM SUPPLY , M0-1170 , REGU L ATOR BYPASS a n d/o r M0-1171 , MANUAL UNLOADER on 1C07 , (b) Start 1P-32 MECHANICAL VACUUM PUMP per 01 691 , (c) When reactor pressure approaches 500 ps i g o r cooldown rate cannot be controlled w it h i n the limit , then secure SJAEs and Offgas System pe r 01 691 and 01 672 , respectively , if not previously secured , (d) If not using EHC Pressure Set to control plant cooldown , then at 1C07 , use the PRESSURE SET ADJUST pushbuttons to maintain A[B] PRESSURE SET DEMAND between 150 and 50 ps i g above r eactor pressure as reactor pressu r e decreases.

Otherwise , N/A. IPOl4 , At approx i mately 400 psig , secure the No. The Feed Pumps and N/A N/A N/A NIA Section 4 oper a t i ng feed pump per 01 644. Condensate Pumps can be t ripped step (8) from the Control Room if necessary.

IPOl 4 , When RHR Shutdown Cooling Isolation No , t his system can be placed in N/A N/A N/A N/A S e ction 4 Interlocks can b e reset service from the Control Room if st e p (9) (approximately 100 ps i g), reset the nec e ssary. i solation , then initiate Shutdown Cooling per 01149. IPOl4 , Perf o rm the following after the turbine trip , No. These systems can be left i n N/A N/A N/A N/A S ect i on 4 if needed: (a) Ver i fy that Reactor Coolant service if necessary.

step(10) Chloride and Conductiv i ty analys i s has been performed , (b) Operate the Tu r bine Lube Oil and Turning Gear System per 01 693.3 , (c) Shutdown the Ma i n Generator p e r 0 1 698 , (d) Shutdown the Ma i n Turbine per 01 693.1.

Procedure Step Action If action not performed, does Building Elevation Room Mode Section this prevent shutdown or and Step cooldown?

IPOl4 , Shutdown the following systems as No. These systems can be left in N/A N/A N/A N/A Section 4 directed by the CRS/OSM. service if necessary. step (11) (a) Isolated Phase Bus Cooling per 01 698 , (b) Stator Water Cooling per 01 697 , (c) H 2 Seal Oil per 01 695.1 , (d) H 2 and CO 2 Gas per 01 695.2 , (e) Secure SJAEs per 01 691 and Offgas per 01 672 if not previously performed.

IPOl 4, Perform the following at approximately 50 No. The Feed Pumps and N/A N/A N/A N/A Section 4 psig: (a) Close the BYPASS VALVE Condensate Pumps can be tripped step (12) OPENING JACK SELECTOR , (b) Line up from the Control Room if and place RFP Stuffing Box Pump 1 P-134 necessary. in operation to maintain Seal Water Drain Tank 1T-135 level. IP014 , When steam seal pressure cannot be No. The MSIVs can be closed if N/A N/A N/A N/A Section 4 maintained or the turbine shaft has cooled necessary to limit plant cooldown step (13) per 01 693.3, open Condenser Vacuum rate. Breaker valves V-03-67 and V-03-73. IPOl 4 , Secure MECHANICAL VACUUM PUMP No. The MSIVs can be closed if N/A N/A N/A N/A Section 4 1 P-32 when no longer required per 01 necessary to limit plant cooldown step (14) 691. rate. IP014 , When the condenser is at atmospheric No. The MSIVs can be closed if N/A N/A N/A N/A Section 4 pressure , secure the Turbine Steam Seal necessary to limit plant cooldown step (15) Svstem per 01 692. rate. IPOl4 , Shut down the operating condensate No. The Feed Pumps and N/A N/A N/A N/A Section 4 pump per 01 644 when no longer required Condensate Pumps can be tripped step (18) for RPV Level Control or Hotwell cleanup from the Control Room if recirculation. necessary. Conclusion of manual action evaluation for EALs RA3 and HAS is shown below: EALs RA3 and HA5 are not applicable to DAEC because the evaluation has shown that there are no 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.

All areas outside the Control Room that contain equipment necessary for normal plant operation , cooldown and shutdown do not require physical access to operate.

Development of EAL Threshold values from NEE-323-CALC-005 Calculated values are provided in Calc-005 as shown below. Table3-Recommended RA1 , R S1 , and RG1 EAL Thr:eshokfs (Modes 1 , 2 , 3) R~Pant RAi. RSI RG1 µWee pcf/cc Jld/CC rurbine lklilding

.1.58f.~2 .1.5BE--Ol .:1..SSf+CJO R eactor Boilding l..22 E~l. 1.2Z E-01 :1...22.E+CJO Offgas 5tad: 4.39£+-01.

4.39E-+-02 4.39E+o3 URPSF 1.51f~02 1 51E-01 1.51E+oo*

Tab 1'e 4-Recommended R A1.,. RS1 , and RG1 EA.L Thresholds (Mod es 4 , 5) ReleasePo~

RAl RSI RGI pci/cc ild/cc Ll<lifcc T u rbi ne B ding 1.30!:-02 1.30E-01 1.30 E...OO Reactor B uil ding 1.0'lE-02 .1.01E-Ol :l-01Eto0 Offgas stadr: i,52E;-01 4_52Eim 4_52 E+o3 l.l.RPSF 1.25£-02 1.25E-Ol. 1.25ETDQ*

Per Des ig n Input 5.8 the results en EAL threshold v alues exceed <lhe range o f the monitor. The following table of threshold values was developed for use in the DAEC EAL scheme by averaging the separate Mode 1-3 and Mode 4-5 thresholds from Calc-005, and then rounding the average values for ease of EAL evaluator use , as well as to provide a step-wise progression through the emergency classification. Resulting values are shown in the Alert , SAE , and GE columns below: Table R-1-Effluent Monitor Class i fication Thresholds Monitor GE SAE Alert NOUE Reacto r Building ventil a tion r ad monitor 1.1E+OO uci/cc 1.lE-0 1 uci/cc 1.lE-02 uci/cc 8.0E-04 uci/cc (K a m a n 3/4 , 5/6, 7 /8) C/l T u r b i ne Building vent il a tion rad m o nito r 1.4E+O O uci/cc 1.4E-01 u c i/cc 1.4E-02 uc i/cc 8.0E-04 u c i/cc ::, (Kam an 1/2) 0 a., C/l Offga s Stack rad monitor ro 4.5E+03 uci/cc 4.5 E+02 uci/cc 4.SE+Ol uci/cc 2.0E-01 uci/cc l.9 (Kaman 9/10) L L RPSF rad monitor 1.4E-01 u c i/cc 1.4E-0 2 uci/cc 1.2E-03 uci/cc (Kaman 12) ---G S W r ad monitor 1.7E+04 cps 1.5E+03 cps (R IS-4767) ------'D RHRSW & ESW rad monitor ::, 1.2E+04 cps 8.4E+02 cps 0-(RM-1997) ------:.:J RHRSW & ESW Ruptur e Di sc rad monitor ------1.8E+04 c p s 1.0E+03 c ps (RM-4268)

DAEC EOP BASES DOCUMENT EOP 3 -SECONDARY CONTAINMENT CONTROL GUIDELINE DISCUSSION SF/L-4 Spent Fuel Pool level drops to 16.36 ft D SF/L-5 D Operate Spent Fuel Pool sprays (SAMP 712) . .-Use only systems not required for adequate co r e cool i ng. BASES-EOP 3 Rev. 13 Page 29 of 29 If spent fuel pool level cannot be controlled using alternate or external makeup sources, sprays are used to add water to the spent fuel pool, cool exposed bundles, and reduce radioactivity releases.

However , spray operation may damage electrical equipment and flood lower elevations of the secondary containment, complicating implementation of other emergency response strategies, and runoff from sprays could spread radioactivity release. Use of sprays is therefore delayed until it is determined that spent fuel pool level cannot be maintained above the top of the fuel racks. As long as the spent fuel assemblies are covered with water , the fuel will not overheat and efforts should focus on providing sufficient makeup flow to keep the assemblies submerged.

The lowest measurable spent fuel pool level using the wide range instrument is 16.16 ft., approximately one foot above the top of the spent fuel racks. The action level in SF/L-4 corresponds to NEI 12-02 Level 3 , the level at which fuel remains covered but actions to implement make-up water addition should no longer be deferred.

The " before" condition permits appropriate anticipatory action based on the spent fuel pool leakage rate, radiation levels, available resources , and the time required to place sprays in service. Steps to prepare spray equipment for use should be initiated while radiation levels permit access to the refueling floor and timed to optimize use of available resources.

As in Steps SF/T-3 and SF/L-3, available spray sources may be alternated between RPV injection and spent fuel pool spray modes as long as adequate core cooling can be maintained , but maintaining adequate core cooling takes precedence over spent fuel pool cooling (refer to the discussions of Steps SF/T-3 and SF/L-3 above).

AC Sources -Operating B 3.8.1 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.1 AC Sources -Operating BASES BACKGROUND DAEC The unit Class 1 E AC Electrical Power Distribution System AC sources consist of the offsite power sources (preferred and alternate preferred), and the onsite standby power sources (Diesel I Generators (DGs) 1 G-31 and 1 G-21 ). As discussed in UFSAR Section 3.1.2.2.8 (Ref. 1 ), the design of the AC Electrical Power System provides independence and redundancy to ensure an available source of power to the Engineered Safety Feature (ESF) Systems via essential buses 1A3 and 1A4. The Class 1 E AC Distribution System is divided into redundant load groups , so loss of any one group does not prevent the minimum safety functions from being performed. Each load group has connections to two preferred offsite power supplies and a single DG. Offsite power is supplied to the 161 kV and 345 kV switchyards from the transmission network by six transmission lines. The 345 kV switchyard and the 161 kV switchyard are connected via the autotransformer , and both sections of the switchyard are connected to the transmission grid by at least two independent lines. From the 161 kV switchyard (the preferred power source), a single overhead transmission line feeds the startup transformer.

From the startup transformer , dual isolated secondary windings provide feeds to the 4160 volt essential buses , 1 A3 and 1 A4 , through separate bus supply lines and circuit breakers.

The startup transformer is sized to supply all plant power (both essential and non-essential loads) during unit startup. From the tertiary winding on the autotransformer (the alternate preferred power source), a single 34.5 kV underground line feeds the standby transformer.

From the standby transformer , a single 4160 volt line feeds both essential buses through separate bus supply circuit breakers.

A detailed description of the offsite power network and circuits to the onsite Class 1 E essential buses is found in the UFSAR , Sections 8.2.1.3 and 8.3.1.1.5 (Ref. 2). An offsite circuit consists of all breakers , transformers , switches , interrupting devices , cabling , and controls (continued)

B 3.8-1 TSCR-044A BASES BACKGROUND (continued)

DAEC AC Sources -Operating B 3.8.1 required to transmit power from the offsite transmission network to the onsite Class 1 E essential bus or buses. Startup transformer (1 X3) provides the normal source of power to the essential buses 1 A3 and 1 A4. If either 4.16 kV essential bus loses power , an automatic transfer from the startup transformer to the standby transformer (1X4) occurs. The startup transformer and standby transformer are both sized to accommodate the starting of all ESF loads on receipt of an accident signal. Emergency loads are sequenced onto the essential buses regardless of the source of power (onsite or offsite). The onsite standby power source for4.16 kV essential buses 1A3 and 1A4 consists of two DGs. DGs 1G-31 and 1G-21 are dedicated to essential buses 1A3 and 1A4 , respectively.

A DG starts automatically on a Loss of Coolant Accident (LOCA) signal (i.e., low reactor water level signal or high drywell pressure signal) or on an essential bus degraded voltage or undervoltage signal. After the DG has started, it automatically ties to its respective bus after offsite power is tripped as a consequence of essential bus undervoltage or degraded voltage, independent of or coincident with a LOCA signal. The DGs also start and operate in the standby mode without tying to the essential bus on a LOCA signal alone. Following the trip of offsite power , non emergency loads powered from essential buses are load shed. When the DG is tied to the essential bus, loads are then sequentially connected to its respective essential bus. The sequencing logic controls the permissive and starting signals to motor breakers to prevent overloading the DG. In the event of a loss of both the preferred power source and the alternate preferred power source , the ESF electrical loads are automatically connected to the DGs in sufficient time to provide for safe reactor shutdown and to mitigate the consequences of a Design Basis Accident (OBA) such as a LOCA. Certain required plant loads are returned to service in a predetermined sequence in order to prevent overloading of the DGs in the process. Within 25 seconds after the initiating signal is received , all automatic and permanently (continued)

B 3.8-2 TSCR-111 ANNUNCIATOR RESPONSE PROCEDURE ARP 1C08A GENERATOR AND AUXILIARY POWER Usage Level Reference Use Record the following:

Dateffime:

______ / ____ Initials:

__ _ NOTE: User shall perform and document a Temp Issue/Rev. Check to ensure revision is current , in accordance with procedure use and adherence requirements. Prepared By: I Date: ------------Print Signature CROSS-DISCIPLI N E REVIEW (AS REQUIRED)

Reviewed By: I Date: Print Signature Reviewed By: I Date: Print Signature PROCEDURE APPROVAL Approved By I Date: ---------------------

'I Print Signature Page 1 of 140 Rev. 89 1 2 3 A AUX XFMR TO 1A 1 BUS 1A1 S/U X FM R TO 1A1 BREAKER 1A101 LOCKOU T TRIP BREAKER 1A102 OR TRIP LOSS OF VOLTAGE TRIP 1A1 TO XFMR 1X11 1A1 TOXFMR 1X71 1A1 TO XFM R 1X51 BREAKER 1A107 BREAKER 1A 108 BREAKER 1A109 B TRIP TR I P TRIP LC 1611162 XFMR 1X11 TOLC181 XFMR 1X51 TO LC 165 BREAKER 18101 CROSS TIE BREAKER BREAKER 16501 T RIP 16107 TRIP TRIP C LC 181 BREAKER LC 185/186 LOAD CENTER 185 18102 , 18103 CROSSTIE BREAKER 18502 BREAKER 18104 OR 18105 18505 1 8503 OR 1 8504 TRIP TRIP TR I P D ALARM WINDOW ENGRAVINGS AND GRID LAYOUT 1C08A Same on annunciator panel 1 COBB for Division 2 4 5 6 7 8 9 STBY XFMR TO 1A3 S/U X F MR TO 1A3 S T ARTUP XFMR UNINTERRUPTIB LE AC 125VDC BREAKER 1A301 BUS 1A3 BREAKER 1A302 1X3 1Y23 UNDERVOL T AGE SYST E M 1 LOCKOUT TRIP OR TROU BL E TRIP TR I P TROUBLE INVERTER TROUBLE LC XFMR 1X91 INSTRUMENT AC SWITCHY ARD SUPPLY LC XFMR 1X31 BREAKER 1A312 MAIN GENERATOR 1Y21 125VDC BREAKER 1A110 BREAKER 1A303 OR I MPROPER PHAS E UNDERVOL TAGE CHARGER 1012 TRIP TRIP MCC 1891 BKR 16903 SEQUENCE OR TROUBLE TRIP INVERTER TROUBLE 125VDC LC 183 BREAKER INSTRUMENT AC 1Y11 125VDC SYSTEM 1 18301, 18302 BUS 1A3 STARTUP XFMR UNDERVOLTAGE CHARGER 1D120 BATTERY 101 1 8303 OR 1 8304 LOSS OF VOL TAGE LOCKOUT TRIP OR TROUBLE DISCONNECTED TRIP INVERTER TROUBLE MCC 1B34A/1844A "A" DIESEL GEN MCC 1834A TIE BREAKER 4KV BUS DIESEL FUEL OIL 1G-31 TIE BKR 183401 AUTO TRANSFER STORAGE TANK H-35 TRIP 183402 OR 1 84402 INOP LO LEVEL CONTROL P OWER TRIP FAILURE Page 2 of 140 10 11 12 "A" DIESEL G E N A DG TO BUS 1A3 "A" DIESEL GEN 1G-31 BREAKER 1A311 1 G-31 RUNNING TRIP LOCKOUT TRIP "A" DIESEL GEN " A" DIESEL GEN FUEL O IL DAY TANK 1G-31 "A" DIESEL GEN 1T-37A PHASE OVERCURRENT 1G-31 LO-LO-LEVEL OR GROUND FAULT OVERSPEED TRIP AUX BOILER "A" DIESEL GEN "A" DIESEL GEN FUEL TANK 1T-34 PANEL 1C-93 1G-31 LO LEVEL TROUBLE ENGIN E CRANKING "A" DIESEL GEN 1G-31 " A" DIESEL GEN " A" DIESEL GEN 1G-31 1G-31 AUTOSTART ENGINE SHUTDOWN START FA IL URE INHIBITED Rev. 89 125V DC SYSTEM 1 TROUBLE ANNUNCIATOR PANEL: COORDINATES

PAGE: 1 of 125 VDC SYSTEM 1 TROUBLE (GROUND OR LOW VOLTAGE) 1C08A A-9 3 TITLE: 1.0 PROBABLE CAUSE(S) / INITIATING DEVICE(S)

/ SETPOINT(S) 1.1 Ground fault on 125 VDC System 1 Positive to ground Relay 64 9 Volt differential or negative to ground Relay 64 1.2 125 VDC System 1 low Relay 1 D 10-27 105 VDC (dee) voltage 1.3 Positive or negative metering FU 3 amp fuse blown blown fuse blown 2.0 AUTOMATIC ACTIONS 2.1 If due to a complete loss of 125 VDC System 1: a. Various control systems half trip. b. Scoop Tube for Recirc Pump A locks up. c. Breaker 1 B3401 auto trips open after 6 second time delay , 1 B4401 auto closes. d. Static Switch JS1501 transfers from Inverter 1D15 to Regulating Transformer 1 Y1A. 2.2 If the 125 VDC System 1 is not lost, no AUTOMATIC ACTIONS occur. Page 36 of 140 Rev.89 ANNUNCIATOR PANEL: COORDINATES:

PAGE: 1 of 125V DC CHARGER 1D12 TROUBLE 1C08A B-9 3 TITLE: 125 voe CHARGER 1012 TROUBLE 1.0 2.0 NOTE This is a normal alarm anytime Charger 1012 is being changed over to Charger 10120. PROBABLE CAUSE(S) I INITIATING DEVICE(S)

I SETPOINT(S) 1.1 AC Breaker 1012-01 open Relay K-8 OPEN position 1.2 DC Breaker 1012-02 open Relay K-7 OPEN position AND DC Breaker 1012-03 open Relay K-7 OPEN position 1.3 Charger Failure Relay K-5 < 5 Amps (dee) 40 second time delay 1.4 Reverse Current Relay K-4 Reverse Current Detected 1.5 DC Undervoltage Relay K-2 105 voe (dee) 1.6 AC Undervoltage Relay K-1 340 VAC (dee) AUTOMATIC ACTIONS 2.1 If due to CAUSE 1.3 , 1.4 , or 1.5 , Charger 1D12 front panel trouble light illuminates.

Page 69 of 140 Rev.89 ANNUNCIATOR PANEL: COORDINATES

PAGE: 1 of 125V DC CHARGER 1D120 TROUBLE 1C08A C-9 3 TITLE: 125 voe CHARGER 10120 TROUBLE 1.0 2.0 NOTE This is a normal alarm anytime 1D120 is being changed over to Chargers 1D12 or 1 D22. PROBABLE CAUSE(S) I INITIATING DEVICE(S)

I SETPOINT(S) 1.1 AC Breaker 1D120-01 open Relay K-8 Open Position 1.2 DC Breaker 1D120-02 open Relay K-7 OPEN Position 1.3 DC Breaker 1D120-03 open Relay K-7 OPEN Position 1.4 Charger Failure Relay K5 < 5 Amps (dee) 40 Second TD 1.5 Reverse Current Relay K-4 Reverse Current Detected 1.6 DC Undervoltage Relay K-2 105 voe (dee) 1.7 AC Undervoltage Relay K-1 340 VAC (dee) AUTOMATIC ACTIONS 2.1 If due to CAUSE 1.4 , 1.5 , or 1.6 , Charger 1D120 front panel trouble light illuminates.

Page 102 of 140 Rev. 89 TYPE AC safety buses 125 voe buses 250 voe buses Distribution Systems -Operating B 3.8.7 Table B 3.8.7-1 (page 1 of 1) AC and DC Electrical Power Distribution Systems VOLTAGE DIVISION 1 <a> DIVISION 2<a> 4160 V Essential Bus 1 A3 Essential Bus 1 A4 480V Load Centers Load Centers 1B3 , 1B9 1B4, 1B20 480V Motor Control Motor Control Centers Centers 1 B32 , 1 B34 1B42, 1B44 125 V Distribution Panels Distribution 1010, 1011, Panels 1013 1020, 1021, RCIC Motor Control 1023 Center 1014 250V N/A Distribution Panel 1040 Motor Control Centers 1 041 and 1 042 <a> Each division of the AC and DC electrical power distribution systems is a subsystem. DAEC B 3.8-73 Amendment 223 I AOP 302.1 LOSS OF 125 voe POWER ABNORMAL OPERA TING PROCEDURE AOP 302.1 LOSS OF 125 VDC POWER U sage Level Reference Use NOTE This AOP is normally coordinated by the Reactor Operator.

Record the following: Date/Time: ______ I ____ Initials:

__ _ NOTE: User shall perform and document a Temp Issue/Rev. Check to ensure revision is current , in accordance with procedure use and adherence requirements. Enter the following as applicable:

LOSS OF 125 voe 1011 PAGE 2 LOSS OF 125 voe 1013 PAGE 8 LOSS OF 125 voe 1014 PAGE 14 LOSS OF 125 v o e DIV I PAGE 17 LOSS OF 125 voe 1 D21 PAGE 28 LOSS OF 125 voe 1 D23 PAGE 35 LOSS OF 125 voe DIV II PAGE 43 COMPLETE LOSS OF 125 voe PAGE 56 I AOP 302.1 Page 1 of 63 Rev. 58 AOP 302.1 LOSS OF 125 VDC POWER LOSS OF 125 VDC DIV I IMMEDIATE ACTIONS 1. Place RX WATER LEVEL CONTROL INPUT SELECT HSS-4560 in A LEVEL at 1C05. 2. Place 1 OR 3 ELEMENT CONTROL SELECT HSS 4450 in 1 ELEMENT at 1C05. AUTOMATIC ACTIONS 1 G001 Voltage Regulator transfers to manual with no adjustment capability Loss of 1 A 1 /1 A3 breaker control MG Set A Scoop Tube Power Failure Lockout initiates (no Amber lockup light) SBGTS A SV-5801A fails closed , SV-5815A , SV-5817A and SV-5825A fail open MCC 1 B34A/1 B44A will auto transfer to 1 B44 JS1501 will transfer from Inverter 1D15 to Reg Trans 1Y1A Group 3A Primary Containment Isolation (Lockout relay w i ll not trip) Rx FEEDWATER FLOW B FEEDLINE Fl-1626 fails downscale STEAM FLOW B STEAMLINE Fl-4409 fails downscale Recirc Lube Oil Pumps 1 P202A & 1 P202B trip , and 1 P202C auto starts on low oil pressure (continued operation of the A Recirc MG Set is allowable in this condition).

B GEMAC LEVEL Ll4560 fails low HPCI will not trip on high level Loss of 1D10 with B LEVEL selected and no operator action results in: Feedwater control opens Feed Reg Valves Reactor water level goes high "B" Reactor Feed Pump and Main Turbine Trip on high level ("A" Reactor Feed Pump cannot be tripped remotely)

Reactor Scram (Main turbine trip) Loss of 1A1 and 1A2 power (failure to auto transfer)

If >26% power , Reactor Recirc. Pumps 1 P-201A and 1 P-201 B trip (RPT) I AOP 302.1 Page 17 of 63 Rev. 58 AOP 302.1 LOSS OF 125 voe POWER LOSS OF 125 VDC DIV I FOLLOW-UP ACTIONS NOTE Follow up actions may be performed in any order. 1. Establish critical parameter monitoring of RPV Water Level , as priorities allow. 2. Stabili z e reactor powe r level and maintain recirculation loop flows balanced. Use manual control of the A MG Set and Recirc Pump from the MG Set Room i n accordance with 01 264. NOTE Buses 1 A 1 and 1 A2 will not auto transfer to the startup transformer on a turbine trip. 3. Transfer Bus 1A2 to the startup transformer per 01 304.1. 4. IF 1A4 has power available THEN verify TIE BREAKER 184401 MCC 1B44A/1834A is closed. 5. IF a Reactor Scram occurs: THEN perform the following: a. Verify main turbine trip. b. Verify the H and I breakers open. c. Only after H and I are open , direct an operator to trip the GENERATOR EXCITER FIELD BREAKER locally. 6. Reference EPIP 1.1 for EAL assessment , for instrumentation, perform alarm panel checks as needed to confirm threshold is met. 7. Suspend all evolutions in progress associated with electr i cal switchgear and switching operations. 8. Locally operate affected switchgear to start and stop equipment as required. I AOP 302.1 Page 18 of 63 Rev. 58 AOP 302.1 LOSS OF 125 voe POWER LOSS OF 125 VDC DIV I FOLLOW-UP ACTIONS (continued)

NOTE Loss of 125 VDC DIV 1 causes a loss of 1 A 1 control power. If 1 A 1 is on the aux transformer , all 1 A 1 loads will remain energized with no automatic trips or starts until the Main Generator is tripped. Likewise , when the main generator is tripped , All 1A1 loads will be lost until manually restored. 1A1 loads:

A Feed Pump

A Condensate Pump

A Recirc MG Set

A Gire Water Pump

Load Center 181

Load Center 185

Load Center 1 87 9. IF 1A1 isdeenergized

a. Trip 1A101. b. Strip 1 A 1 loads. c. Close 1A102. (1A103)-no 211" trip (1A106) (1A104) (1A 105) (1A107) (1A109) (1A108) THEN reenergize 1A 1 locally: d. Restart loads as required. NOTE If operation of the RCIC System is required while the Division 1125 VDC System is deenergized , use the HPCI System in manual mode. 1 O. Evaluate the status of the 125 VDC Electrical Distribution System to determine the cause of the malfunction. I AOP 302.1 Page 19 of 63 Rev. 58 I I J AOP 302.1 LOSS OF 125 voe POWER LOSS OF 125 voe DIV I FOLLOW-UP ACTIONS (continued)

11. IF 1010 is totally deenergized THEN perform the following: 12. 13. 14. 15. 16. a. Open all branch circuit breakers at Panels 1010 , 1011 , 1013 , and 1014. b. Verify MCC 1 B32 energized. c. Locally inspect circuit breakers at Panels 1010 , 1011 , 1013 , and 1014. d. Verify Battery Room ventilation. e. Locally inspect Battery 101 f. Locally inspect Battery Charge r s 1012 and 10120. g. Restore power to 1010 and branch panels. h. Comply with Tech Spec Requirements for " Distribut i on Systems -Operating" or " Distribution Systems -Shutdown", as app l icable. i. Comply with ACP 1412.4 Impairments to Fire Protection System. IF 1D10 is not lost , but one or THEN investigate and evaluate the status of more loads are lost the individually affected loads and comply with Tech Spec Requirements for " Distribution Systems -Operating" or " Distribution Systems -Shutdown", as applicable. IF 125 voe Div I Battery 101 is THEN comply with Tech Spec Requirements made or found to be for " DC Sources -Operating" or " DC inoperable Sources -Shutdown" , as applicable. IF 125 voe Div I Battery 101 is THEN Verify battery cell parameters per < 110 voe STP 3.8.4-02 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . WHEN Div I 125 voe System THEN send an operator to TIE BREAKER is restored and the use 1 B3401 MCC 1 B34N44A to press of-TIE BREAKER reset button prior to closing. 1 B3401 is desired WHEN power is restored T H EN reset A Scoop Tube lockout per 01264. I AOP 302.1 Page 20 of 63 Rev. 58 AOP 302.1 LOSS OF 125 voe POWER LOSS OF 125 voe DIV I PROBABLE ANNUNCIATORS 1C08A , A-2 Bus 1 A 1 LOCKOUT TRIP OR LOSS OF VOLT AGE STARTUP XFMR 1X3 TROUBLE 125 voe SYSTEM 1 TROUBLE 125 voe CHARGER 1D12 TROUBLE 125 voe SYSTEM 1 BATTERY 1D1 DISCONNECTED BUS 1A3 LOSS OF VOLTAGE A-7 A-9 B-9 C-4 C-6 C-8 C-9 D-5 D-9 INSTRUMENT AC 1Y11 UNDERVOLTAGE OR INVERTER TROUBLE 125 voe CHARGER 10120 TROUBLE MCC 1834A TIE BKR 183401 TRIP A DIESEL GEN 1G-31 CONTROL POWER FAILURE 1C08C , A-3 8-3 B-6 MAIN GENERATOR LOCKOUT RELAY CKT LOSS OF 125 voe MAIN GENERATOR VOLTAGE REGULATION IN MANUAL Hz/STATOR COOLING PANEL 1C83 LOSS OF DC PROBABLE INDICATIONS Annunciators

-Loss of power to the following panels: 1C03 1C04 1C14 1C05 1C08B 1C09A 1 C08 -Loss of the following: 1C23A 1C24A 1C22 1C25A 1C26A 1C34 1C35A 4160V BUS 1A1 switchgear control , indication and automatic trip functions 480V LC 181 switchgear control and indication 480V LC 187 switchgear control and indication 480V LC 185 switchgear control and indication 4160V BUS 1A3 switchgear control , indication and automatic trip functions 480V LC 1 B3 switchgear control and indication 480V LC 1 B9 switchgear control and indication TIE BREAKER 1 B3401 MCC 1 B34A/1 B44A control and indication GENERATOR EXCITER FIELD BREAKER control and indication 1 CO? -Loss of the following: EMERGENCY BEARING OIL PUMP 1 P-40 control and indication I AOP 302.1 Page 21 of 63 1C40 1C40A Rev. 58 AOP 302.1 LOSS OF 125 voe POWER LOSS OF 125 voe DIV I PROBABLE INDICATIONS (continued) 1 C06 -Loss of the following: A CIRC WATER PUMP 1P-4A control and indication A CONDENSATE PUMP 1 P-8A control and indication A REACTOR FEEDWATER PUMP 1 P-1A control and indication A GSW PUMP 1 P-89A control and indication A[C] RWS PUMP 1 P-117A and C control and indication 1 C05 -Loss of the following: A WIDE RANGE LEVEL Ll-4539 i nd i cation fails low B GEMAC LEVEL Ll-4560 fails low B REACTOR PRESS Pl-4564 fails low A CRD PUMP 1 P-209A control and indication Rx FEEDWATER FLOW B FEEDLINE Fl-1626 fails downscale STEAM F LOW B STEAMLINE Fl-4409 fails downscale 1 C04 -Loss of the following: MG SET LUBE OIL PUMP 1 P-202A control and indication MG SET LUBE OIL PUMP 1 P-202B control and indication A MG SET SPEED CONTROL control and indication A MG S E T EM ERG AUX OIL PUMP 1 P-204A control and indication RCIC TURBINE CONTROL VALVE HV-2406 position indication RCIC System Drain Valves RCIC STEAM LINE DRAIN ISOL CV-2410 , and CLOSED RADWASTE DISCH ISOL CV-2435 RCIC System Condensate Pump Motor and Motor Operated Valves control and indication Inboard MSIVs indication and DC Solenoid Valve control 1 C03 -Loss of the following: A CORE SPRAY PUMP 1P-211A control and indication A RHR PUMP 1 P-229A and C RHR PUMP 1 P-229C control and indication A RHRSW PUMP 1 P-22A and C RHRSW PUMP 1 P-22C control and indication A RHR HX SHELL OUTBD VENT M0-2044A and A RHR HX SHELL INBD VENT M0-2044B percent indication HPCI STEAM LINE DRAIN ISOL CV-2211 and CLOSED RADWASTE DISCH ISOL CV-2234 control and indication I AOP 302.1 Page 22 of 63 Rev. 58 AOP 302.1 LOSS OF 125 voe POWER LOSS OF 125 VDC DIV I ............................................................... INFORMATION

1010 125 VDC Distribution Panel loads: 1 D10 ckt 04 1D13 125 VDC Distribution Panel C 1D10 ckt 05 1D14 125 voe RCIC MCC 1 D1 O ckt 06 1 D11 125 VDC Distribution Panel A 1 D10 ckt 07 Instrument AC Inverter 1D15 Supply 1 D11 125 VDC Dist r ibution Panel A loads: 1 D 11 ckt O 1 Load Center 1 B 1 switchgear control 1 D11 ckt 02 RWCU F/D Panel 1 C82 annunc i ators 1 D11 ckt 03 Load Center 185 switchgear control 1 D11 ckt 04 Load Center 187 switchgear control 1 D 11 ckt 05 4160V Bus 1 A 1 switchgear control 4KV BREAKER 1A101 AUX XFMR TO BUS 1A1 4KV BREAKER 1A102 STARTUP XFMR TO BUS 1A1 Reactor Feed Pump 1 P-1 A Supply Breaker 1 A 103 Reactor Recirculation MG Set 1 G-201A Supply Breaker 1A 104 Circulating Water Pump 1 P-4A Supply Breaker 1 A 105 Condensate Pump 1 P-8A Supply Breaker 1 A 106 FEEDER BREAKER 1A107 1A1 TO LC XFMR 1X11 FEEDER BREAKER 1A108 1A1 TO LC XFMR 1X71 FEEDER BREAKER 1A109 1A1 TO LC XFMR 1X51 FEEDER BREAKER 1A110 1A1 TO SWYD LOAD CENTER Reactor pressure Channel B calculation and indication Reactor water level Channel B calculation and indication 1 D11 ckt 06 Load Center 189 switchgear control 1 D11 ckt 07 MCC breaker 183401 control (Normal) 1 D11 ckt 08 Main Generator excitation control Generator Exciter Field Breaker control and indication Motor Driven DC Regulator Setpoint Adjust (1 COS) Motor Driven AC Regulator Setpoint Adjust (1 COS) Exciter Field Flashing Regulator Transfer and Lockout Relay Exciter Field Bridge Overcurrent Alarm Generator Field Bridge Over temperature Alarm Exciter Field Current Limit Circuit Volts/Hertz Protective Panel 1 D11 ckt 09 Generator H2 Cooling Panel 1 C83 Associated Generator Trip and Load Runback Relays Annunciators I AOP 302.1 Page 23 of 63 Rev.58 AOP 302.1 LOSS OF 125 VDC POWER LOSS OF 125 VDC DIV I * * * * * * * * *** * * * * * * * * * * * * * * * * ** * * * * * * * * * * * * * * * * ** * * * * * * * ** .. ** .. *INFORMATION

.... * .. ** .. ** .. *********

1 D11 125 VDC Distribution Panel A loads (continued)

1 D1 1 ckt 10 Main T r ansformer 1X1 control power 1 D11 ckt 11 1 G-31 Diesel Gen. Control Panel 1 C117 1 D 11 ckt 12 1 G-31 Diesel Gen. Control Panel 1 C 117 1 D11 ckt 13 Startup Transformer 1X3 control power 1 D11 ckt 15 Core Spray Channel A Relay Vertical Board 1 C43 Core Spray System Loop A Logic 1 D11 ckt 17 Radwaste Panel 1 C84 annunciators 1 D11 ckt 18 1 G-31 Diesel Gen. Exciter Panel 1 C93 1 D11 ckt 19 Standby Gas Treatment System Control Panel 1 C24A control PASS System Valves SV-4594A , SV-4595A , and SV-8772A (FC) A SBGTS valves SV-5801A , SV-5815A , SV-5817A , and SV-5825A (CV-5815A, CV-5817A , and CV-5825A (FO) (AV5801 (FC) A SBGTS vent shaft Rad Monitor Aux Relay 95-K134A (PCIS GP 3A input) A SBGTS Fire Deluge SV-5837A (CV-5837A (FC)) A SBGTS Preheater control (TORUS) EXTERNAL VACUUM BKR ISOL (CV-4304 (FO)) CAMS Loop A Isolation Valve control and indication Offgas Stack Exhaust Fan 1V-EF-18A remote control Panel 1 C24A annunciators Panel 1 C25A annunciators 1 D11 ckt 20 Turbine Building and Control Room HVAC Panel 1 C26 SFU Fire Deluge SV-7328A (CV-7328A (FC)) I AOP 302.1 A DIESEL GENERATOR 1G-31 Room Supply Fan 1V-SF-20 remote control A DIESEL GENERATOR 1G-31 Room Supply Fan 1V-SF-20 dampers D0-7001A and D0-7002A position indication 1V-SFU-30A valves CV-7301A and SV-7318A (AV-7301A and AV-7318A (FO)) Miscellaneous Reactor , Turbine , and Control Building isolation dampers 1 C23A annunciators 1 C26A annunciators and indications Page 24 of 63 Rev. 58 AOP 302.1 LOSS OF 125 voe POWER LOSS OF 125 voe DIV I ** ** * * * * * * * * * **** * * * * * * * * ** ** * * * * ** * * * * ** * * * * * * * * * * * * * * ********INFORMATION

- - - - 1D13 125V DC Distribution Panel C loads: 1D13 ckt 01 Reactor Recirculation Pump MG Set 1 G-201 A Control Panel 1 C112A and Scoop Tube Power Failure Lock circuitry 1D13 ckt 13 Recirculation Pump MG Set 1 G-201 A control (Normal and Standby Power) MG Set A Emergency Lube Oil Pump 1 P-204A control and indication Loss of Division I ATWS/ARI/RPT Trip System (101313 only) 1D13 ckt 02 Reactor Core Cooling Bench board 1 C03 RHR Heat Exchanger A Vent Valves M0-2044A and M0-2044B position indication (Zl-2044A and Zl-2044B)

HPCI System Drain Valves SV-2211 and SV-2234 (CV-2211 and CV-2234) control and indication Position indication for CV-4309 (ZS-4309) 1D13 ckt 03 Reactor Water Cleanup and Recirculation Bench board 1 C04 RCIC Inverter RCIC Governor Valve HV-2406 Position indication RCIC System Drain Valves SV-2410, and SV-2435 (CV-2410 and CV-2435) control and indication 1D13 ckt 04 Annunciator Power Panels 1 C03 , 1 C04 , 1 C05 , 1 C08B , 1 C34 , Panel 1 C22 Frequency Converter 1D13 ckt 05 CAD Panel 1 C35A , CAM Panel 1 C09A SV-4332A, Upper Drywell Spray CAD N2 Primary Containment Isolation SV-4334A North Torus Spray Header Primary Containment lsol SV-4332B , and SV-4334B control and indication 1 C35A Annunciators 1 C09A Annunciators 1C014A EOP Annunciators 1D13 ckt 06 1 C40 Annunciators 1D13 ckt 07 1 C32 Channel A RHR Relay Vertical Board RHR Loop A Logic I AOP 302.1 HPCI Low Water Level initiation signal HPCI Isolation Logic A HPCI Rx Hi-Level trip logic (1/2 of logic, HPCI will not trip on Hi-Level) Page 25 of 63 Rev. 58 AOP 302.1 LOSS OF 125 voe POWER LOSS OF 125 voe DIV I * * ** * * * * * * * * * * ** * * * * ** * * *** * * * * * * * * ** * * * * * * * * * * * * * * * * * * ********INFORMATION

- - - - 1013 125 voe Distr i bution Panel C loads: 1013 ckt 08 Reactor Protection System Channel A Vertical Board 1C15 Recirculation Pump A Trip circuitry Backup Scram Valve SV-1840A (FC) 1013 ckt 09 Inboard Isolation Valve Relay Panel 1 C41 Inboard MSIVs position indication and DC solenoid valve control 1013 ckt 10 1C40A Annunciators 1013 ckt 11 EBO Pump 1 P-40 Starter Control (HS-3151)

& Indication Emergency Bearing Oil Pump 1 P-40 control and indication 1013 ckt 12 Generator and Plant Relay Panel 1 C31 Generator Primary Lockout Relay 286/P Startup Transformer

-Bus 1A3 Breaker 1A302 Lockout Relay Essential Bus 1A3 Load Shedding Circuit Non-Essential Auto Transfer and Load Shed 1013 ckt 14 Auto Slowdown Panel 1C45 ADS main control power 1013 ckt 15 4160V Bus 1A3 switchgear control 4KV BREAKER 1A301 STANDBY TRANSFORMER TO BUS 1A3 4KV BREAKER 1A302 STARTUP TRANSFORMER TO BUS 1A3 FEEDER BREAKER 1A303 1A3 TO LC XFMR 1X31 Core Spray Pump 1 P-211A Supply Breaker 1A304 RHR Pump 1 P-229A Supply Breaker 1A305 RHR Pump 1 P-229C Supply Breaker 1A306 RHR SW Pump 1 P-22A Supply Breaker 1A307 RHR SW Pump 1P-22C Supply Breaker 1A308 GSW Pump 1 P-89A Supply Breaker 1A309 CRD Pump 1 P-209A Supply Breaker 1A310 A DG TO BUS 1A3 BREAKER 1A311 FEEDER BREAKER 1A312 , 1A3 TO LC XFMR 1X91 Essential Bus 1 A3 Degraded Voltage Detection Circuit 1 D 13 ckt 16 Load Center 183 switchgear control 1013 ckt 17 RCIC Relay Vertical Board 1C30 RCIC Turbine Speed Controller RCIC Turbine Trip circuitry RCIC Initiation and Isolation Relay Logic A RCIC Instrumentation 1013 ckt 19 NSSS Temperature and Leak Detection Panel 1C21 RCIC Area Steam Leak Detection circuitry Division I RCIC Timer Logic Division I 1013 ckt 20 Remote Shutdown Panels 1 C389 and 1 C390 Transfer Switch Position Status Indication I AOP 302.1 Page 26 of 63 Rev. 58 AOP 302.1 LOSS OF 125 voe POWER LOSS OF 125 voe DIV I * * * * ** * * * * * * * * *** * * * * * * * * * * * * ** * * ** * * * * * * * * * * * * * * * * * * * * * *******INFORMATION

- - - - 1 D14 125 VDC RCIC MCC loads: 1 D14 ckt 01 Steam Outboard Isolation Valve M0-2401 1 D14 ckt 02 Steam Supply Valve M0-2404 1 D14 ckt 03 Turbine Stop Valve M0-2405 1 D14 ckt 04 Bypass to Condensate Valve M0-2426 1 D14 ckt 05 Suct i on from Condensate Sto r age Tank Valve M0-2500 1 D14 ckt 06 Minimum Flow Bypass Valve M0-2510 1 D14 ckt 07 Normally Open Discharge Valve M0-2511 1014 ckt 08 Normally Closed Discharge Valve M0-2512 1 D14 ckt 09 Test Discharge Valve M0-2515 1014 ckt 10 Suction at Pool Valve M0-2516 1 D14 ckt 11 Supp r ession Pool Suction Valve M0-2517 1014 ckt 12 Gland Seal Vacuum Pump 1P-227 1 D14 ckt 13 Gland Seal Vacuum Tank Condensate Pump 1 P-228 I AOP 302.1 Page 27 of 63 Rev. 58 (1)-EOF DAEC EMERGENCY PLAN EMERGENCY COMMUNICATIONS FIGURE F-5 DAEC TELEPHONE SYSTEMS ----------

---1 I r----, I I I I Operational Support Cen t er (Access ControQ Sa tellite Communicat i ons .,,. .... .,,. .... .,,. ........ .,,. .... ,--------, Room ........ .,,.

emergency unltS l ed i nes (Bl u e P h o n es) i n Contro l Room , T SC , CAS , SAS Local Telephone Company Central Office DAEC PBX R oom DAEC M i crowave t o A llia n t Tower 'Alliant Tower Microwave to D A EC Offs tt e Laboratory and Decontamin at i o n C e n ter

FPL E Duane Arnold Corporate Offices

County Sheriff's Offices

Palo F r e D epartm e n t

Mercy Hospl1I 'state Highway Patrol

Sta t e E mNgeney Management DMsfon

UniYersty of Iowa *N R C 'DOE 'FEMA

L i m C:,unty Emergen cy Management

Be n tor County Emergency Management Nonn1I Telephone S e rvices Qwest (Cedar Ra p ids) To Ot ier Bel C e n tr a l Offices EOF Emergency Ope r ations Fac i l i ty (1) Denotes a Ded i cated Line Joint Public Information Center SECTION 'F' Rev. 29 Page 15 of 17

,----------

DAEC EMERGENCY PLAN SECTION 'F' Rev. 29 EMERGENCY COMMUNICATIONS Page 16 of 17 FIGURE F-6 FEDERAL TELEPHONE SYSTEM (FTS-2001)

NRC ERDS EOF Lo ca l TSC (D ENS © HPN @ RSCL © P M C L @ MCL DAECEMERGENCYPLAN SECTION 'F' Rev. 29 EMERGENCY COMMUNICATIONS Page 17 of 17 FIGURE F-7 ALL-CALL TELEPHONE SYSTEM LINN COUNTY EOC LINN COUNTY Sheriffs Office Backup facility Main Facility .___ _________ _J---~ Sm!!!lt DAECEMERGENCYPLAN SECTION 'E' Rev. 23 NOTIFICATION METHODS AND PROCEDURES Page 3 of 7 1.0 PURPOSE (1) This section descr i bes the methods and procedures used by FPLE Duane Arnold to transmit emergency information to the Emergency Response Organization , local and state authorities , and subsequently , from such authorities to the public. Details required in the initial and follow-up message are described , along with a description of the types of news statements that will be used to provide the public with information and protective actions. 2.0 REQUIREMENTS (1) Methods used to accomplish notification of the Emergency Response Organization include the use of call lists contained in the Emergency Telephone Book , pager and automated telephone callout process. (2) The Emergency Telephone Book includes phone numbers and pager numbers (where applicable) of emergency response personnel who may be required to respond to an emergency condition. It also includes the 24-hour telephone numbers of local , state , and federal support agencies including the NRC. The NRC would normally be notified using the NRC ENS Telephone (FTS-2001 System) from the Control Room. The state and counties would normally be notified by dedicated microwave telecommunications link. 2.1 INITIAL NOTIFICATION (1) After declaration of an emergency condition , the Operations Shift Manager/ Supervisor will ensure that the following personnel and agencies are notified:

Linn and Benton Counties State of Iowa NRC Operations Center

Emergency Coordinator

Emergency Response and Recovery Director

NRC Res i dent Inspectors (2) Verification of Notification (a) The authenticity of initial notifications provided to Linn and Benton Counties and the State of Iowa do not require verification if the notification is made by the dedicated phone system. (b) Local and state agencies notified by commercial communication system (telephone or facsimile) may require ver i fication of the identity and authenticity of the caller and the message received.

BASES-DAEC EOP BASES DOCUMENT BREAKPOINTS Rev. 14 EOP BREAKPOINTS Page 7 of 14 BREAKPOINTS FOR REACTOR LEVEL CONTROL Page 1 of 2 RPV Level Item of Interest Significance (inches) +211 High Level Trip Setpoint, ¢ Loss of high pressure injection (FW, Main Turbine Trip HPCI , RCIC) ¢ Loss of 100% Heat Sink +170 Low Water Level Scram, ¢ RPS defeats needed in A TWS PCIS Groups 2, 3 , 4 ¢ Containment Isolation, Isolations

¢ Shutdown Cooling Valves Close ---+119.5 High Pressure Injection , ¢ HPCI/RCIC Auto lnitiat i'ori} PCIS Group 5 Isolation , ¢ RWCU Isolation ARI ARI Initiation

& Recirc Pump A TWS Trip ¢ +87 Two Feet Below During ATWS if power >5% or unknown , Feedwater Sparger lower level to +87 inches to reduce core inlet subcooling

+64 ECCS Auto Start, ¢ ADS Timers start PCIS Group 1 Isolation

¢ CS/RHR Auto Initiation MSIVs close and result in loss of main condenser

+15 Top of Active Fuel (TAF) ¢ Loss of Adequate Core Cooling (ACC) (Note 1) through core submergence

¢ If no preferred Injection Subsystem is available, maximize injection with Alternate Injection Systems in EOP 1 when level < +15" Note 1: +15 inches is used for TAF than O inches for the following reasons:

To allow monitoring RPV level on the Wide Range instrumentation

-prevents risk of uncovering the core if using Fuel Zone instruments.

Fuel Zone instruments use the same tap as jet pump instrumentation and any flow through the jet pumps including LPCI flow will cause the Fuel Zone instruments to read high.

Emergency Preparedness Program Frequently Asked Question (EPFAQ) EPFAQ Number: Originator:

Organization:

Relevant Guidance:

2016-002 David Young NEI NEI 99-01, Methodology for Development of Emergency Action Levels, Revisions 4 and 5; and NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors , Revision 6. NUMARC/NESP-007 , Methodology for Development of Emergency Action Levels. Applicable Section(s):

Initiating Condition (IC) HA2 in NEI 99-01 , Revisions 4 and 5 , and NUMARC/NESP-007 , " FIRE or EXPLOSION Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown" Status: NOTE: ICs CA6 and SA9 in NEI 99-01, Revision 6: " Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode" Definition of VISIBLE DAMAGE in NEI 99-01, Revisions 4, 5 and 6 , and NUMARC/NESP-007 Complete Based on NRC staff consideration of industry comments provided by letter dated February 16 , 2017 (ADAMS Accession No. ML 17079A228), a revision to these /Cs was proposed at the public meeting held on April 4 , 2017. These changes were attached to the public meeting notice (ADAMS Accession No. ML 17089A458). Based on comments provided by the industry during the April 4 , 2017 public meeting , the NRC staff revised the proposed revisions to these /Cs. QUESTION OR COMMENT: A review of industry Operating Experience has identified a need to clarify an aspect of the definition of VISIBLE DAMAGE as it relates to the I Cs cited above; adding this clarity is necessary to minimize the potential for an over-classification of an equipment failure. There may be cases where VISIBLE DAMAGE is the result of an equipment failure and limited to the failed component (i.e., the failure did not cause damage to any other component or a structure). The current definition of VISIBLE DAMAGE does not adequately differentiate between damage resulting from , and affecting only , the failed piece of equipment vs. an equipment failure causing damage to another component or a structure (e.g., by a failure-induced fire or explosion). Can the definition of VISIBLE DAMAGE be clarified to help avoid an inappropriate emergency declaration in cases where an equipment failure does not result in damage to another component or a structure (i.e., VISIBLE DAMAGE affects only the failed component)?

A related question is also posed -Consistent with the approach used in other ICs, should a note be added to preclude an emergency declaration if the safety system affected by a hazard was not functional before the event occurred (e.g., tagged out for maintenance)?

PROPOSED SOLUTION:

Yes; the sentence below may be added to the definition of VISIBLE DAMAGE [as defined in NEI 99-01 , Revisions 4 , 5 , and 6]. Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) i s not VISIBLE DAMAGE. From a plant safety and change-in-risk perspective , the consequences from the failure of a 1 Emergency Preparedness Program Frequently Asked Question (EPFAQ) piece of equ i pment , accompanied by a hazard (e.g., a fire or explosion) that does not damage any other equipment or a structure , are essentially the same as the equipment failing with no attendant hazard. Neither event would appear to meet the definition of an Alert because the outcome does not involve an actual or potential substantial degradation of the level of safety of the plant (e.g., there has been no significant reduct i on in the margin to a loss or potential loss of a fission product barrier). Nuclear power plants are designed with redundant safety system trains that are required to be separated (i.e., installed in separate plant areas or have separation w i thin an individual area). Absent any collateral damage to another component or a structure , a hazard associated with an equipment failure does not affect the ability to protect public health and safety , and there is no additional response benefit to be gained by declaring an emergency.

The normal plant organization has sufficient resources and adequate guidance to respond to an equipment failure -guidance includes operating procedures and Technical Specifications

the fire protection

[program], industrial safety and corrective action programs; and work management and maintenance requirements. Concerning the second question , an emergency declaration would not be appropriate in r esponse to a hazard affecting a piece of equipment or system that was non-functional prior to the event (e.g., tagged out for maintenance). For this reason and consistent with the approach used in other I Cs , the following note may be added to IC HA2 (NEI 99-01 R4 and R5), or ICs CA6 and SA9 (NEI 99-01 R6). Note: If the affected safety system (or component) was already non-functional before the event occurred , then no emergency classification is warranted. Consistent with the guidance in Regulatory Issue Summary (RIS) 2003-18 , Supplement 2 , Use of Nuclear Energy Institute (NE/) 99-01 , " Methodology for Development of Emergency Action Levels ," Revision 4 , dated January 2003 , it is reasonable to conclude that the changes proposed above would be considered as a " deviation." NRC RESPONSE:

The proposed guidance 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 a Notification of Unusual Event (NOUE) 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 reliabil i ty 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 guidance 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. IC HA2 (NEI 99-01 R4 and R5; NUMARC/NESP-007), or ICs CA6 and SA9 (NEI 99-01 R6), do not directly escalate to a Site Area Emergency or a General Emergency due to a hazardous event. The F i ssion 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 proposed addition of the following notes , applicable to I Cs HA2 (NEI 99-01 R4 and R5; NUMARC/NESP-007), or ICs CA6 and SA9 (NEI 99-01 R6), provide further clarification as to how these Alert emergency classifications are considered.

The revisions to these EALs , 2 Emergency Preparedness Program Frequently Asked Question (EPFAQ) including the addition of the notes, are consistent with the current NRC-endorsed Alert classification language. 1. Adding the following note to the applicable EALs , per this EPFAQ, is acceptable as it meets the intent of the EALs , is consistent with other EALs (e.g., EAL HAS from NEI 99-01 , Revision 6; this revision was endorsed by the NRC in a letter dated March 28 , 2013 , available at ADAMS Accession No. ML 12346A463), and ensures that declared emergencies are based upon unplanned events with the potential to pose a radiological risk to the public. If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred , then this emergency classification is not warranted. 2. Adding the following note to help explain the EAL is reasonable to succinctly capture the more detailed information from the Basis section related to when conditions would require the declaration of an Alert. If the ha z ardous 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. Revising the EALs and the Basis sections to ensure potential escalations from a NOUE 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. Revising the definition for VISIBLE DAMAGE is appropriate as this definition is only used for these EALs and the revised EALs are based upon SAFETY SYSTEM trains rather than individual components or structures. All of the changes discussed above are addressed in the attached markups to NEI 99-01 , Revision 6. Licensees that use NESP-007 , NEI 99-01 Revision 4 , or NEI 99-01 Revision 5 EAL schemes can adopt this language in the relevant format the staff approved for their use. Consistent with the guidance in Regulatory Issue Summary (RIS) 2003-18 , Supplement 2 , Use of Nuclear Energy Institute (NE/) 99-01 , " Methodology for Development of Emergency Action Levels ," Revision 4, dated January 2003 , a licensee's scheme change based on this EPFAQ should be considered as a " deviation" because a classification based on NRC-endorsed industry guidance in NEI 99-01 , Revisions 4 , 5 and 6 , as well as in NUMARC/NESP-007 , could be different from a classification based on this EPFAQ. RECOMMENDED FUTURE ACTION(S):

0 INFORMATION ONLY , MAINTAIN EPFAQ [gj UPDATE GUIDANCE DURING NEXT REVISION 3

Emergency Preparedness Program Frequently Asked Question (EPFAQ) CA6 ECL: Alert Initiating Condition:

Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode. Operating Mode Applicability:

Cold Shutdown , Refueling Example Emergency Action Levels: Notes:

If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred , then this emergency classification is not warranted.

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. (1) Basis: 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 Sh i ft Manager AND b. 1. Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode. AND 2. EITHER of the following:

Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode , or

Event damage has resulted in VISIBLE DAMAGE to the second train of a SAFETY SYSTEM needed for the current operating mode. 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 class i fied , 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 4 Emergency Preparedness Program Frequently Asked Question (EPFAQ) such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance for criteria 1.b.1 of th i s EAL; 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 address damage to a SAFETY SYSTEM train that is in service/operation since i ndications 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. Operators will make a determination of VISIBLE DAMAGE based on the totality of available event and damage report i nformation. This is intended to be a brief assessment not requiring lengthy analysis or quant i fication 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 AS1. Developer Notes: For (site-specific hazards), developers should consider including other significant , site-specific hazards to the bulleted list contained in EAL 1.a (e.g., a seiche). Nuclear power plant SAFETY SYSTEMS are comprised of two or more separate and redundant trains of equipment in accordance with site-specific design criteria. ' EGL Assignment Attributes:

3.1.2.B 5 Emergency Preparedness Program Frequently Asked Question (EPFAQ) SA9 ECL: Alert Initiating Condition:

Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode. Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Example Emergency Action Levels: Notes:

If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then this emergency classification is not warranted.

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. (1) Basis: 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. 1. Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode. AND 2. EITHER of the following:

Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode, or

Event damage has resulted in VISIBLE DAMAGE to the second train of a SAFETY SYSTEM needed for the current operating mode. 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 6 Emergency Preparedness Program Frequently Asked Question (EPFAQ) such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance for criteria 1.b.1 of this EAL; 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 address damage to a SAFETY SYSTEM train that is in service/ope r ation 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. 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 ICs FS1 or AS1. Developer Notes: For (site-specific hazards), developers should consider including other significant , site-specific hazards to the bulleted list contained in EAL 1.a (e.g., a seiche). Nuclear power plant SAFETY SYSTEMS are comprised of two or more separate and redundant trains of equipment in accordance with site-specific design criteria. ECL Assignment Attributes

3.1.2.B 7 Emergency Preparedness Program Frequently Asked Question (EPFAQ) 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. 8 EPFAQ Number: Originator:

Organization:

Relevant Guidance:

Applicable Section(s):

2018-04 David Young NEI This question concerns NEI 99-01 , Development of Emergency Action Levels for Non-Passive Reactors , Revision 6 and EPFAQ 2016-002. Initiating Conditions (ICs) CA6 and SA9 , and the associated Emergency Action Levels (EALs) and Bases Date Accepted for Review: 5/31/2018 Status: Under Review QUESTION OR COMMENT: Background EPFAQ 2016-002 provided guidance intended to 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. In responding to the EPFAQ , the staff determined that revising the EALs and the Basis sections of ICs CA6 and SA9 would be appropriate to ensure potential escalations from a NOUE to an Alert , due to a hazardous event , occur when there is: (1) a hazardous event , and (2) one SAFETY SYSTEM train having performance issues as a result of the hazardous event, and (3) either the second SAFETY SYSTEM train is having performance issues or VISIBLE DAMAGE sufficient to be concerned that the second SAFETY SYSTEM train may have operability or reliability issues. The response to EPFAQ 2016-002 works well for situations involving a safety system with two trains (a typical configuration)

however, industry operating experience indicates that additional clarification is needed for three other cases as described in the questions below. Because this EPFAQ is based on material in EPFAQ 2016-002 , the response to this EPFAQ may be considered only by sites that have implemented EPFAQ 2016-002 in a manner approved through an NRC Safety Evaluation Report (SER). Question Concerning ICs CA6 and SA9 , how should an event leading to indications of degraded performance and/or VISIBLE DAMAGE be classified when: 1. The event affects equipment common to two or more safety systems or safety system trains? For example , a unit with a tank that is the water source for multiple safety injection systems or trains , such as a Refueling Water Storage Tank (RWST). 2. The event affects a safety system that has only one train. For example , a Boiling Water Reactor (BWR) unit with a single-train Reactor Core Isolation Cooling (RCIC) or High-Pressure Coolant Injection (HPCI) system. 3. The event affects two trains of a safety system having more than two trains. For example, a unit that has an Auxiliary/Emergency Feedwater system with three trains. 1 05/21/2018 Emergency Preparedness Program Frequently Asked Question (EPFAQ) PROPOSED SOLUTION:

The following answers to the above questions are proposed: 1. An event affecting equipment common to two or more safety systems or safety system trains (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the common equipment) should be classified as an Alert under CA6 or SA9, as appropriate to the plant mode. By affecting the operability or reliability of multiple system trains , the loss of the common equipment effectively meets the two-train impact criteria that underlie the EALs and Bases. 2. An event affecting a single-train safety system (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the one train) would not be classified under CA6 or SA9 because the two-train impact criteria that underlie the EALs and Bases would not be met. If an event affects a single-train safety system , then the eme r gency classification should be made based on plant parameters/symptoms meeting the EALs for another IC. Depending upon the circumstances , classification may also occur based on Shift Manager/Emergency Director judgement.

3. An event that affects two trains of a safety system (e.g., one train has indications of degraded performance and the other VISIBLE DAMAGE) that also has one or more additional trains should be classified as an Alert under CA6 or SA9 , as appropriate to the plant mode. This approach maintains consistency with the two-train impact criteria that underlie the EALs and Bases , and is warranted because the event was severe enough to affect the operability or reliability of two trains of a safety system despite plant design criteria associated with system and system train separation and protection.

Such an event may have caused other plant impacts that are not immediately apparent.

As stated above, this EPFAQ may be considered only by sites that have implemented EPFAQ 2016-002 in a manner approved through an NRC Safety Evaluation Report (SER). With this proviso met , the response to EPFAQ 2018-004 would then provide clarification of expected emergency classifications for cases not explicitly addressed by ICs CA6 and SA9 (from NEI 99-01 , Revision 6), and EPFAQ 2016-002; therefore , implementation of the guidance in this EPFAQ would improve the accuracy and timeliness of a classification following a hazardous event affecting a safety system. Moreover , the answers provided in EPFAQ 2018-004 would result in EAL interpretations that are consistent with the meaning and intent of NRC-approved EAL bases such that the classification of the event would not be different from that approved by the NRC in a site-specific application. For this reason , it is reasonable to conclude that incorporation of the guidance from this EPFAQ into an NRC-approved site-specific scheme reflecting the guidance in EPFAQ 2016-002 would be considered a " difference" in accordance with Regulatory Issue Summary (RIS) 2003-18 , Supplement 2 , Use of Nuclear Energy Institute (NE/) 99-01 , " Methodology for Development of Emergency Action Levels ," Revision 4 , dated January 2003. This " difference" determination is contingent upon incorporating any or all of the three answer statements (as applicable to a facility) verbatim; any change to the scope or intent of the answers would make incorporation into a site-specific scheme a " deviation" per RIS 2003-018 , Supplement

2. 2 05/31/2018 Emergency Preparedness Program Frequently Asked Question (EPFAQ) NRC RESPONSE:

RECOMMENDED FUTURE ACTION(S):

0 INFORMATION ONLY , MAINTAIN EPFAQ UPDATE GUIDANCE DURING NEXT REVISION 3 05/31/2018 BASES-DAEC EOP BASES DOCUMENT BREAKPOINTS Rev. 14 EOP BREAKPOINTS Page 7 of 14 BREAKPOINTS FOR REACTOR LEVEL CONTROL Page 1 of 2 RPV Level Item of Interest Significance (inches) +211 High Level Trip Setpoint ,

Loss of high pressure injection (FW , Main Turbine Trip HPCI, RCIC)

Loss of 100% Heat Sink +170 Low Water Level Scram ,

RPS defeats needed in ATWS PCIS Groups 2 , 3 , 4

Containment Isolation , Isolations

Shutdown Cooling Valves Close +119.5 High Pressure Injection ,

HPCI/RCIC Auto Initiation PCIS Group 5 Isolation ,

RWCU Isolation ARI ARI Initiation

& Recirc Pump ATWS Trip * +87 Two Feet Below During ATWS if power >5% or unknown, Feedwater Sparger lower level to +87 inches to reduce core inlet subcooling

+64 ECCS Auto Start ,

ADS Timers start PCIS Group 1 Isolation

CS/RHR Auto Initiation MSIVs close and result in loss of main condenser

+15 Top of Active Fuel (TAF)

Loss of Adequate Core Cooling (ACC) (Note 1) through core submergence

If no preferred Injection Subsystem is available , maximize injection with Alternate Injection Systems in EOP 1 when level < +15" Note 1: +15 inches is used for TAF than O inches for the following reasons:

To allow monitoring RPV level on the Wide Range instrumentation

-prevents risk of uncovering the core if using Fuel Zone instruments.

Fuel Zone instruments use the same tap as jet pump instrumentation and any flow through the jet pumps including LPCI flow will cause the Fuel Zone instruments to read high.

SAG~ HYDROGEN CONTROL CAUTIONS 0 hsuction Operatio n of HPC I , RC I C , Core Sp r ay , or RHR wit from the torus and pump n ow~ the NPSH or v ortex l i mit may damage equipment

0 ntrations H2 and 02 instruments may Indicate higher conee than actua ll y ex is t ins ide the containment follCl'Ning break L OCA due to moistu,e condensation In the s l i nes. During the 24 hr period fol l O'Mng a l arge brea the monitOfS shou l d om be Independently used to operationa l decisions but may be used for trendng Detail D I Norma l R e l e a se R a t e Limit s A determination that the offsite release rate Is below normal limits may be made bi/ either:

Containment Atmosphere Radation Monitor GASEOUS Channels ~cale and operable. -MonitoredlocalyonRJT-8102A/Bat1C-219A/B or RR4379A/B at IC-29 (blLw CNlnnel). , Containment sam?e (PASAP 7.4). Detail E I Spray S ourc es

RHR (01149) 3

RHR SeNice Water (AIP 401) , Fire System (AIP 404)

Wei Water (AIP 403)

GSW (AJP 403)

ESW (AIP 402) , Condensate Service Water (AlP 405) a l arg e ample klOCA , support I ( START ) While i n t h i s p r o c e dur e: I F H 2 m. mooitor I s unavailable Drywel l pressure drops below 2.0 psig Torus pressure drops below 2.0 psig H-1 DRYWELL I F ....... offsite releau rate Is expected to stay bfHow normal limits (DetailD), THEN .vent and purge the primary containment:

OK m defeat isolation*~

high radiation (Defeat 9). -If pneumatic supplies are unava~able , use $AMP 706 , Venting the Primary ContaiMl9nt Follo'Ning Lou of Pnet.matic SuPP)'. 1. Vent as follOW&'

IF ....... torus water tevel is bd2t£ 16 ft, THEN .vent the drywel through themn&(SEP 301.1).

IF ....... toruswaterleveUs~

16ft. OR. .... the torus cannot be vented , THEN .. vent directly from the s1caadl. (SEP 301.2). 2. I F ........ the primary con t a i nment can be vented , T HEN ... purge t h e~ with~ using N 2 purge (SEP 303.2). 3. Stop the vent and purge (If not req u ired by o t her SAG steps) when:

Hydrogen la no l onger detected in the drywe ll , OR

Offsite release rate reaches norma l limits (D etail 0). I F ....... offsite r e l ease rate is expected to stay below Genera l Emergency Leve l s (EAL RG1), OR ..... RPV water levef "11Cm be mainta in ed above +15 in. (TAF), THEN .. vent and purge the primary containment OK m defeat&ll isolations (O.faat 10). -If pneumatic supplies are una-aiable , use SAMP 706 , Venting the Primary Containment Following Lo u of Pneumatic Suppty. 1. Vent as followa:

I F... . t orus water l evel i s~ 16 ft , T HEN .ven t t he drywe ll through the 1'llJ,ll (SEP 30 1.1).

I F ....... t orus wate r l eve l is~ 16 ft , OR ..... the torus canno t be vented , THEN .. vent direc tl y from the (SEP 301.2). 2. IF ........ the primary containment can be vented , THEN ... purge the doctld, v.1th at max flow using N 2 purge (SEP 303.2}. 3. Stop the ve nt and purge (if not re"'1 ired by other SAG steps) when:

Hydrogen Is no longer det ecte d in either the drywel or the torus , OR

Hydrogen Is no longer detected In the drywe H J.O.d. drywell oxygen la less than 5%, OR

RPV water l eve l can be maintained above +15 in. (TAF) lad offs i te r elease rate reaches a General Emergency Leve l (EAL RG1). Vent and purge the primary cootainment OK to defeat Ill Iso l atio n s (Oefea l 10). OK to exceed re l ease rate limits. I f pneuma tic suppliH are unava i lable , use SAMP 706 , Venting the Primary Contalm,ent Folowing Loss of Pneumatic Supply. 1. Vent as folows:

IF ....... torus water l ev .. la bmfttt 16 ti , THEN .vent the dryweN through the twm (SEP 30 1.1).

IF ....... toruswaterleveHs~

16ft, OR ..... the torus cannot be vented , THEN . vent directly from the (S EP 301.2). 2. I F ....... the primary containment can be vented , T HEN .purge the~ v.1th at max flow. Use whicheve r method wi ll r e d uce hydrogen be l ow 6% or oxygen below 5% faster:

Air purge (SEP 303. 1)

N;i purge (SEP 303.2) 3. I F...... permitted by SAG-1 , THEN .. operate drywe ll sprays (Detail E). TH EN Notify Chemistry t o manuany sample the drywetl and torus for H 2 and~ (PASAP 2.6). Verify conta i nment sp,ays Isolate. T erm i nate torus sprays. T ORUS I F ....... offsite release rate is expected to stay below normal limits (Detail 0), THEN .. vent and purge the primary containment H-7 OK m defeat isolation*

uam high radiation (Defeat 9). -I f pneumatic supplies are unavailable , use SAMP 706 , Venting the Primary Conta i Ml9nt Fo l lo'Ning loss of Pneuma tic Supply. 1. Vent as folows:

IF .....*. torus water l evel is bd2t£ 16 ft, THEN ... vent directly from the 1s2wl (S EP 301.1).

IF ........ the torus cannot be vented , AND .... t o rus water l evel is 13.5 ft , THEN ... vent the torus t hrough the~ (SEP 301.2). 2. I F ........ the primary conta i nment can be vented , THEN ... purge the 1m.l.li with~ using N 2 purge (SEP 303.2). 3. Stop t he vent an d purge (i f not required by othe r SAG steps) when:

Hydrogen is no l onger detected in the torus , OR

Offsite release rate reaches normal limits (Detail 0). I F ....... offs i te ,eleaae rate Is expected to stay be l ow Genera l Emergency leve l s (EAL RG 1), OR ..... RPV water l evef an.nm be maintained above +15 in. (TAF), THEN .. vent and pu r ge the primary containment OK m defeat &II isolations (Defeat 10). -II pneumatic supplies are unavailable. uae SAMP 706 , Venting the Primary Conta inme nt Folowing loS5 of Pneuma tic Supply. 1. Ventasfol l ov.-s:

I F ....... torus water l evel is~ 1 6 n , T H EN .. vent directly f rom the 1l2Wi (SEP 301.1).

I F ....... the t o r us canno t be vented , AND .. t orus water l eve l is 1 3.5 ft, THEN .. vent the t orus t hrough the .d!Y:m.D. (SEP 301.2). 2. I F ......*. the primary containment can be vented , THEN ... purge the 1m.l.ll. v.1th at max flow using N 2 purge (S EP 303.2). 3. Stop the vent and purge (if not re quired by other SAG steps) when.

Hydrogen Is no longe r detected i n either the drywell or the t orus , OR

Hydrogen is no l onger detected I n the torus j.[lg, torus oxygen is less than 5%, OR

RPV water l eve l can be m aintained above +15 I n. (TAF) j.[lg, offsite release , ate r eaches a General Emergency Level (EAL RG1). Vent and purge the primary conta i nment --OK to de!eat all isolations (Defeat 10). _.. OK t o exceed re l ease , ate Um i ts. -I f pneumatic supplies are una v ailable , use SAMP 706 , Venting the Primary Con ta inme nt Fo l lo'Ning Loss of Pneumatic Supply. 1. IF ....... permitted by SAG-1 , THEN ... operate torus sprays (Detail E). 2. Vent as follows.

IF ..... t o ruswater l evelis~16tl , THEN. vent directly from the 1m.l& (S EP 301.1 ).

1 F ....... the tonts cannot be vented , AND.. torus water l eve l is~ 13.5 ft , T HEN .. vent the torus through th e~ (SEP 301.2). 2. I F ....... the pr i mary containmen t can be vente d , THEN .. purge the m1.Wl with~ at max f\ow. Use whi c hever method wi ll r educe h ydrogen below 6°,4 or oxygen below 5'"' faster:

Air purge (SEP 303.1)

purge (SEP 303.2) DUANE ARNOLD ENERGY CENTER SAG-3 HYDROGEN CONTROL REV. 6 Contr;i'led Copy .... I EOP3 SECONDARY CONTAINMENT CONTROL Ar eatemp,e,al ure Area radiation level AruwaterleYfl F~ Poo l e.vi aust Ft1S4131A(8]H IH1 Radle¥ela IIOOl'e ln<<lhr RB Vent Shalt Rad Mon itor Rllll7l506A(B]Radlevet.

at)Ote lmR/hf IIDOve Mu Mo,n,ai Op,tnl!l1111 Umt abo<re Max Normal Operating Umtt IIOOl'a Mu Norn,al Op.,.all1111 Limit ..... OllgNv.rtf'lpa HI-Hllrlp Tablal5 {Tatilel5 Tab6el5 F ... Pl:IOl~RIS11'1A{81R-.....LaWa...,..

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..... ENT TORUSCATVIIW(.EA!TOIFF TORUS CATVIIW(.

YEST Off' TORUSCATVIIW(.!!ICUnMEST DFf' TORUS CATVIIW(.

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...... i....._."""poo1-D ..... ....,altld_.......

_ __,._....

Seco nd ary Containment Limits TRl'T'CR~Clil TRl'T'DR~C,,2 Tl'lfTDR2CCIDBCIII TAIT'DR2CCIDBCll2

=~;", l RITOR~C,,11']

TRITOA2C2SA{BIC,,1 TRITOA2425AC,,2 TRITOR2425A(BIC,,1 Tll'TOR2C25o'1Clll TWTtlll:24258Cll2 fWTtlll:2225#!,C,,)

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...........

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1.2.7

HSM Dose Rates with a Loaded 24P , 52B or 61 BT DSC Limit/Specification

Applicability

Objective: Action: 61BT DSC Dose Rate Thresholds

= 2 X TS limits Therefore:

3 feet from HSM Surface = 800 mrem/hr Outside HSM Door -Centerline of DSC = 200 mrem/hr End Shield Wall Exterior =40mrem/hr Surveillance

Basis: Dose rates at the following locations shall be limited to levels which are less than or equal to: a. 400 mrem/hr at 3 feet from the HSM surface. b. Outside of HSM door on center line of DSC 100 mrem/hr. c. End shield wall exterior 20 mrem/hr. This specification is applicable to all HSMs which contain a loaded 24P, 52B or 61BT DSC. The dose rate is limited to this value to ensure that the cask (DSC) has not been inadvertently loaded with fuel not meeting the specifications in Section 1.2.1 and to maintain dose rates as-low-as-is-reasonably achievable (ALARA) at locations on the HSMs where surveillance is performed , and to reduce s i te exposures during storage. a. If specified dose rates are exceeded , the following actions should be taken: 1. Ensure that the DSC is properly positioned on the support rails. 2. Ensure proper installation of the HSM door. 3. Ensure that the required module spacing is maintained. 4. Confirm that the spent fuel assemblies contained in the DSC conform to the specifications of Section 1.2.1. 5. Install temporary or permanent shielding to mitigate the dose to acceptable levels in accordance with 10 CFR Part 20 , 10 CFR 72.104(a), and ALARA. b. Submit a letter report to the NRC within 30 days summarizing the action taken and the results of the surveillance , investigation and findings.

The report must be submitted using instructions in 10 CFR 72.4 with a copy sent to the administrator of the appropriate NRC regional office. The HSM and ISFSI shall be checked to verify that this specification has been met after the DSC is placed into storage and the HSM door is closed. The basis for this limit is the shielding analysis presented in Section 7.0 , Appendix J , and Appendix K of the FSAR. The specified dose rates provide as-low-as-is-reasonably-achievable on-site and off-site doses in accordance with 10 CFR Part 20 and 10 CFR 72.104(a). Certificate of Compliance No. 1004 A-78 Amendment No. 9 , Revision No. 1 Development of EAL Threshold values from NEE-323-CALC-001 Calculated values were added to the typical background readings of these monitors, and then rounded to aid in evaluator use of the EALs. Resulting values used in the DAEC Fission Product Barrier chart are shown below:

Fuel Clad Barrier: o Fuel Clad Barrier LOSS 4.A = Drywell Monitor (9184A/B) reading greater than 2000 R/hr. o Fuel Clad Barrier LOSS 4.B = Torus Monitor (9185A/B) reading greater than 200 R/hr.

RCS Barrier: o

RCS Barrier LOSS 4.A = Drywell Monitor (9184A/B) reading greater than 5 R/hr after reactor shutdown. (minimum serviceable threshold value accounting for scale of monitor) o

Calculated Torus Monitor (9185A/B) response is below scale of monitor and not used.

CTMT Barrier: o

CTMT Barrier LOSS 4.A = Drywell Monitor (9184A/B) reading greater than 5000 R/hr. o

CTMT Barrier LOSS 4.A = Torus Monitor (9185A/B) reading greater than 500 R/hr.

BASES-DAEC EOP BASES DOCUMENT BREAKPOINTS Rev. 14 EOP BREAKPOINTS Page 12 of 14 BREAKPOINTS FOR PRIMARY CONTAINMENT PRESSURE CONTROL Pressure Item of Interest Significance (psig) 53 Primary Containment When PCPL is reached, containment venting (Torus) Pressure Limit (PCPL) is required.

-21.4 Pressure Suppression Pressure Suppression Pressure exceeded for (Torus) normal torus level >11 Drywell Sprays Drywell sprays may be initiated if drywell (Torus) parameters are within the Drywell Spray (11.15) Initiation Limit and torus level is less than 13.5 feet 11.4 Drywell Spray Initiation Above 11.4 psig drywell pressure, drywell (Drywell)

Limit (DWSIL) Break spray initiation is unrestricted by the DWSIL. Point <11 Torus Spray Initiation Start torus sprays prior to 11 psig, if possible. (Torus) Pressure If pressure is exceeded before torus sprays (11.15) are initiated

-initiate them anyway 2 Drywell High Pressure ECCS Initiation , Isolations and RPS defeats (Drywell)

Scram Setpoint may be needed , EOP 1 and EOP 2 entry 1 Drywell N2 Makeup Drywell N2 makeup supply isolates if drywell (Drywell)

Isolation pressure exceeds 1 psig EOP3 SECONDARY CONTAINMENT CONTROL F~PoolE.1.t11u11 RIS4131A(B]

H1 HI Redl.,N M>IMll lmRlhr RB Vent Shatt Rad Mon itor RIM11'iOM{BIRldLIYMI aborie lmlllh, Area lempe,ature Area 1ad l 1tlon level AIH water 1,evel at,o,,e Mu Nomwl Op,tn,tlng Utnlt 1boYe Mu Notffill ap.ratlng Llrrit at,o,,,e Mu N on N I 0pa ... 1 1 no Limit CAUTIONS F F\1111Pool~~RIS41)1A{8!lt--.L

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.........

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....... ENT ""'""""""""'

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IICICllt-Affl RCIC Et.ER COOLER AM81ENT """"°""""""'

Rae R<X>M CIFFERENTW.

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BASES-DAEC EOP BASES DOCUMENT BREAKPOINTS Rev. 14 EOP BREAKPOINTS Page 12 of 14 BREAKPOINTS FOR PRIMARY CONTAINMENT PRESSURE CONTROL Pressure Item of Interest Significance (psig) 53 Primary Containment When PCPL is reached , containment venting (Torus) Pressure Limit (EQ PL) is required.

-21.4 Pressure Suppression Pressure Suppression Pressure exceeded for (Torus) normal torus level >11 Drywell Sprays Drywell sprays may be initiated if drywell (Torus) parameters are within the Drywell Spray (11.15) Initiation Limit and torus level is less than 13.5 feet 11.4 Drywell Spray Initiation Above 11.4 psig drywell pressure , drywell (Drywell)

Limit (DWSIL) Break spray initiation is unrestricted by the DWSIL. Point <11 Torus Spray Initiation Start torus sprays prior to 11 psig , if possible. (Torus) Pressure If pressure is exceeded before torus sprays (11.15) are initiated

-initiate them anyway 2 Drywell High Pressure ECCS Initiation , Isolations and RPS defeats (Drywell)

Scram Setpoint may be needed, EOP 1 and EOP 2 entry 1 Drywell N2 Makeup Drywell N2 makeup supply isolates if drywell (Drywell)

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I AOP 901 EARTHQUAKE PROBABLE ANNUNCIATORS None PROBABLE INDICATIONS 1C35 -The amber DESIGN BASIS EARTHQUAKE (DBE) light is ON. -The amber OPERATING BASIS EARTHQUAKE (OBE) light is ON. -The amber .01 G RECORDERS RUNNING l i ght is ON. -The white CONTINUITY light is OFF. -The Seismic Wailing Alarm is sounding. -Building vibration. A Cooling Tower Valve House -No power indicating l i ght is operable. I AOP 901 Page 11 of 16 Rev. 30 I AOP 901 EARTHQUAKE

.................

............................................. INFORMATION**

- - - - .................................................... . Earthquake OBE DBE Ground Acceleration 0.06g 0.12g I AOP 901 Page 13 of 16 Rev.30 AOP 915 SHUTDOWN OUTSIDE CONTROL ROOM SECTION 1 I TRANSFER OF CONTROL TO THE REMOTE SHUTDOWN PANEL CONDITIONAL ST A TEMENTS IF while performing this procedure: IF Control Room access is regained AND personnel are available THEN when directed by the Emergency Response and Recovery Director resume control of unaffected components from the Control Room NOTE AND maintain control of Division II components from 1 C388 until operability of Control Room instruments, indications and controls has been verified.

Operations personnel evacuate to the Remote Shutdown Panel except: the STA , Shift Communicator , and on-site personnel not on shift evacuate to the TSC.

The preferred evacuation route to the Remote Shutdown Panel is out the back door of the Control Room, and down the stairs. Emergency lighting is provided for this path.

The alternate evacuation route to the Remote Shutdown Panel is out the front door of the Control Room, and down the stairs to access control. Emergency lighting is provided for this path.

Since fire induced failure in 1 C05 could adversely affect manual scram circuits , the initiation of A TWS ARI/RPT provides a redundant and diverse means of control rod insertion. CAUTION For Control Room evacuation as the result of a fire , transfer of control at panels 1 C388 , 1 C389 , 1 C390 , 1 C391, 1 C392 is required to be completed within 20 minutes. I AOP 915 Page 4 of 94 Rev. 571 V -------------------**. 3.4 REACTOR COOLANT SYSTEM CRCS) 3.4.6 RCS Specific Activity RCS Specfffc Activity 3.4.6 LCO 3.4.6 The specific activity of the reactor coolant shall be limited to DOSE EQUIVALENT I-131 specific activity~

0.2 µCi /gm. APPLICABILITY:

MODE 1. MODES 2 and 3 with any main steam line not isolated.

ACTIONS CONDITION REQUIRED ACTION COMPLEfIQij TIME A. Reactor coo 1 ant -----*-------NOTE-----------

specific activity LCO 3.0.4.c is applicable.

> 0.2 µCi/gm and ----------------------------2.0 µCi/gm DOSE Oetennine DOSE Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months EQUIVALENT I-131. A.1 EQUIVALENT I-131. MID A.2 Restore DOSE 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months EQUIVALENT I-131 to within limits. B. Required Action and B.l Oetennine DOSE Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion EQUIVALENT I-131. Ti~ of Condition A not met. AtID .QB B.2.1 Isolate all main 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months steam lines. Reactor Coo 1 ant specific activity>

2.0 DB µCi/gm DOSE EOUIVALEUT I-131. (continued) 2.0 uci/gm chosen as EAL threashold since levels above that activity \_) directly influence continued plant operation.

OAEC 3.4-13 Af!O 255" I

\_) RCS Operational LEAKAGE 3.4.4 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Operational LEAKAGE LCO 3.4.4 RCS operational LEAKAGE shall be limited to: a. s 5 gpm unidentified LEAKAGE: b. s 25 gpm total LEAKAGE averaged over the previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period: and c. s 2 gpm increase in unidentified LEAKAGE within the previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period in MODE 1. APPLICABILITY

MODES 1. 2. and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Unidentified LEAKAGE A.1 Reduce LEAKAGE to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months not within limit. within limits. OR Total LEAKAGE not within limit. . . B. Unidentified LEAKAGE 8.1 Reduce unidentified 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months increase not within LEAKAGE increase to 1 imit. within limits. OR (continued)

Developer Notes: For EAL #I leak rate value, entered the higher of 10 gpm or the value specified in DAEC's Technical Specifications for this type of leakage. 5 gpm per DAEC Tech Specs, so 10 gpm used in EAL 1 For E A L #2 enter the higher of 25 gpm or the value specified in DAEC's Technical Specifications for this type of leakage. )-__../ DAEC uses a total leakage (identified+

unidentified) spec of 25 gpm averaged over 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, so 25 gpm used in the EAL DAEC 3.4-8 Amendment 223 (9) During the approach to criticality , the operator withdrawing control rods should pause long enough between control rod notches to allow neutron count rate and period to stabilize , thus allowing a slow and controlled approach t o the critical condition. (10) When a control rod reaches position 48, perfo r m a coupling check by attempting to withdraw the rod past position 48. If uncoupling should occur , stop control rod withdrawal, notify the CRS , and perform ARP 1 C05A , D-7 (ROD OVERTRAVEL OUT). (11) If criticality occurs significantly earlier or later than expected , notify the CRS. ( 12) Approach the power range on a stable period of about 60-150 seconds. Do not achieve a sustained period of less than 50 seconds. If the period becomes too short, insert the notch and monitor for subcriticality.

(13) Each operable IRM channel must be indicating at least 5/40 scale on Range 1 prior to SRM count rate exceeding 10 6 cps with SRMs fully inserted.

One IRM recorder on each RPS System should be in second speed (30 s/div) during startups while in the IRM Range. However , during extended stable operation in the IRM Range , it is permissible to shift the recorders to normal speed (30 min/div). (14) Reactor plant heatup with M0-4629 and M0-4630, A/B RECIRC PUMP DISCH BYP in the closed position may cause bonnet over pressurization , resulting in failure of the valve to open due to pressure lock and damage to valve internals. (15) Do not establish a vacuum in the main condenser until: (a) Steam seals are in operation. (b) Turbine is on turning gear. (c) Lube Oil Temperature>

80°F. (16) Do not exceed a reactor pressure of 400 psig unless a reactor feed pump is in operation or the MSIVs are closed and the RCIC or HPCI Systems are operating.

(17) Do not retract IRMs until the MODE SWITCH i s in RUN. (18) Do not operate the mechanical vacuum pump above 10% reactor power to minimize the possibility of a hydrogen explosion or an untreated radioactivity release. (19) Place the MODE SWITCH in RUN prior to reaching 12% reactor powe r. I IPOl 2 Page 5 of 46 Rev. 148 BASES-DAEC EOP BASES DOCUMENT BREAKPOINTS Rev. 14 EOP BREAKPOINTS Page 8 of 14 BREAKPOINTS FOR REACTOR LEVEL CONTROL Page 2 of 2 RPV Level Item of Interest Significance (inches) -25 Minimum Steam Cooling

No guarantee that fuel cladding RPV Water Level temperature can be kept <1500 °F (MSCRWL)

ED required in EOP 1 before -25 inches

SAG Entry in EOP 1 if cannot restore and ma i ntain level above -25 inches and spray cooling cannot be establ i shed

Lower end of level control band in A TWS level/power control

Loss of ACC in A TWS Steam Cooling & SAG Entry -39 Elevation of top of Jet

RPV water level following OBA LOCA Pump Suction

SAG Entry in EOP 1 if cannot restore and (-2/3 Core Height) ma i ntain level above -39 inches while spray cooling ATTACHMENT 5 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELATING TO LICENSE AMENDMENT R E QUEST TSCR-166 UPDATED DAEC EAL SCHEME WALLBOARDS

[FOR INFORMATION ONLY] 2 pages fo l low

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NG-18-0090, Response to Request for Additional Information Regarding License Amendment Request (TSCR-166), Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors. (: Difference between revisions

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NG-18-0090, Response to Request for Additional Information Regarding License Amendment Request (TSCR-166), Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors. (: Difference between revisions

Revision as of 11:49, 21 September 2018

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GENERAL COMMENT

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