NG-17-0235, Attachment 2: Duane Arnold Energy Center, License Amendment Request TSCR-166, Clean Copy of the Proposed DAEC EAL Scheme

Attachment 2: Duane Arnold Energy Center, License Amendment Request TSCR-166, Clean Copy of the Proposed DAEC EAL Scheme

ML17363A074
Person / Time
Site:	Duane Arnold
Issue date:	12/15/2017
From:	NextEra Energy Duane Arnold
To:	Office of Nuclear Reactor Regulation
Shared Package
ML17363A067	List: ML17363A069 ML17363A071 ML17363A072 ML17363A074 ML17363A075 ML17363A076 ML17363A078 ML17363A079 ML17363A082 ML17363A083 ML17363A085 ML17363A086 ML17363A088 ML17363A098 NG-17-0235, Attachment 1: Duane Arnold Energy Center, License Amendment Request TSCR-166, Redline Markup of NEI 99-01, Revision 6. Part 1 of 2 NG-17-0235, Attachment 1: Duane Arnold Energy Center, License Amendment Request TSCR-166, Redline Markup of NEI 99-01, Revision 6. Part 2 of 2 NG-17-0235, Attachment 2: Duane Arnold Energy Center, License Amendment Request TSCR-166, Clean Copy of the Proposed DAEC EAL Scheme NG-17-0235, Duane Arnold Energy Center - Calc No. NEE-323-CALC-003, Documentation of the RU1 Emergency Action Levels NG-17-0235, Duane Arnold Energy Center EAL Classification Matrix Hot Conditions (Modes 1, 2, 3) NG-17-0235, Duane Arnold Energy Center, Calc No. NEE-323-CALC-001, Primary Containment Radiation EAL Threshold Determination NG-17-0235, Duane Arnold Energy Center, Calc No. NEE-323-CALC-002, Dose Rate Evaluation of Reactor Vessel Water Levels During Refueling for EAL Thresholds NG-17-0235, Duane Arnold Energy Center, Calc No. NEE-323-CALC-004, Revised Liquid Radiological EALs Per NEI 99-01 NG-17-0235, Duane Arnold Energy Center, Calc No. NEE-323-CALC-005, Revised Gaseous Radiological EALs Per NEI 99-01 Rev. 06 NG-17-0235, Duane Arnold Energy Center, Calc No. NEE-323-CALC-005, Revised Gaseous Radiological EALs Per NEI 99-01 Rev. 6 NG-17-0235, Duane Arnold Energy Center, Calc No. NEE-323-CALC-006, Correlation of Drywell Radiation Monitors for 300 Uci/Gm Dose Equivalent Iodine NG-17-0235, Duane Arnold Energy Center, 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. NG-17-0235, EAL Classification Matrix Cold Conditions (Modes 4,5, and Defueled)
References
NG-17-0235
Download: ML17363A074 (362)

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{{#Wiki_filter:ATTACHMENT 2 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER LICENSE AMENDMENT REQUEST TSCR-166 CLEAN COPY OF THE PROPOSED DAEC EAL SCHEME 130 pages follow

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Duane Arnold Energy Center -(DAEC) Emergency Action Levels Technical Bases Document TBD,2018 1 TABLE OF CONTENTS 1 BASIS FOR EMERGENCY ACTION LEVELS ...*.........*..........*....*.....**.*..*.***.........**...*.*..

1 1.1 OPERA TING REACTORS ........................................................ ......................................... 1 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSl) ..................................... 2 1.3 NRC ORDER EA-12-051 ................................................................................................ 3 2 KEY TERMINOLOGY USED IN DAEC EAL SCHEME ..................................................... 4 2.1 EMERGENCY CLASSIFICATION LEYEL (ECL) ............................................................... 4 2.2 INITIATING CONDITION (IC) ****************************************************************************************** 6-2.3 EMERGENCY ACTION LEVEL (EAL) ***************************************************************************** 6 2.4 FISSION PRODUCT BARRIER THRESHOLD ..................................................................... 6 3 DESIGN OF THE DAEC EMERGENCY CLASSIFICATION SCHEME ............................. 7 3.1 ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS (ECLs) **************

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7 3.2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS ******************** 11 3.3 DAEC-SPECIFIC ORGANIZATION AND PRESENTATION OF GENERIC INFORMATI0Nl2 3.4 IC AND EAL MODE APPLICABILITY ................................................. .......................... 13 4 DAEC SCHEME DEVELOPMENT

• .*.*.*.*.*.*.*........****....*.*.*.....*..****..*.*.*..**.**.*........*.*.***.

14 4.1 GENERAL DEVELOPMENT PROCESS ............................................................................ 14 4.2 CRITICAL CHARACTERISTICS ...................................................................................... 14 4.3 INSTRUMENTATION USED FOR EALs ************************************************************************** 15 4.4 EAL/THRESHOLD REFERENCES TO AOP AND EOP SETPOINTS/CRITERIA

• • • • • • • • ***** 15 5 GUIDANCE ON USING THE DAEC EALS *.....*.**......**.**...****.....****.**.....*.*....****.....*.*.*..

16 5.1 GENERAL CONSIDERATIONS ........................................................................................ 16 5.2 CLASSIFICATION METHODOLOGY

18 5.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS ........................................ 18 5.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION

18 5.5 CLASSIFICATION OF IMMINENT CONDITIONS ............................................................. 19 5.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING

19 5.7 CLASSIFICATION OF SHORT-LIVED EVENTS *************************************************************** 20 5.8 CLASSIFICATION OF TRANSIENT CONDITIONS ............................................................ 20 5.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION

21 5.10 RETRACTION OF AN EMERGENCY DECLARATION ....................................................... 21 11 6 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS ........................ 22 7 COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION ICS/EALS .................... 38 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS .............. 60 9 FISSION PRODUCT BARRIER ICS/EALS ****************************************************************** 62 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS ......... 77 11 SYSTEM MALFUNCTION ICS/EALS *************************************************************************** 99 APPENDIX A -ACRONYMS AND ABBREVIATIONS ........................................................ A-1 APPENDIX B -DEFINITIONS

11*****n**************************************************************************B*1 111 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, Criteriafor 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-l.

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 ofIC HUl and IC HAl 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 off site 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 if requested). 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 CPR 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 (NODE) Alert Site Area Emergency (SAE) General Emergency (GE) 2.1.1 Notification of Unusual Event (NODE) 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 offsite for more than the immediate site area. Purpose: The purpose of the General Emergency declaration is to initiate predetermined 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 I 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 I Cs 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 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 reporting 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 of ECL 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 ofICs 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 (NOUE) 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 P AG 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 P AG 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 P AG 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. 9 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 of risk-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. 10 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 I Cs 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 I Cs 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/trip to shut down the reactor, natural phenomena (e.g., an earthquake), or man-made hazards such as a toxic gas release. 11 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 / 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 I Cs 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. 12 3.4 IC AND EAL MODE APPLICABILITY The DAEC emergency classification scheme was developed recognizing that the applicability of I Cs 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) 13 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 facilitate timely and accurate classification of fission product barrier losses and/or potential losses. The user-aid logic is consistent with the classification logic presented in Section 9.

• The ICs, 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.
• ICs, 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 concmTent events or conditions.

14 !

• 4.3 INSTRUMENTATION USED FOR EALs '4.4 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. 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 CPR 50.54( q) is required. 15 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 refen-ed to as Fission Product Ban-ier 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. For EAL thresholds that specify a duration of the off-normal condition, the NRC expects that the emergency declaration process run concurrently with the specified threshold duration. Once the off-normal condition has existed for the duration specified in the EAL, no further effort on this declaration is necessary-the EAL has been exceeded. Consider as an example, the EAL "fire which is not extinguished within 15 minutes of detection." On receipt of a fire alarm, the plant fire brigade is dispatched to the scene to begin fire suppression efforts.

• If the fire brigade reports that the fire can be extinguished before the specified duration, the emergency declaration is placed on hold while firefighting activities continue.

If the fire brigade is successful in extinguishing the fire within the specified duration from detection, no emergency declaration is wan-anted based on that EAL. 16

• If the fire is still burning after the specified duration has elapsed, the EAL is exceeded, no further assessment is necessary, and the emergency declaration would be made promptly.

As used here, "promptly" means at the first available opportunity (e.g., if the Shift Manager is receiving an update from the fire brigade at the 15-minute mark, it is expected that the declaration will occur as the next action after the call ends).

• If, for example, the fire brigade notifies the shift supervision 5 minutes after detection that the brigade itself cannot extinguish the fire such that the EAL will be met imminently and cannot be avoided, the NRC would not consider it a violation of the licensee's emergency plan to declare the event before the EAL is met (e.g., the 15-minute duration has elapsed).

While a prompt declaration would be beneficial to public health and safety and is encouraged, it is not required by regulation.

• In all of the above, the fire duration is measured from the time the alarm, indication, or report was first received by the plant operators.

Validation or confirmation establishes that the fire started as early as the time of the alarm, indication, or report. 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 ascertain 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 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. 17 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. For example: If an Alert EAL and a Site Area Emergency EAL are met, whether at one unit or at two different units, a Site Area Emergency should be declared. There is no "additive" effect from multiple EALs meeting the same ECL. For example: If two Alert EALs are met, whether at one unit or at two different units, an Alert should be declared. Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarificatzon o/NRC Guidance for Emergency Notifications During Quickly Changing Events. 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. 18 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. The following approach to downgrading or terminating an ECL is recommended. ECL Action When Condition No Longer Exists Unusual Event Terminate the emergency in accordance with plant procedures. Alert Downgrade or terminate the emergency in accordance with plant procedures. Site Area Emergency with no Downgrade or terminate the emergency in long-term plant damage accordance with plant procedures. Site Area Emergency with Terminate the emergency and enter recovery in long-term plant damage accordance with plant procedures. General Emergency 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. 19 5.7 CLASSIFICATION OF SHORT-LIVED EVENTS As discussed in Section 3.2, event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed. If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration. Examples of such events include a failure of the reactor protection system to automatically scram/trip the reactor followed by a successful manual scram/trip or an earthquake. 5.8 CLASSIFICATION OF TRANSIENT CONDITIONS Many of the ICs 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 performed in accordance with procedures. EAL momentarily met but the condition is corrected prior to an emergency 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. 20 5.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION In soine 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. 21 6 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS 22 RU1 ECL: Notification of Unusual Event Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODAM limits for 60 minutes or longer. Operating Mode Applicability: All Emergency Action Levels: Notes:

• The Emergency Director should declare the Unusual Event promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.
• If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 60 minutes.
• If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.

RUl.l RUl.2 RUl.3 Reading on ANY of the following effluent radiation monitors greater than the reading shown for 60 minutes or longer: Effluent Monitor Classification Thresholds Monitor NOUE Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8) l.OE-03 uCi/cc "' Turbine Building ventilation rad monitor (Kaman 1/2) = l.OE-03 uCi/cc 0 Q,I "' Off gas Stack rad monitor (Kaman 9/10) 2.0E-01 uCi/cc c., LLRPSF rad monitor (Kaman 12) l.OE-03 uCi/cc 'C GSW rad monitor (RIS-4767) 2.0E+03 CPS *= RHRSW & ESW rad monitor (RM-1997) 8.0E+02 CPS O" RHRSW & ESW Rupture Disc rad monitor (RM-4268) l.OE+03 CPS Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer. Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the ODAM limits for 60 minutes or longer. 23 Definitions: None Basis: This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or 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 administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that carinot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs. more fully addresses the spectrum of possible accident events and conditions. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. EAL RUl .1 -This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways. EAL RUI .2 -This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL will typically be associated with planned batch releases from non-continuous release pathways ( e.g., radwaste, waste gas). EAL RUI .3 -This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways ( e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.). Escalation of the emergency classification level would be via IC RAl. 24 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of water level above irradiated fuel. Operating Mode Applicability: All Emergency Action Levels: RU2 RU2.1 a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:

• Report to control room (visual observation)
• Fuel pool level indication (Ll-3413) less than 36 feet and lowering
• WR GEMAC Floodup indication (LI-4541) coming on scale AND b. UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors.
• Spent Fuel Pool Area, RI-9178
• North Refuel Floor, RI-9163
• New Fuel Vault Area, RI-9153
• South Refuel Floor, RI-9164
• NW Drywell Area Hi Range Rad Monitor, RIM-9184A
• South Drywell Area Hi Range Rad Monitor, RIM-9184B 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. REFUELING PATHWAY: The reactor refueling cavity, spent fuel pool and fuel transfer canal. 25 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 level 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 bridge 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 level 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 valid 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 Technical 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 3 7 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. Increased 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. 26 RA1 ECL: Alert Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE.

• Operating Mode Applicability:

All Emergency Action Levels: Notes:

• The Emergency Director should declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
• If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
• If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
• The pre-calculated effluent monitor values presented in EAL RAl .1 should only be used for emergency classification assessments until the results from a dose assessment using actual RAl.1 RAl.2 RAl.3 RAl.4 meteorology are available.
• Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer: Effluent Monitor Classification Thresholds Monitor Alert Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7 /8) l .OE-02 uCi/cc "' = Turbine Building ventilation rad monitor (Kaman 1/2) l .OE-02 uCi/cc 0 Q.I "' Offgas Stack rad monitor (Kaman 9/10) 4.5E+Ol uCi/cc C, LLRPSF rad monitor (Kaman 12) l .OE-02 uCi/cc GSW.rad monitor (RIS-4767) 2.0E+04 CPS 't:l *a RHRSW & ESW rad monitor (RM-1997) 1.0E+04 CPS O' RHRSW & ESW Rupture Disc rad monitor (RM-4268) 2.0E+04 CPS Dose assessment using actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond SITE BOUNDARY.

[Preferred] Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for one hour of exposure. Field survey results indicate EITHER of the following 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 indicate thyroid CDE greater than 50 mrem for one hour of inhalation.

27 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 (PA Gs). It includes both monitored and rm-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 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 ad,dresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1 % of the EPA P AG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. Escalation of the emergency classification level would be via IC RS 1. 28 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, RI-9178
• North Refuel Floor, RI-9163
• New Fuel Vault Area, RI-9153
• South Refuel Floor, RI-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 ft. Definitions:

REFUELING PATHWAY -The reactor refueling cavity, spent fuel pool and fuel transfer canal. Basis: This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant. 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. 29 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 ICs. EALRA2.l This EAL escalates from RU2 in that the loss oflevel, 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 frradiated fuel assembles stored in the pool. Escalation of the emergency classification level would be via ICs RSI or RS2. 30 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.l Dose rate greater than 15 mR/hr in ANY of the following areas:

• Control Room ARM (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 F ICs. 31 RS1 ECL: Site Area Emergency Initiating Condition: Release of gaseous radioactivity resulting in off site 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 Site Area Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
• If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
• If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
• The pre-calculated effluent monitor values presented in EAL RS 1.1 should only be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

RSI.I RSl.2 RSl.3 Reading on ANY of the following radiation monitor greater than the reading shown for 15 minutes or longer: Effluent Monitor Classification Thresholds Monitor SAE Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8) l.OE-01 uCi/cc "' = Turbine Building ventilation rad monitor (Kaman 1/2) l.OE-01 uCi/cc 0 "' C,I Offgas Stack rad monitor (Kaman 9/10) 4.5E+02 uCi/cc c.:, LLRPSF rad monitor (Kaman 12) l.OE-01 uCi/cc Dose assessment using actual meteorology indicates doses greater than I 00 mrem TEDE or 500 mrem thyroid CDE 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 greater than I 00 mR/hr expected to continue for 60 minutes or longer.
• Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation.

32 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 (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. *

• 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 RSl.3. Escalation of the emergency classification level would be via IC RG 1. 33 ECL: Site Area Emergency Initiating Condition: Spent fuel 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 loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity. Escalation of the emergency classification level would be via IC RG 1 or RG2. 34 RG1 ECL: General Emergency Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE. Operating Mode Applicability: All Emergency Action Levels: Notes:

• The Emergency Director should declare the General Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
• If an ongoing release is detected and the release start time is unlmown, assume that the release duration has exceeded 15 minutes.
• If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
• The pre-calculated effluent monitor values presented in EAL RG 1.1 should only be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

RGl.l RGl.2 RGl.3 Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer: Effluent Monitor Classification Thresholds Monitor GE "' = Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7 /8) l.OE+OO uCi/cc 0 Turbine Building ventilation rad monitor (Kaman 1/2) l .OE+OO uCi/cc Q,I "' l:'lS t-' Offgas Stack rad monitor (Kaman 9/10) 4.5E+03 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 following at or beyond the SITE BOUNDARY:

• Closed window dose rates greater than 1,000 mR/hr expected to continue for 60 minutes or longer.
• Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrem for one hour of inhalation.

35 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 (PA Gs). 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 RGI.3. 36 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 General Emergency promptly upon determining that 60 minutes 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. 37 7 COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION ICS/EALS 38 CU1 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of RPV inventory for 15 minutes or longer. Operating Mode Applicability: 4, 5 Emergency Action Levels: Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. CUI.I CUl.2 UNPLANNED loss of reactor coolant results in RPV level less than ANY of the following for 15 minutes or longer: a. In Mode 4, RPV water level less than 170" OR b. In Mode 5, ifRPV level band is established above the RPV flange and RPV water level drops below the RPV flange. OR c. In Mode 5, ifRPV level band is established below the RPV flange and RPV water level drops below RPV level band. 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 ari 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. 39 EAL CUI .1 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 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level. EAL 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. 40 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 Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. CU2.1 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 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. Systems 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 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. 41 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. 42 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 Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. CU3.l CU3.2 UNPLANNED increase in RCS temperature to greater than 212°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. 43 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. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation to Alert would be via IC CAI based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria. 44 I I . I 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 Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. CU4.1 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. Systems 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. Minimum DC bus voltage selected due to automatic trip of the inverters at 105 VDC decreasing. Depending upon the event, escalation of the emergency classification level would be via IC CAI or CA3, or an IC in Recognition Category R. 45 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 (incluqing fixed cell phone system)
• Control Room fixed satellite phone system
• FTS Phone system CU5.3 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 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.). 46 EAL CU 5 .1 addresses a total loss of the communications methods used in support of routine plant operations. 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 CU5.3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. 47 ECL: Alert Initiating Condition: Loss ofRPV inventory. Operating Mode Applicability: 4, 5 Emergency Action Levels: Note: The Emergency Director should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. CAI.I CAI.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 CA1 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 CAI.I, 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 CAI.I 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. 48 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 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. The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS I If the RPV inventory level continues to lower, then escalation to Site Area Emergency would be via IC CSI. 49 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: Note: The Emergency Director should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. CA2 CA2.1 Loss of ALL offsite and ALL onsite AC Power to 1A3 and IA4 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. Systems 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 I or RS 1. 50 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 Alert 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 the following table. Table: RCS Heat-up Duration Thresholds RCS Integrity CONTAINMENT CLOSURE Heat-up Duration Status Intact Not applicable 60 minutes* Established 20 minutes* Not intact 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). 51 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 CS 1 or RS 1. 52 CA6 ECL: Alert Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the cunent operating mode. Operating Mode Applicability: 4, 5 Emergency Action Levels: Notes: CA6.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 wananted.
• If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded perfo1mance to at least one train of a SAFETY SYSTEM, then this emergency classification is not wananted.

a. b. The occunence of ANY of the following hazardous events:

• Seismic event (earthquake)
• Internal or external flooding event
• High winds or tornado strike
• FIRE
• EXPLOSION
• River level above 757 feet
• River Water Supply (RWS) pit low level alarm
• Other events with similar hazard characteristics as determined by the Shift Manager or Emergency Director AND 1. 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, OR
• Loss of the safety function of a single train SAFETY SYSTEM. 53 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. Systems 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 performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM 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 system issues without compromising public health and safety from radiological events. Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. Escalation of the emergency classification level would be via IC RS 1. 54 CS1 ECL: Site Area Emergency Initiating Condition: Loss ofRPV inventory affecting core decay heat removal capability. Operating Mode Applicability: 4, 5 Emergency Action Levels: Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded. CSI.l CSI.2 CSI.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 ANY of the following:

• Drywell Monitor (9184A/B) reading greater than 5.0 R/hr
• Erratic source range monitor indication
• 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 I) 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. 55 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 RCS/reactor vessel level cannot be restored, fuel damage is probable. Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified RCS/reactor vessel levels ofEALs CSl.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 ofRPV 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 CSI.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 ofleakage 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. 56 CG1 ECL: General Emergency Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with containment challenged. Operating Mode Applicability: 4, 5 Emergency Action Levels: Note: The Emergency Director should declare the General Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded. CGI.1 CGI.2 a. RPV level less than +15 inches for 30 minutes or longer. AND b. ANY indication from the Secondary Containment Challenge Table (see below). a. RPV level cannot be monitored for 30 minutes or longer. AND b. Core uncovery is indicated by ANY of the following:

• Drywell Monitor (9184A/B) reading greater than 5.0 R/hr
• Erratic source range monitor indication
• 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 Challenge Table (see below). Secondary Containment Challenge Table
• 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) ofEOP 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.

57 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 potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA P AG exposure levels offsite for more than the immediate site area. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable. With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is 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 bum 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 containmenL 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 CG 1.2.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. 58 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 ofleakage 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. 59 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS 60 ECL: Notification of Unusual Event Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY. Operating Mode Applicability: All Emergency Action Levels: E ... HU1 E-HUI.1 Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than the values shown below on the surface of the spent fuel cask. 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 HUI and HAI. 61 9 FISSION PRODUCT BARRIER ICS/EALS No 62 Table 9-F: DAEC EAL Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Barriers FAlALERT FSl SITE AREA EMERGENCY FGlGENERALEMERGENCY ANY Loss or ANY Potential Loss of either the Loss or Potential Loss of ANY two barriers. Loss of ANY two barriers and Loss OR Fuel Clad OR RCS barrier. Potential Loss of the third ban*ier. Operating Mode Applicability: 1, 2, 3 Operating Mode Applicability: 1, 2, 3 Operating Mode Applicability: 1, 2, 3 .. ' ,, '",.,_ ,---'. ,' " RCS Barrier : *' ,'j. ' .,i F*~el Clad ~artier '" ,ContainmentBarrier' H 1 ,, " ' a'<- ~' ,, LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS 1. Primary Containment Conditions

1. Primary Containment Conditions

1. Primary Containment Conditions Not Applicable Not Applicable A. Primary Not Applicable A. UNPLANNED A. Torus pressure containment rapid drop in greater than 53 . pressure greater Drywell pressure psig 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 63 ', Fuel Clad Barrier .. ' ' I* *, RCS Bari:ier 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. RCS Leak Rate 3. RCS Leak Rate 3. RCS Leak Rate 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 Feed water, 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) ofEOP (MSOL) ofEOP 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.

64 ,F*uefCiadiBarvier ' .* ...

• Cohtain:tnJnt Barrier*;*
• r *, ., RCSBa~rier

..* ,, ,,. <. * :" ,, * ' ,< t, * ... .. -* 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 200 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. 65 Basis Information For DAEC EAL Fission Product Barrier Table 9-F 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. Primary Containment Conditions There is no Loss or Potential Loss threshold associated with Primary Containment Condition.

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 corresponds 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. 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 ofRPV 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 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 ofRPV inventory. 66 DAEC FUEL CLAD BARRIER THRESHOLDS ( cont.): 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 SAS or SS5 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. RCS Leak Rate There is no Loss or Potential Loss threshold associated with RCS Leak Rate. 4. Primary Containment Radiation Loss 4.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 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 BaiTier. Loss 4.B The Torus radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the 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 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. The radiation monitor reading in this threshold is 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. 67 i I DAEC FUEL CLAD BARRIER THRESHOLDS (cont.): 5. Other Indications Loss 5.A Results obtained from. procedure PASAP 7.2, Fuel Dam.age Assessment, indicate at least 5% fuel clad dam.age. There is no Potential Loss threshold associated with Other Indications. Developer Notes: Loss and/or Potential Loss 5.A Developers should determine if other reliable indicators exist to evaluate the status of this fission product barrier ( e.g., review accident analyses described in the site Final Safety Analysis Report, as updated). The goal is to identify any unique or site-specific indications that will promote timely and accurate assessment of barrier status. Any added thresholds should represent approximately the sam.e relative threat to the barrier as the other thresholds in this column. Basis information for the other thresholds may be used to gauge the relative barrier threat level. 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 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. 68 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 l.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 con-esponds to the top of active fuel (TAF) and is used in the EOPs to indicate challenge to core cooling. The RPV water level threshold is the same as Fuel Clad ban-ier Potential Loss threshold 2.A. Thus, this threshold indicates a Loss of the RCS ban-ier and Potential Loss of the Fuel Clad ban-ier 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 baITier) 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 of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this RCS ban-ier 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 ofRPV inventory. The term, "cannot be restored and maintained above," means the value ofRPV 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 cun-ent 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. 69 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 (MSCR WL ). 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 SS5 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 Normal 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. 70 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.

Developer Notes: Loss and/or Potential Loss 5.A Developers should determine if other reliable indicators exist to evaluate the status of this fission product barrier ( e.g., review accident analyses described in the site Final Safety Analysis Report, as updated). The goal is to identify any unique or site-specific indications that will promote timely and accurate assessment of barrier status. Any added thresholds should represent approximately the same relative threat to the barrier as the other thresholds in this column. Basis information for the other thresholds may be used to gauge the relative barrier threat level. 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. 71 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 Banier 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. Fallowing 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 RI 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. 72 DAEC CONTAINMENT BARRIER THRESHOLDS (cont.): 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 banier could occur. Potential Loss l .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 HCTL is a function ofRPV 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.

73 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 dam.age is possible. BWR EPGs/SAGs specify the conditions that require primary containment flooding. When primary containment flooding is required, the EPGs are exited and SAGs are entered. Entry into SAGs 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. RCS Leak Rate 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. 74 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 ofradioactivity 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 baiTier potentially lost in the event that barrier status cannot be monitored.

75 [THIS PAGE IS LEFT BLANK INTENTIONALLY] 76 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS 77 ECL: Notification of Unusual Event Initiating Condition: Confinned SECURITY CONDITION or threat. Operating Mode Applicability: All Emergency Action Levels: HU1 HUI.I A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by DAEC Security Shift Supervision. HUI.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 HGI. 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. 78 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 HUl .1 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 10 CFR § 2.390 information. EAL HUl.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 HUl .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 infonnation. 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 HAl. 79 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 1 C35. 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 SA8. 80 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 HU3.5 HU3.6 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 offsite 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. River level above 757 feet. River Water Supply (RWS) pit low level alarm. 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. Systems 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. 81 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. 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 Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011. EAL HU3.5 addresses the observed effects of flooding in accordance with AOP 902 (Flood). Plant site finished grade is at elevation 757.0 feet. Personnel doors and railroad and truck openings at or near grade would require protection in the event of a flood above elevation 757.0 feet. Therefore, EAL 6 uses a threshold of flood water levels above 757.0 feet. . EAL HU3.6 addresses the effects ofloss ofriver water make-up capability. The intake structure for the safety-related water supply systems (river water, RHR service water, and emergency* service water) is located on the west bank of the Cedar River. River levels below the intake structure inlet or a blockage of the intake would result in a loss of the ability to provide make-up water for safety-related systems. The overflow weir is at elevation 724 feet 6 inches. River level at or below this elevation will result in all river flow being diverted to the safety related water supply systems. The top of the intake structure around the pump wells is at elevation 724 feet. If the river water level dropped to this level, the pump suction would have no continuous supply. Blockages of the intake structure may result from debris, ice, or aquatic life. A loss of flow into the intake structure, due to a blockage or low river level, will result in the pit level lowering to the alarm setpoint (723.0 feet) and a resulting alarm in the Control Room. Therefore, this EAL uses a threshold of low pit level as a potential substantial degradation of the ultimate heat sink capability. Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C. 82 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:

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

HU4.l a. A FIRE is NOT extinguished within 15-minutes of ANY of the following FIRE HU4.2 HU4.3 HU4.4 detection indications:

• Report from the field (i.e., visual observation)
• Receipt of multiple (more than 1) fire alarms or indications
• Field verification of a single fire alarm AND b. The FIRE is located within ANY 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 Safe ShutdownNital Areas Category Area Electrical Power 1G31 DG and Day Tank Rooms, 1G21 DG and Day Tank Rooms, Battery Rooms, Essential Switchgear Rooms, Cable Spreading Room Heat Sink/ Torus Room, Intake Structure, Pumphouse Coolant Supply Containment Drywell, Torus Emergency NE, NW, SE Comer Rooms, HPCI Room, RCIC Room, RHR Systems Valve Room, North CRD Area, South CRD Area, CSTs Other Control Building, Remote Shutdown Panel 1C388 Area, Panel 1C55/56 Area, SBGT Room 83 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 iflarge 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 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished ( e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc. Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report. 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. 84 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. 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 CPR 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. l 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. 85 In addition, Appendix R to IO 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 SA8. 86 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. Systems 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. 87 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: HALI 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). 88 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 .1 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 ofreadiness. 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. 89 HAS ECL: Alert Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations. Operating Mode Applicability: All Emergency Action Level: HA5.l An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel (1C388). 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. 90 HA6 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.l 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. 91 ECL: Site Area Emergency Initiating Condition: HOSTILE ACTION within the PROTECTED AREA. Operating Mode Applicability: All Emergency Action Levels: HS1 HSI.I 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]. 92 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. 93 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 Site Area Emergency promptly upon determining that 20 minutes has been exceeded, or will likely be exceeded. HS5.l a. An event has resulted in plant control being transferred 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 afire, transfer 9/ control at panels JC388, JC389, JC390, JC391, and JC392 is required to be completed within 20 minutes." Escalation of the emergency classification level would be via IC FG 1 or CG 1. 94 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. 95 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.l 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: I. 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 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. 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. 96 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. 97 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.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 IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area. 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 per.son(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. 98 11 SYSTEM MALFUNCTION ICS/EALS 99 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 Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. SUI.I Loss of ALL offsite AC power capability to 1A3 AND 1A4 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 SAI. 100 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 Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. SU3.l a. Definitions: An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer.

• Reactor Power
• RPV Water Level
• RPV Pressure
• Primary Containment Pressure
• Suppression Pool Level
• Suppression Pool Temperature 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. Systems 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. 101 An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CPR 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 ofreactivity 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. 102 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.l 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 hours 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 SU4.1, 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 I Ci/sec. For EAL SU4.2, coolant samples exceeding the 2.0 µCi/gm dose equivalent I-131 concentration require prompt action by DAEC Technical Specifications and are representative of minor fuel cladding degradation. Escalation of the emergency classification level would be via I Cs F Al or the Recognition Category R ICs. 103 ECL: Notification of Unusual Event Initiating Condition: RCS leakage for 15 minutes or longer. Operating Mode Applicability: 1, 2, 3 Emergency Action Levels: SUS Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes 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 normal 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. 104 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. 105 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 1 COS 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 1C05 are successful in lowering reactor power below 5% power

• Manual Scram Pushbuttons
• Mode Switch to Shutdown
• Alternate Rod Insertion (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 trip). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. 106 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.

107 ECL: Notification of Unusual Event Initiating Condition: Loss of ALL onsite or offsite communications capabilities. Operating Mode Applicability: l, 2, 3 Emergency Action Levels: SU7.1 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 off site 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 off site 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.). 108 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 response 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. 109 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 Alert promptly upon determining that 15 minutes 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 addit~onal 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. Systems 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 SUI. 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 SS 1. 110 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 Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. SA3.l a. An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer.

• Reactor Power
• RPV Water Level
• RPV Pressure
• Primary Containment Pressure
• Suppression Pool Level
• Suppression Pool Temperature AND b. ANY of the following transient events in progress.
• Automatic or manual runback greater than 25% thermal reactor power
• Electrical load rejection greater than 25% full electrical load
• Reactor scram
• ECCS actuation
• Thermal power oscillations greater than 10% 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. Systems 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. 111 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 rep01ied 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 ofreactivity 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 FSI or IC RSl. 112 SA6 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. 113 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 FS 1. Absent the plant conditions needed to meet either IC SS6 or FS 1, 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). 114 SA8 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: SA8.1

• 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 the SAFETY SYSTEM, then this emergency classification is not warranted.
• For a single train SAFETY SYSTEM, degraded performance which results in loss of the safety function of the SAFETY SYSTEM a. b. The occurrence of ANY of the following hazardous events:
• Seismic event (earthquake)
• Internal or external :flooding event
• High winds or tornado strike
• FIRE
• EXPLOSION
• River level above 757 feet
• River Water Supply (RWS) pit low level alarm
• Other events with similar hazard characteristics as determined by the Shift Manager or Emergency Director AND 1. 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, OR
• Loss of the safety function of a single train SAFETY SYSTEM. 115 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 prefened 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. Systems 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 cunent operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance for criteria SA8.1.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. Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. Escalation of the emergency classification level would be via IC FS 1 or RS 1. 116 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 Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. SS 1.1 Loss of ALL offsite and ALL onsite AC power to 1A3 and 1A4 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. Systems 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, FGI or SGl. 117 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: Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. SS2 SS2.1 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. Systems 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. Minimum DC bus voltage selected due to automatic trip of the inverters at 105 VDC decreasing. 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. 118 SS6 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.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) AND c. EITHER of the following conditions exist:
• RPV level cannot be restored and maintained above -25 inches. OR
• HCL (Graph 4 ofEOP 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. 119 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 Action Level: Note: The Emergency Director should declare the General Emergency promptly upon determining that 4 hours has been exceeded, or will likely be exceeded. SGI.1 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 hours 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. Systems 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 hour 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. 120 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. 121 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 General Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. SG2.1 a. Loss of ALL offsite and ALL onsite AC power to 1A3 and 1A4 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. Systems 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. Minimum DC bus voltage selected due to automatic trip of the inverters at 105 VDC decreasing. 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. 122 I I 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 .................................

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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 ........................................................................................................................ Initiating 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 I 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 TAP ..................................................................................................................... Top of Active Fuel TSC .......................................................................................................... Technical 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 PAG 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 P AG 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 offsite response or monitoring are expected unless 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 riot 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 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 off site 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 LICENSE AMENDMENT REQUEST TSCR-166 DEVIATIONS AND DIFFERENCES MATRIX 100 pages follow

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, DAEC DEVIATIONS AND DIFFERENCES MATRIX TABLE OF CONTENTS

GENERAL COMMENT

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

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

1 ABNORMAL RAD LEVELS/ RADIOACTIVE EFFLUENT ICS/EALS ................................................................... 5 COLD SHUTDOWN/ REFUELING SYSTEM MALFU~CTION ICS/EALS ........................................................ 20 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS ....... .-............................................ 36 FISSION PRODUCT BARRI ER ICS/EALS ....................................................................................................... 38 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS .............................................. 47 ! SYSTEM MALFUNCTION ICS/EALS .................................

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

63 APPENDIX A-ACRONYMS AND ABBREVIATIONS .................................................................................... 84 ---* -* :--*. . .. '":~ . APP EN DIX B -DEFINITIONS ................................................................................................................

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

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

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

9?. DAEC DEVIATIONS AND DIFFERENCES MATRIX

GENERAL COMMENT

S Page 1 GLOBAL#l References to NEI 99-01 GL0BAL#2 Effective date GL0BAL#3 Defined terms in Appendix B; Title Case GL0BAL#4 PWR specific references GLOBAL#S Recognition Category A-Abnormal Radiation Levels/Radiological Effluent category and Emergency Action Levels; AU, AA, AS, and AG GL0BAL#6 Permanently Defueled Section GL0BAL#7 Acknowledgments, Notice and Executive Summary GL0BAL#8 Parameters or indications listed in EALs GL0BAL#9 Site specific information or indication statements GL0BAL#10 Operating Mode Applicability lists mode names (i.e., Power Operation, Startup) GL0BAL#11 Developer's Notes GL0BAL#12 Example EAL statement GL0BAL#13 The following terms: "all, any, or, either" are sometimes capitalized and/or balded in ICs and EALs GL0BAL#14 Defined terms are only listed in APPENDIX B -DEFINITIONS GL0BAL#15 Term "emergency buses" DAEC DEVIATIONS AND DIFFERENCES MATRIX DAEC J ustifi cati'on ,' 'f' Replaced with DAEC Difference Convert generic guidance to DAEC specific. Replaced with TBD, 2018 Difference Convert generic guidance to DAEC specific. Defined terms in Appendix B; Difference All defined terms in Appendix B used in the Upper Case document are in upper case (CAPs} to indicate that the terms are defined. PWR references removed Difference DAEC is a BWR Recognition Category R-Difference DAEC implemented the optional Abnormal Radiation designation of "R" for radiological related Levels/Radiological Effluent items to maintain continuity with previous category and Emergency Action practice at DAEC. Levels; RU, RA, RS, and RG Deleted references to Difference Not Applicable to DAEC Permanently Defueled Station Deleted Difference Not Applicable to DAEC Some parameters or indications Difference Tables or bullets were created to present listed in EALs were placed in DAEC-specific information in a manner tables or bulletized lists. familiar to and desired by scheme users. "Site specific information or Difference Cornpliance with intent of the guidance. indications" were replaced with DAEC-specific information or indications where applicable. Operating Mode Applicability lists Difference Mode numbers used for consistency with mode numbers (i.e., 1, 2, etc.) DAEC procedures and training. Developer's Notes deleted Difference Developer's notes are not reflected in the implementation of the EALs. "Example" deleted from Difference In adopting the EAL, the "example" status statement is no longer applicable. Consistently capitalized and Difference Capitalized and balded conditional terms in balded the following terms: "ALL, I Cs and EALs for consistency based on user ANY, OR, EITHER" in ICs and EALs. feedback. Defined terms are also listed as in Difference Aid to the user to present all needed separate section of each IC/EAL information within the same section of the where the terms are used. Basis document. Replaced with "essential buses" Difference Changed to reflect DAEC nomenclature 2

• Valic{atfon*1, .... # None None None None None None None None None None None None None None None sectjc;m*

NEl,9Q,-01Rev .. 6 COVER PAGE Development of Emergency Action Levels for Non-Passive Reactors Introduction Acknowledgments, Notice and Executive Summary TDC 1. Regulatory Background TDC 1.1 Operating Reactors TDC 1.2 Permanently Defueled Station TDC 1.3 Independent Spent Fuel Storage Installation (ISFSI) TDC 1.4 NRC Order EA-12-051 TDC 1.5 Applicability of Advance and Small Modular Reactor Designs TDC 3.Design of the NEI 99-01 Emergency Classification Scheme TDC 3.3 NSSS Design Differences TDC 3.4 Organization and Presentation of Generic Information TDC 4.0 Site-Specific Scheme Development TDC 4.4; 4.5; 4.6; 4.8 TDC 4.7 Developer and User Feedback TDC Appendix C-Permanently Defueled Station ICs/EALs 1.1 Regulatory Background 1.2 Permanently Defueled Station 1.3 1.3 Independent Spent Fuel Storage Installation (ISFSI) DAEC DEVIATIONS AND DIFFERENCES MATRIX :. *. DAEC* : .. Change*. ** Ju~tification Duane Arnold Emergency Action Difference Changes made to adapt the generic NEI Level Technical Bases Document guidance to a DAEC-specific document Deleted Difference Not Applicable to DAEC 1. Basis for Emergency Action Difference Title change Levels 1.1 Regulatory Background Difference Title change Deleted section Difference Not Applicable to DAEC 1.2 Independent Spent Fuel Difference Re-numbered Storage Installation (ISFSI) 1.3 NRC Order EA-12-051 Difference Re-numbered Deleted section Difference Not Applicable to DAEC 3. Design of the DAEC Emergency Difference Title Change Classification Scheme Deleted section Difference Changes made to adapt the generic NEI guidance to a DAEC-specific document Changed to 3.3 DAEC 3.4 Difference Changes made to adapt the generic NEI Organization and Presentation of guidance to a DAEC-specific document Generic Information 4.0 DAEC Scheme Development Difference Title change Deleted sections Difference Changes made to adapt the generic NEI guidance to a DAEC-specific document Deleted section Difference Changes made to adapt the generic NEI guidance to a DAEC-specific document Regulatory Background Difference Changes made to adapt the generic NEI guidance to a DAEC-specific document and removed developer information Section deleted Difference Not Applicable to DAEC 1.2 Independent Spent Fuel Difference Re-numbered section. Storage Installation (ISFSI) 3 ValJpation. .,#: None None None None None None None None None None None None None None None None None None , letti6n:, (i, , 7,\, "*-, , __ :. ,' I'"~, 1.4 1.5 2 3 3.1 3.2 3.3 3.4 3.5 4 5 6 -11 : ", , , -'NE1 1:99£oiRev.:;5: 1.4 NRC Order EA-12-051 Applicability to Advanced and Small Modular Reactor Designs KEY TERMINOLOGY USED IN NEI 99-01 DESIGN OF THE NEI 99-01 EMERGENCY CLASSIFICATION SCHEME Assignment of Emergency Classification Levels (ECLs) Types of Initiating Conditions and Emergency Action Levels Text referring to NSSS design differences for various types or plants; Developer guidance Organization and Presentation of Generic Information Mode of Applicability Matrix; Typical BWR Operating Modes Site Specific Scheme Development Guidance GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS Recognition Category IC/EAL Matrixes DAEC DEVIATIONS AND DIFFERENCES MATRIX DAEC 1.3 NRC Order EA-12-051 Section deleted KEY TERMINOLOGY USED IN DAEC EAL SCHEME DESIGN OF THE DAEC EMERGENCY CLASSIFICATION SCHEME Assignment of Emergency Classification Levels (ECLs) Types of Initiating Conditions and Emergency Action Levels Deleted DAEC-Specific Organization and Presentation of Generic Information Deleted "Permanently Defueled" section of matrix; replaced Typical BWR Operating Modes with DAEC-specific Operating Modes Development of the DAEC Emergency Classification Scheme GUIDANCE ON USING THE DAEC EALS removed 4 Justi.ficatibri' ,i: Difference Re-numbered and removed wording to add these readings (DAEC installation completed). Difference Not Applicable to DAEC Difference Minor changes to reflect DAEC-specific implementation. Difference Changes made to adapt the generic NEI guidance to a DAEC-specific document Difference Verbatim Difference Difference Difference Difference Difference Difference Changes made to adapt the generic NEI guidance to a DAEC-specific document, removed references to PWRs, and removed developer information. Guidance is now DAEC specific Renumbered to 3.3, made DAEC-specific, and deleted developer information Renumbered to 3.4, removed PWR information, removed permanently defueled, and inserted DAEC Operating Modes to comply with the document intent. Updated to reflect DAEC specific scheme development process. Added text from Section IV.H.7 of ISG-01 explaining how to treat concurrent time periods when making an emergency declaration. Information was added to address a frequently asked question by the DAEC operators. Matrixes were intended for use by EAL developers. Inclusion in licensee scheme is not desired. None None None None None None None None Vl None V2 None DAEC DEVIATIONS AND DIFFERENCES MATRIX ABNORMAL RAD LEVELS/ RADIOACTIVE EFFLUENT ICS/EALS 5 DAEC DEVIATIONS AND DIFFERENCES MATRIX $ec,tic>n

,: h, . .,, ::; ... NEl,9~.~Ql.Rev

.. 6*: .. ;; .* . Ju.stificatie:>n

• r: Vc11idatio11*

1. ,\ .. ... QAEC" ,, Change ,, ,, -: \* 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 .... ::, specific effluent release or longer. <C 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 .Sedion ,: ".4'. , -~ . NEl.99~01 Rev .. 6 ' 's, -,, .-,* -.. '* , Justification . Validation

1. / (1) Reading on ANY effluent (1) Reading on ANY of the Difference See Global Comments #8, 9, 12, & 13. V3 radiation monitor greater following effluent radiation than 2 times the (site-monitors greater than the Reworded EAL statement to remove specific effluent release reading shown for 60 operator confusion as to whether they controlling document) minutes or longer: needed to multiply the values of the limits for 60 minutes or following table by 2 or if the value provided longer: (site-specific

[inserted Table of DAEC-already was 2X. Wording now matches monitor list and threshold specific radiation monitors wording of RS1 and RG1 allowing for easier values corresponding to 2 and threshold values] 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 = Q current radioactivity current radioactivity u -discharge permit for 60 discharge permit for 60 ,....; ,:;) 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 ,NEI 99~01: Rev., ij ' Justification Validation 4{ 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 V4 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 N~l 99101. R~v. 6 . DAEC
• Change . , Justification*

Validation

1. b. UNPLANNED increase in b. UNPLANNED rise in area Difference Global Comments #9 & 13 vs 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 Area, radiation monitors)

Rl-9178

• North Refuel Floor, RI-9163
• New Fuel Vault Area, -Rl-9153
• South Refuel Floor, RI-s::: 0 9164 N
• NW Drywell Area Hi ::) <C Range Rad Monitor, RIM-9184A
• South Drywell Area Hi Range Rad Monitor, RIM-91848 Intent and meaning of the EALs are not altered. 9 DAEC DEVIATIONS AND DIFFERENCES MATRIX Sec.tiQn <
• NEt99-:01 Rev. 6 Df\EC. ., Cl)ange .. ., *= . Justification Validation

1. Recognition Category:

AA1 RA1 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 COE. Operating Mode of Applicability: Operating Mode of Applicability: Verbatim None All All (1) Reading on ANY of the (1) Reading on ANY of the Difference Global Comment #8, 9, 12 & 13 V6, V7 following radiation following radiation monitors monitors greater than the greater than the reading reading shown for 15 shown for 15 minutes or minutes or longer: longer: (site-specific monitor list and threshold values) [inserted Table of DAEC-specific radiation monitors and threshold values] .-1 (2) Dose assessment using (2) Dose assessment using actual Difference Global Comment #9 None actual meteorology meteorology indicates doses Added bracketed 'Preferred' to reinforce indicates doses greater greater than 10 mrem TEDE the 4th Note of the IC than 10 mrem TEDE or 50 or 50 mrem thyroid COE at or mrem thyroid COE 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 COE at or thyroid COE 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 Section NEI 99-01 Rev."*5 DAEC Chari"ge . .Ju~tification Validation

1. * '"' /" ! :, (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 C: 0 mR/hr expected to or longer. .-t continue for 60
• Analyses of field survey <t <t minutes or longer. samples indicate thyroid ~. Analyses of field survey CDE greater than 50 samples indicate mrem for one hour of thyroid CDE greater inhalation.

than 50 mrem for one hour of inhalation. Intent and meaning of the EALs are not altered. 11 DAEC DEVIATIONS AND DIFFERENCES MATRIX Settion. -f., t* \

• NEI 99-01 Rev.* 6 ,, DAEC *change Justification

'. Validation

1. . . .. .. . . *.' ., , .. .. : .. , . *,. C .,.,-{;' M 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 V8 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). 9184B (3) Lowering of spent fuel pool (3) Lowering of spent Difference Global Comment #9 V9 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 J"'!!.,,'

-' NEI 99-01 Rev.' 6 " .* .. "' Justification

Validation

1. ,* 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) rt, (2) An UNPLANNED event Not used at DAEC Difference EALs RA3 and HAS are not applicable.to V10 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 of the 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 , . Justification Recognition Category: AS1 RS1 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 of the Difference Global Comment #8, 9, 12 & 13 V11 following radiation following radiation monitors monitors greater than the greater than the reading reading shown for 15 shown for 15 minutes or minutes or longer: longer: .... (site-specific monitor list and II) <t threshold values) [inserted Table of DAEC-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 S.ection. NEI g~:.oJ.. Rev., .6 .* DAE~ . Change Justification ' ,, ,', ' J *validation#' .. ' ' ., :r***** ,, (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 .... II) <C Intent and meaning of the EALs are not altered. 15 DAEC DEVIATIONS AND DIFFERENCES MATRIX

• section: NEI 99~0i"Rev.

6 DAEC ,ch~nge Justification 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 V12 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 V12 II) <t level to (site-specific Level level to 16.36 feet 3 value). Intent and meaning of the EALs are not altered. 16 DAEC DEVIATIONS AND DIFFERENCES MATRIX <sedi6n'> NEI 99-01 Rev. :6 .. DAEC / :change Ju~tificatipn . *Validation#" 4',,* ,. -, ***.. ,, 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 of the Difference Global Comment #8, 9, 12 & 13 V13 following radiation following radiation monitors monitors greater than the greater than the reading .-I reading shown for 15 shown for 15 minutes or t.!1 minutes or longer: longer: <C (site-specific monitor list and [inserted Table 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 Settfon DAEC . "Justification Validatitjp (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.

.-I mrem for one hour of (!) <C , inhalation. Intent and meaning of the EALs are not altered. 18 DAEC DEVIATIONS AND DIFFERENCES MATRIX C SectJon 'i: * ,,,,-,, , ~:* ; 0 , ,;. , 7

• Z '*' ,; a '" . ,. :,:* *::-' DAEC Ch~_nge -,, 'Justification

,"*, Vaiidation

1. c NEI 99-01 Rev. 6 ' '. ,~ ,,*.,* , "A ;,: ** *1*, '-:z ,, Recognition Category:

AG2 RG2 Difference Global Comment #5 None Initiating Condition: Spent fuel Spent fuel pool level cannot be Difference Global Comment #9 V12 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 V12 (!) <t 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 DAEC Chang~. . . )ustifii:ation .

• Validation.#

, '~ _, ' < ~,

0 ' 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 ANY of the following or RPV [BWR]) level less for 15 minutes or longer: than a required lower limit for 15 minutes or longer. a. In Mode 4, RPV water level less than 170" OR b. In Mode 5, if RPV level band is established above the RPV flange and RPV water level .-1 :::::, drops below the RPV flange. u OR c. In Mode 5, if RPV level band is established below the RPV flange and RPV water level drops below RPV level band. (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. 21 L __ <<' ,., . ',, *Section N ::, u Recognition Category: CU2 Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes or longer. Operating Mode Applicability: Cold Shutdown, Refueling, Defueled (1) a. AC power capability to (site-specific emergency buses) is reduced to a single power source for 15 minutes or longer. AND b. Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS. DAEC DEVIATIONS AND DIFFERENCES MATRIX bAEC CU2 Loss of all but one AC power source to essential buses for 15 minutes or longer. Operating Mode Applicability: 4, 5, Defueled (1) a. AC power capability to 1A3 and 1A4 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. 22 ;change : Verbatim Difference Difference Difference

, ~u~tificatipn

Global Comment #11, 14 Global comment #15 Global Comment #10 Global Comment #9, 12, & 13 Intent and meaning ofthe EALs are not altered. None None None V14

. .. *.* ,* Section M :::::, u NEI 99-01 Rev~ 6 Recognition Category: CU3 Initiating Condition: UNPLANNED increase in RCS temperature. Operating Mode Applicability: Cold Shutdown, Refueling (1} UNPLANNED increase in RCS temperature to greater than (site-specific Technical Specification cold shutdown temperature limit}. (2} Loss of ALL RCS temperature and (reactor vessel/RCS [PWR] or RPV [BWR]} level indication for 15 minutes or longer. DAEC DEVIATIONS AND DIFFERENCES MATRIX ,. DAEC CU3 UNPLANNED increase in RCS temperature. Operating Mode Applicability: 4, 5 (1} UNPLANNED increase in RCS temperature to greater than 212°F (2} Loss of ALL RCS temperature and RPV level indication for 15 minutes or longer 23 C:hange Verbatim Verbatim Difference Difference Difference Global Comment #11, 14 Global Comment #10 Global Comment #9 & 12 Global Comment #4 & 13 Intent and meaning of the EALs are not altered. Validation

1. .,

   • ,* ., .. ; ., ..... * ;< .... None None None Vl None DAEC DEVIATIONS AND DIFFERENCES MATRIX Section bAEC C Change , ~. Ju~titi,,::ation Ce )', ' . V,;;ilidati~l'l

• 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 (1) Indicated voltage is less (1) Indicated voltage is less than Difference Global Comment #9, 12, 13 V15 o:::t ::::) 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 Ln ::::, u DAEC DEVIATIONS AND DIFFERENCES MATRIX ** ,, .. DAEC Recognition Category: CU5 CU5 Initiating Condition: Loss of all Loss of all onsite or offsite onsite or offsite communications communications capabilities. capabilities. Operating Mode Applicability: Cold Shutdown, Refueling, Defueled Operating Mode Applicability: 4, 5, Defueled {1} Loss of ALL of the following {1} Loss of ALL of the following onsite communication methods: (site-specific list of communications methods) onsite communication methods:

• Plant Operations Radio System
• In-Plant Phone System
• Plant Paging System (Gaitronics}

(2} Loss of ALL of the following (2} Loss of ALL of the following ORO communications methods: (site-specific list of communications methods) 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 25 , 'I .. :change
• Justification . Verbatim Verbatim Difference Global Comment #10 Difference Global Comment #9, 12 & 13 Difference Global Comment #9 & 13 None None None V16 V16 V17 DAEC DEVIATIONS AND DIFFERENCES MATRIX Sec:tfon NEI 99-01 Rfiv.-6 DAEC Change Justificaticin

, ' Validation

1. ,*": *" > _,. (3) Loss of ALL of the following (3) Loss of ALL of the following Difference Global Comment #9, 12 & 13 V16 NRC communications NRC communications methods: methods: (site-specific list of

• FTS Phone system -communications methods)
• All telephone lines (PBX .... C: and commercial) 0 u -* Cell Phones (including r.n ::::, fixed cell phone system) u
• Control Room fixed satellite phone system Intent and meaning of the EALs are not altered. 26 0' Section Recognition Category:

CA1 Initiating Condition: Loss of (reactor vessel/RCS [PWR] or RPV [BWR]} inventory. Operating Mode Applicability: Cold Shutdown, Refueling (1) Loss of (reactor vessel/RCS [PWR] or RPV [BWR]} inventory as indicated by level less than (site-specific level}. (2} a. (Reactor vessel/RCS [PWR] or RPV [BWR]} level cannot be monitored for 15 minutes or longer AND b. UNPLANNED increase in (site-specific sump and/or tank} levels due to a loss of (reactor vessel/RCS [PWR] or RPV [BWR]} inventory. DAEC DEVIATIONS AND DIFFERENCES MATRIX DAEC CA1 Loss of RPV inventory. Operating Mode Applicability: 4, 5 (1} Loss of RPV inventory as indicated by level less than 119.5 inches (2) 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 of RPV inventory. 27 '. Justification ,,, I, ,,,,', , '"' Chang~ Verbatim Global Comment #11, 14 Difference Global Comment #4 Difference Global Comment #10 Difference Global Comment #4, 9 & 12 Difference Global Comment #4 Difference Global Comment #4, 9 & 13 Intent and meaning of the EALs are not altered. None None None V18 None None DAEC DEVIATIONS AND DIFFERENCES MATRIX DAEC fhange** ... . . Ju.stifi~~tion 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 <C u Defueled (1) Loss of ALL offsite and ALL (1) Loss of ALL offsite and ALL Difference Global Comment #9, 12 & 13 V14 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

• Section NEI 99:01 Rev.'.6 DAEC ~hange Justification Validation n: 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: 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 V1 RCS temperature to temperature to greater than greater than (site-specific 212°F for greater than the Technical Specification duration specified in the cold shutdown following table: temperature limit) for greater than the duration specified in the following table. Table: RCS Heat-up Duration Threi hnl~e .,CS Heat-up Duration Thresha Difference Global Comment #4 None Containment ,~u.-Up l"l'I RCS Status Closure Status Duration Containment Closure Changed "RCS Status" to "RCS Integrity" to <C u Intact (but not at -----o* Status match current site nomenclature reduced inventory Not applicable h.f\min""° ...... * [PWR]) Intact Not Applicable Established N%intac\

• Established mmu es Not intact (or at reduced inventory

[PWR]) Not Established O minutes Not Established

• If ari,R..CS heat removal system is in operatior
• If an RCS heat removal system is in operatio .vwim t JM 1 Wcs 1 ~mperature is being reduced, frame and RCS temperature is being reduce ' tnejt 15reot applicable.

app 1ca e. (2) UNPLANNED RCS pressure (2) UNPLANNED RCS pressure Difference Global Comment #4 & 9 V19 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 ofthe EAL (This EAL does not apply during water-solid plant conditions. [PWR]) Intent and meaning of the EALs are not altered. 29 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99.:01 Rev.,-6 DAEC ~hange Justification Validation

1. p :!t,* <Ac:£:, s"* Recognition Category:

CA6 CA6 Verbatim Global Comment #11, 14 None 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: 4, Difference Global Comment #10 None Cold Shutdown, Refueling 5 (1) a. The occurrence of ANY of (1) a. The occurrence of ANY of Difference Global Comment #9, 12 & 13 the following hazardous the following hazardous events: events:

• Seismic event
• Seismic event (earthquake) (earthquake)
• Internal or external
• Internal or external flooding event flooding event I.C
• High winds or tornado
• High winds or tornado <( u strike strike
• FIRE
• FIRE
• EXPLOSION
• EXPLOSION
• (site specific hazards)
• River level above 757 V20
• Other events with feet V21 similar hazard
• River Water Supply characteristics as (RWS) pit low level alarm determined by the Shift
• Other events with Manager similar hazard characteristics as determined by the Shift Manager or Emergency Director 30 DAEC DEVIATIONS AND DIFFERENCES MATRIX AND AND b. EITHER of the following:

b. 1. Event damage has Deviation Adopted ~he revised EAL wording provided V22 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 V22 1. The event has caused following:

in approved EAL FAQ 2016-02; with the VISIBLE DAMAGE to a

• Event damage has addition of a 3rd choice due to DAEC having SAFETY SYSTEM caused indications single train SAFETY SYSTEM -component or structure of degraded .... needed for the current performance to a C 0 operating mode. second train of the u -ID SAFETY SYSTEM <C u 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.
• Loss of the safety Intent and meaning of the EALs are not function of a single altered. train SAFETY SYSTEM. 31 DAEC DEVIATIONS AND DIFFERENCES MATRIX Recognition Category:

CS1 CS1 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: 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 V23 not established. not established. AND AND .-I 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 V23 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 DAEC DEVIATIONS AND DIFFERENCES MATRIX (3) a. (Reactor vessel/RCS [PWR] (3) 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. AND AND b. Core uncovery is indicated

b. Core uncovery is indicated Difference Global Comment #9 &13 V24 by ANY of the following:

by ANY of the following:

• (Site-specific radiation
• Drywell Monitor -monitor) reading greater (9184A/B) reading +,i s::: than (site-specific value) greater than 5.0 R/hr 0
• Erratic source range
• Erratic source range .-I II) monitor indication monitor indication u [PWR]
• UNPLANNED level rise in
• UNPLANNED increase in Drywell/Reactor Building (site-specific sump Equipment or Floor Drain and/or tank) levels of sump, or Suppression sufficient magnitude to Pool of sufficient Intent and meaning of the EALs are not indicate core uncovery magnitude to indicate altered. * (Other site-specific core uncovery indications) 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: 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 V23 or RPV [BWR]) level less inches for 30 minutes or .-I (!) 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 V24 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 Verbatim None Containment Challenge Containment Challenge Table (see below}. Table (see below} Containment Challenge Table Difference Global Comment #9 V25 ONTAINMENT CLOSURE not established*

Containment Challenge Table C V26

• CONTAINMENT CLOSURE not established xplosive mixture) exists inside containment NPLANNED increase in containment pressure
• Drywell Hydrogen or Torus Hydrogen gre AND Drywell Oxygen or Torus Oxygen gre ~condary containment radiation monitor reading ite specific value) [BWR]
• UNPLANNED increase in containment pre -* Secondary containment radiation monito -safe operating limits (MSOL) of EDP 3, Ta
• If CONTAINMENT CLOSURE is *If CONTAINMENT CLOSURE is Verbatim re-established prior to exceeding re-established prior to exceeding the 30-minute time limit, then the 30-minute time limit, then Intent and meaning of the EALs are not declaration of a General declaration of a General altered. Emergency is not required.

Emergency is not required. 35 DAEC DEVIATIONS AND DIFFERENCES MATRIX INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS 36 . . *, Section :* I .-1 ::::) :::c I LU Recognition Category: E-HU1 Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY. Operating Mode Applicability: All (1} Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than (2 times the site-specific cask specific technical specification allowable radiation level} on the surface of the spent fuel cask. DAEC DEVIATIONS AND DIFFERENCES MATRIX DAEC E-HUl Damage to a loaded cask CONFINEMENT BOUNDARY. Operating Mode Applicability: All (1) Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than the values shown below on the surface of the spent fuel cask. 61BTDSC 3 feet from HSM Surface Outside HSM Centerline of DSC End Shield Wall Exterior 800 mrem/hr 200 mrem/hr 40 mrem/hr 37 :'I Change Verbatim Verbatim Verbatim Difference ..

• Justification
• , Global Comment #8, 9, 12 & 14 Intent and meaning of the EALs are not altered. None None None V27 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 NEl99-Di Rev. 6 , "'" ' .. DAEC C~an~,e ' f* : ... : ' ;,,., .. : )fr' " _Justi!!fatipn _; 1~, Va!idajipn;#

... r, . . ",~ , . ' ... . " *.* :" .. 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 Deleted Difference applicability carried over onto Table 9-F EAL listing. None 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 9-F: EAL Fission Product Barrier Renumbered and re-labeled due to deletion of Tables 9-F-1 & 3. Table Thresholds for LOSS or POTENTIAL LOSS of Table Thresholds for LOSS or Difference Added None Barriers POTENTIAL LOSS of Barriers 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 Information 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 Sub-Category " 1. RCS Activity Renam~d-to

• 1~ Primclr.v

..* Containment Cohditions 2; RPVWater . -. . . . Level .. .. .. . : , 3. RCS. Leak Rate . -,, " "' DAEC DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of Fuel Clad Barrier Loss Potential Loss Loss Potential Loss A. (Site-specific indications that reactor coolant activity is greater than 300 µCi/gm dose equivalent 1-131}. Not Applicable Not Applicable Not Applicable A. Primary containment A. RPV water level flooding required. cannot be restored and maintained above (site-specific RPV water level corresponding to the top of active fuel} or cannot be A. UNISOLABLE break in ANY of the following: (site-specific systems determined . A. SAG entry is required. Not Applicable A. RPV water level cannot be restored and maintained above +15 inches OR cannot be determined. Not Applicable

• • ** with potential for energy line breaks} A. UNISOLABLE primary system leakage that results in exceeding EITHER of the following:

OR B. Emergency RPV Depressurization.

1. Max Normal Operating Temperature OR 2. Max Normal Operating Area Radiation Level. 40 Difference Renamed Category from "RCS" Activity" to "Primary Containment Conditions" to better align category with thresholds being assessed.

Fuel Clad LOSS 1.A carried over to OTHER INDICATIONS category as LOSS 5.A Difference EPFAQ 2015-004 V28 General Comment #9, 13 Difference Renamed Category from "Not applicable" to "RCS Leak Rate" to better align category with thresholds being assessed.

• 4}::~rfm~rv:'t

.. : Containment.; R~aiat13J, ~:J1c*,. * .. "'":~'.}~:.;,, '!,"'"'-\-,.-., ;,W;;t ~"'* , , 7..,.*, containment radiation monitor reading greater than (site-specific value). A. (site-specific as applicable) A. ANY condition in the opinion of the Emergency Director that indicates Loss of the Fuel Clad Barrier. DAEC DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of Fuel Clad Barrier Potential Loss Not Applicable A. (site-specific as applicable) A. ANY condition in the opinion of the Emergency Director that indicates Potential Loss of the Fuel Clad Barrier. Loss A. Drywell Monitor (9184A/B) reading greater than 200 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. A. ANY condition in the opinion of the Emergency Director that indicates Loss of the Fuel Clad Barrier. 41 Potential Loss Not Applicable Not Applicable A. ANY condition in the opinion of the Emergency Director that indicates Potential Loss of the Fuel Clad Barrier. Difference V29 -V30 Global Comment #9 RCS @300 uci/cc readable on Drywell monitor, so that lower value was used versus value for loss of both RCS and Fuel clad barriers. Torus monitor value follows the developer guidance for "instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory, with RCS radioactivity concentration equal to 300 uCi/gm dose equivalent l-131, into the primary containment atmosphere Difference Global Comment #9 Difference Core damage assessment procedure. Emergency Director changed to Emergency Director to align with site terminology. j::t!~}ii,~ R,~nqmei:Mo .. 6, ** *'('i>ri!J1~5Y:-, :L. *containment. tt>nditit>nl':-1;. DAEC DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of RCS Barrier 0':l\ii:fgg*zo1 RevZG . . DAE<::.,*. ' f, . .::-1,;', *- .,,;-;" ~*, Loss A. Primary containment pressure greater than (site-specific value) due to RCS leakage. A. RPV water level cannot be restored and maintained above 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. Potential Loss Not Applicable Not Applicable A. UNISOLABLE primary system leakage that results in exceeding EITHER of the following:

1. Max Normal Operating Temperature OR 2. Max Normal Operating Area Radiation Level. Loss A. Primary containment pressure greater than 2 psig due to RCS leakage. A. RPV water level 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 De pressurization required.

42 Potential Loss Not Applicable Not Applicable

• 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 Difference Difference Difference V31 Global Comment #9 V23 Global Comment #9, 13 V32 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. Pril'l)ary: ,; A. Primary Not Applicable A. Drywell Monitor Not Applicable Difference Global Comment #9 Coniainmeht containment radiation (9184A/B) reading V29 ' " Radi~tion monitor reading greater than 5 R/hr : greater than (site-after reactor '" specific value). shutdown 5. 'other 'ff A. (site-specific as A. (site-specific as Not Applicable Not Applicable Difference Global Comment #9 lhdicaticms applicable) applicable)

6. e*me'rgency A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in the Difference Emergency Director > ~., Director"'

the opinion of the the opinion of the the opinion of the opinion of the changed to Emergency Judgment Emergency Director Emergency Emergency Emergency Director Director to align with site ,< that indicates Loss Director that Director that that indicates Potential terminology. ";e '-,, of the RCS Barrier. indicates Potential indicates Loss of Loss of the RCS Barrier. ' " ,, Loss of the RCS the RCS Barrier. ,* ,,, ,* Barrier. 43 Sub-Category l 1.;Primary !,<:~ntainment; t* Pressure , f Re.named to . r* ....... . . . :.;* .. *" ,., *. 1_1. Primary i Contaihinerit [ coriditi.ons . f f, ' : I r f\ , ' I* f: V ,, r, fu f-f, ' I t~ 2.RPVWater I O ~' Lev.el [ ,", DAEC DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of Containment Barrier 'NEl99-01 Rev. ~6 , DAEC .. C11~n.~~ Juitification Loss Potential Loss Loss Potential Loss A. UNPLANNED rapid drop A. Primary containment A. UNPLANNED rapid A. Torus pressure Difference Global Comment #9 in primary containment pressure greater than drop in Drywell greater than 53 psig V25 pressure following (site-specific value) pressure following OR V33 primary containment OR Drywell pressure rise B. Drywell or Torus H2 V34 pressure rise B. (site-specific OR cannot be OR explosive mixture) exists B. Drywell pressure determined to be Loss 3.A and 3.B moved to B. Primary containment inside primary response not less than 6% and sub-category 1 "Primary pressure response not containment consistent with LOCA Drywell OR Torus Containment Conditions" consistent with LOCA OR conditions. 02 cannot be as Losses 1.C and 1.D to conditions. C. HCTL exceeded. OR determined to be consolidate concepts into C. UNISOLABLE direct less than 5% single sub-category downstream pathway OR to the environment exists after primary C. HCL (Graph 4 of containment isolation EOP 2) signal exceeded. OR D. Intentional primary containment venting per EOPs Not Applicable A. Primary Not Applicable A. SAG entry is Difference EPFAQ 2015-004 containment required. flooding required. 44

3. ~cs Leak*. Rat*! 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 DAEC DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of Containment Barrier Potential Loss Not Applicable A. Primary containment radiation monitor reading greater than (site-specific value). Loss A. UNISOLABLE primary system leakage that results in exceeding the Max Safe Operating Limit (MSOL) of EOP 3, Table 6 for EITHER of the following:

• Temperature OR
• Radiation Level Not Applicable 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 Cliang~ Difference

<justification . Global Comment #9 V35 Loss 3.A and 3.B moved to category 1 "Primary Containment Conditions" Difference Global Comment #9 V29 DAEC DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of Containment Barrier l l\f ' ';_*, "i( ','

• NE[99-01 R~v. 6 ,. ,' : .. [, i<f; DAE'<: ; .. Cnange . ' ti'll;,le 9-.F-2 * ; ... , J~~tif.i~ati.~,n

.~ I I --*., " t ~. .. : . "'~ . . ,,*' .. * .. l Sub-Category Loss Potential Loss Loss Potential Loss s. 0th.er '} A. (site-specific as A. (site-specific as Not Applicable A. Fuel damage Difference Global Comment #9 ,c applicable)

~ndi~atipns.,* applicable) assessment Core damage assessment IL indicates at least procedure. I 20% fuel clad L., ~,. [ damage . ,* *." .

• r I ' ,, ; r . . .. ... A. ANY condition in the B. ANY condition in the C. ANY condition in the D. ANY condition in the Difference Emergency Director 1 6. Emergency.

I Di~ect'c:,r opinion of the opinion of the opinion of the opinion of the changed to Emergency l Judgment Emergency Director Emergency Director Emergency Director Emergency Director Director to align with site I ! i that indicates Loss of that indicates that indicates Loss that indicates terminology.

r. "" < *., r,: ' the Containment Potential Loss of the of the Containment Potential Loss of the I ' Barrier. Containment Barrier. Containment r* '., .. Barrier. Barrier. ' 46 DAEC DEVIATIONS AND DIFFERENCES MATRIX HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS 47 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section NEl99--01 Rev.<6 C>AEC * , Change* *.

• Justification Recognition Category:

HUl HU1 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). (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 Section NEI 9~_.:.p1 Rev. ,6 , Justj.fjcation' 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 V36 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 Section NEI 99..::01 Rev.<6. , * ,

• i :Justification Vali.cfatic,n

1. 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 M 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 (5) River level above 757 feet. Difference Global Commen.t #9 V37 or technological hazard events) (6) River Water Supply (RWS) pit Difference Global Comment #9 V38 low level alarm. Added as a 6th EAL for this IC versus a list in EAL #5 to maintain consistent format Intent and meaning of the EALs are not altered. so DAEC DEVIATIONS AND DIFFERENCES MATRIX Section . ,,. NEI 99':'01 Rev.' 6. :oAEC Chan~e Justification

i. Valiaatfon

1. 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 Fl RE 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

,::to single fire alarm

• Field verification of a ::, :::c: AND single fire alarm AND b. The FIRE is located within b. The FIRE is located within Difference Global Comment #8, 9, & 13 None ANY of the following plant ANY Table H-1 plant Same room/area listing as current EAL HU2 rooms or areas: rooms or areas. (site-specific list of plant rooms Table H-1 Safe Shutdown/Vital Areas or areas) Category Area Electrical 1G31 DG and Day Tank Rooms, 1G21 Power DG and Day Tank Rooms, Battery Rooms, Essential Switchgear Rooms, Cable Spreading Room Heat Sink/ Torus Room, Intake Structure, Coolant Supply Pumphouse Containment Drywell, Torus Emergency NE, NW, SE Corner Rooms, HPCI Room, Systems RClC Room, RHR Valve Room, North CRD Area, South CRD Area, CSTs Dther Control Building, Remote Shutdown Panel 1C388 Area, Panel lCSS/56 Area, SBGTRoom 51 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99.-01 Rev. 6 * . PAEC Ghange , * , Justification

,: Validation

1. * , (2) a. Receipt of a single fire (2) a. Receipt of a single fire Deviation Global Comment #8, 9 & 13 V39 alarm (i.e., 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 Fl RE 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 +a C: ISFSI outside the plant extinguished within 60 0 Protected Area] minutes of the initial o::I' ::::, 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 S¢ctioh ,* * . .* *Justificatipn ~Valiclatl~n

1. ' 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 in the judgment of which in the judgment of the Emergency Director the Emergency Director indicate that events are in indicate that events are in ,... progress or have occurred progress or have occurred ::::., which indicate a potential which indicate a potential

I: 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 SAFETY systems occurs. SYSTEMS occurs. 53 DAEC DEVIATIONS AND DIFFERENCES MATRIX . DAEC:~. Change . Justification

. ,, *validation

1. R~cognition 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 within the OWNER within the OWNER .-1 <( 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 S~c:tioo NEI 9~Hll:8ev.,,6 Recognition Category: HAS Not used at DAEC Difference EALs RA3 and HAS are not applicable to V39 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 i.n 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

• Sectiot:t:

.J usdficatipn Recognition Category

HA6 HAS Difference Renumbered to align with other similar !Cs 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 I.O (1} An event has resulted in (1} An event has resulted in plant Difference Global Comment #9 & 12 V40 <C :::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 ; Justifi<:atlon Recognition Category: HA7 HA6 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 degradation of the level of the level of safety of the :::c 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 Section * .. Nf:199;.01 Rev.,6 * * .* :1ustification ,* -. *validatidn

1. '; Recognition Category:

HS1 HS1 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 V) 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 Se,ction , NEI 99.,..01 Rev.:.6 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 V40 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) :c specific remote shutdown Remote Shutdown Panel panels and local control (1C388). stations). AND AND Difference Global Comment #4, 9 V40 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 of the EALs are not altered. 59 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section PAEC * ~{,ange* Justification . *. ::validatfon

1. f 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 II) :I: 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 : Section*

• DAE<; Change
• 11,lstification Recognition Category:

HG1 HG1 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

..... 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 .*. l\lEI 99~01J~ev: Y6 '.:Justifh:ation Recognition Category: HG7 HG6 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 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 :c melting with potential for loss of containment integrity loss of containment or HOSTILE ACTION that integrity or HOSTILE results 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

• Secti.oh. . Justificatipn

validation

1. z: 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. .-1 Operating Mode Applicability: Operating Mode Applicability: 1, Difference Global Comment #10 None ::::, Power Operation, Startup, Hot 2, 3 VI 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 for specific emergency buses) 15 minutes or longer. for 15 minutes or longer. Intent and meaning of the EALs are not altered. 64 DAEC DEVIATIONS AND DIFFERENCES MATRIX S~ction:*:.

. N.EI 99:01: Rev: '6

• DAl:C** * * : ,
• Cbange .. , * *. *Justification Valitlati6ri

1. • 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 Applicabi_lity: 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 following parameters from within the Control from within the Control Room for 15 minutes or Room for 15 minutes or longer. longer.

• Reactor Power Difference Global Comment #4, 9 None N [BWR parameter

[PWR

• RPV Water Level :> Vl list] parameter list]
• RPV Pressure Reactor Power Reactor Power RPV Water Level RCS Level
• Primary Containment RPV Pressure RCS Pressure Pressure Primary In-Core/Core
• Suppression Pool Containment Exit Level Pressure Temperature Suppression Pool Suppression Pool Levels in at least
• Level (site-specific Temperature number) two 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 DAE~ Change
• Justification
• :vali.datiOn*

i( 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 {l) {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 V41 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 hours or longer. Intent and meaning of the EALs are not altered. 66 DAEC DEVIATIONS AND DIFFERENCES MATRIX Sec;tion NEI 99:..01 Rev:i:5 . *, * , . *~ . *, *

• l, . *DAEC Change * 'Jus~ificatjon . Recognition Category:

SU4 SUS Verbatim Global Comment #11, 14 None Renumbered to align with other similar !Cs Initiating Condition: RCS leakage RCS leakage for 15 minutes or Verbatim None for 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) RCS unidentified or (1) RCS unidentified or pressure Difference Global Comment #9 & 12 V42 pressure boundary leakage boundary leakage greater greater than (site-specific than 10 gpm for 15 minutes value) for 15 minutes or or longer. longer. <::2' (2) RCS identified leakage (2) RCS identified leakage greater Difference Global Comment #9 V42 ::, greater than (site-specific than 25 gpm for 15 minutes II) 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 " , .DAEC Change

• Justification.
• Valiclatibl'l

1. ~ 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 V43 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. in shutdown the reactor. :::, II) Difference Global Comment #9 AND AND None b. A subsequent manual b. ANY of the following manual action taken at the reactor actions taken at 1COS 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 S~ction -NEI 99~01.Rev

.. i6 -Justification . ~van'aati6n ,if (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 u Alternate Rod Insertion

-* 11'1 ::::> (ARI) II) OR OR Difference Global Comment #4 None 2. A subsequent automatic

2. A subsequent automatic (trip [PWR] / scram scram is successful in [BWRJ) 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

• SectiQri NEI 99~01 Rev; 6 ,.
• PAEC Change * : .. , .. * :,;Justific~tion

S. Val~dati(>n t( 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: 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 V44 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)

I.I) (2) Loss of ALL of the following (2) Loss of ALL of the following Difference Global Comment #9 & 13 V44 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 section *Justification

.\ . (3) Loss of ALL of the following (4) Loss of ALL of the following Difference Global Comment #9 & 13 V44 NRC communications NRC communications methods: methods: -(site-specific list of

• FTS Phone system ...; C: communications methods) All telephone lines ~PBX 0
• u -and commercial)

I.C ::::,

• Cell Phones (including V) fixed cell phone system)
• Control Room fixed satellite phone system Intent and meaning of the EALs are not altered. 71 DAEC DEVIATIONS AND DIFFERENCES MATRIX Sectio.n Nl:i 99-01:ijev.::Ji

~iJustification 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

J are not closed within 15 V) 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 ' NEI 99:.oi:Rev./6

.\< __ ,, '; -Justification . , Vali'datfoti*

1. 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 is reduced ct 1/) buses) is reduced to a to a single power single power source for 15 source for 15 minutes minutes or longer. 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 Section DAEC Change Justification tvalidati&n

1. 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: 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 of the following parameters the following parameters from within the Control from within the Control Room for 15 minutes or Room for 15 minutes or longer. longer. N <C [BWR [PWR parameter Reactor Power Difference Global Comment #4, 8 V>

• None parameter list] list]
• RPV Water Level Reactor Power Reactor Power RPV Pressure RCS Level
• RPVWater Level
• Primary Containment RPV Pressure RCS Pressure Pressure Primary In-Core/Core Exit
• Suppression Pool Containment Temperature Level Pressure Suppression Levels in at least
• Suppression Pool Pool Level {site-specific Temperature number) steam generators Suppression Steam Generator Pool Auxiliary or Temperature Emergency Feed Water Flow AND AND 74 DAEC DEVIATIONS AND DIFFERENCES MATRIX sectibri .!Justification.

b. ANY of the following

b. ANY of the following Difference Global Comment #4, 9 None transient events in transient events in progress.

progress.

• Automatic or manual
• Automatic or manual run back greater than run back greater than 25% thermal reactor 25% thermal reactor -power power .... C:
• Electrical load rejection
• Electrical load rejection 0 u -greater than 25% full greater than 25% full N <( electrical load electrical load II)
• Reactor scram [BWR] /
• Reactor scram trip [PWR]
• ECCS actuation
• ECCS (SI) actuation
• Thermal power
• Thermal power oscillations greater than oscillations greater than 10% 10% [BWR] Intent and meaning of the EALs are not altered. 75 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section
• NEI 99:.01.Rev,<6

r!1ustification,*}

• • Vaiidatfori

1. '" Recognition Category:

SAS SA6 Difference 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 & 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 V43 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. ct II) 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. 76 DAEC DEVIATIONS AND DIFFERENCES MATRIX s~ction * 'Validation

1. ' 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 following hazardous events: events:

• Seismic event
• Seismic event Difference Global Comment #8 & 9 V45 (earthquake) (earthquake)

V46

• Internal or external
• Internal or external flooding event flooding event a,
• High winds or tornado
• High winds or tornado er V) strike strike
• FIRE
• FIRE
• EXPLOSION
• EXPLOSION
• (site-specific hazards)
• River level above 757
• Other events with feet similar hazard
• River Water Supply characteristics as (RWS) pit low level determined by the Shift alarm Manager
• Other events with similar hazard characteristics as determined by the Shift Manager or Emergency Director 77 DAEC DEVIATIONS AND DIFFERENCES MATRIX S,ectiqn . *' NEI 99,~01 Rev. ,Ei. DAE(;. C~ange. < *.**Justification . AND AND b. EITHER of the following:

b.1. Event damage has Deviation Adopted the revised EAL structure and V47 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 V47 2. The event has caused

• Event damage has in approved EAL FAQ 2016-02; with the VISIBLE DAMAGE to a caused indications addition of a 3rd choice due to DAEC having SAFETY SYSTEM of degraded single train SAFETY SYSTEM component or structure performance to a -needed for the current second train of the ...; operating mode. SAFETY SYSTEM C: 0 needed for the en current operating

<C V) mode, or

• The event has resulted in VISIBLE DAMAGE to the second train of a SAFETY SYSTEM needed for the current operating mode.
• Loss of the safety function of a single train SAFETY SYSTEM. Intent and meaning of the EALs are not altered. 78 DAEC DEVIATIONS AND DIFFERENCES MATRIX Recognition Category:

SSl SSl 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 .... Power Operation, Startup, Hot 2, 3 VI VI 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 for 15 minutes or longer. for 15 minutes or longer. '-Intent and meaning of the EALs are not altered. 79 DAEC DEVIATIONS AND DIFFERENCES MATRIX Recognition Category: SSS SS6 Difference Global Comment #11, 14 None Renumbered to align with other similar !Cs 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 V43 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 V'l V'l shutdown the reactor have shutdown the reactor been unsuccessful. have been unsuccessful. AND AND Difference Global Comment #9 V48 C. EITHER of the following

c. EITHER of the following V49 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 Sectioi::i

• . NEI 99-01 Rey. :s. DAEc**. : Change Justification*
• Validation

1. Recognition Category:

SS8 SS2 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 co Standby, Hot Shutdown V) V) (1) Indicated voltage is less (1) Indicated voltage is less Difference Global Comment #9 & 12 V50 than (site-specific bus than 105 VDC on BOTH V51 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. I Intent and meaning of the EALs are not altered. 81 DAEC DEVIATIONS AND DIFFERENCES MATRIX NEI 99-,01 Rev. 6 DAEC Cha~ge ,*, e Justification.:.':.

• Validation

1. .. ,. ,: * * -~, ~'ff.I' , * * > r*** 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 . .-I AND Difference Global Comment #9 & 13 C!J AND V) 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 hours is not hours) is not likely. likely. OR * (Site-specific indication
• Reactor vessel water Difference Global Comment #9 V48 of an inability to level cannot be restored adequately remove heat and maintained above from the core) -25 inches. Intent and meaning of the EALs are not altered. 82 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 DAEC
• Change Justification
• -Validation

1. Recognition Category:

SG8 SG2 Difference Global Comment #11, 14 None Renumbered to align with other similar ICs 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 (!J specific emergency buses) and 1A4 for 15 minutes or II) for 15 minutes or longer. longer. AND AND Difference Global Comment #9 & 13 V50 b. Indicated voltage is less b. Indicated voltage is less V51 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 NEI 99-01 Rev. 6 DAEC* Change Justification Validation

1. 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 V, 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 a:: IXI Regulations Regulations IXI <C CTMT/CNMT

... Containment Difference Not used N/A C z CSF ...... Critical Safety Function Difference Not used N/A <C V, CSFST ... Critical Safety Function Difference Not used N/A 2 > Status Tree z DBA ...... Design Basis Accident Difference Not used 0 N/A a:: u DC. ....... Direct Current DC. ....... Direct Current Verbatim N/A <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 c System System z ECL... .... Emergency Classification ECL. ...... Emergency Classification Verbatim w N/A 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 NEl,99-01 Rev. 6 DAEC . Change Justification Validation

1. 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 V, 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 Cl: N/A a:i a:i IC. .........

lnitiating Condition IC. ......... lnitiating Condition Verbatim N/A <( C ID ......... lnside Diameter ID ......... lnside Diameter Verbatim N/A z <( IPEEE ... lndividual Plant Difference Not used N/A V, Examination of External Events 2 > (Generic Letter 88-20) z 0 ISFSl. ... lndependent Spent Fuel ISFSl. ... lndependent Spent Fuel Verbatim N/A Cl: u Storage Installation Storage Installation <( I Keff ..... Effective Neutron Keff ..... Effective Neutron Verbatim N/A <( X Multiplication Factor Multiplication Factor c LCO ..... Limited Condition of LCO ..... Limited Condition of Verbatim N/A z w Operation Operation c.. c.. LOCA ... Loss of Coolant Accident LOCA ... Loss of Coolant Accident Verbatim <( N/A 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 s e i::t iC> rf .* ' ' iN1;1,:99-01 Rev:* 6' / --.

• DAtC, : ; :*.change,;~.,;}

' *. ' Iustificatibn C> ' . , \Valid-~tion 1P' , ',, ' .,,, *.' ':* 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 -+i NUMARC. ... Nuclear Management NUMARC. ... Nuclear Management Verbatim N/A C: 0 and Resources Council and Resources Council V) OBE ..... Operating Basis OBE ..... Operating Basis z Verbatim N/A 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 0:: cc Calculation (Assessment) Manual Manual cc <t: ORO ..... Offsite Response Difference Not used N/A 0 z Organization <t: PA ......... Protected Area PA ......... Protected Area Verbatim N/A V) PACS .... Priority Information and Difference Not used N/A > z Control System 0 0:: 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 25 Control System 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 Secti6n 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 -+"' RVLIS ... Reactor Vessel Level Difference Not used N/A C: 0 Instrumentation System V) RWCU ... Reactor Water Cleanup RWCU ... Reactor Water Cleanup Verbatim N/A 2 0 SAR ....... Safety Analysis Report Difference Not used N/A SAS ........ Safety Automation Difference Not used N/A > LU System a:: Difference cc SBO ....... Station Blackout Not used N/A cc <C SCBA ..... Se If-Contained Breathing SCBA .... .Self-Contained Breathing Verbatim N/A C 2 Apparatus Apparatus <C SG .......... Steam Generator Difference Not used N/A V) SI. .......... Safety Injection Difference Not used N/A > 2 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 Difference Not used i5 SRO ....... Senior Reactor Operator N/A 2 TEDE ..... Total Effective Dose TEDE ..... Total Effective Dose Verbatim N/A LU Q. Equivalent Equivalent Q. <C TAF ..... Top of Active Fuel Difference TOAF ..... Top of Active Fuel 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 IC

• NEI 99:-01 Rev:*6 DAEC ... Change Justiffcation

'Validation

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 2 have occurred which involve actual or have occurred which involve actual or 0 E IMMINENT substantial core degradation or IMMINENT substantial core degradation or 2 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 o 2 reasonably expected to exceed EPA PAG reasonably expected to exceed EPA PAG w c.. exposure levels offsite for more than the exposure levels offsite for more than the c.. <C 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 potential 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 IC , DAEC *. Change

• Justification Validation

1. 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 II) PAG exposure levels beyond the site PAG exposure levels beyond the site z boundary. boundary. 0 E Emergency Action Level (EAL): A pre-Emergency Action Level (EAL}: A pre-Verbatim None z determined, site-specific, observable determined, site-specific, observable u: 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 al X given emergency classification level. given emergency classification level. c z 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 . :. Change Justifi~ation
  • > ,, . Validatipn

• 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'l specific definition for this term.) Developer between spent fuel and the environment and added site-specific 2 Note -The barrier(s) between spent fuel and once the spent fuel is processed for dry language. 0 E the environment once the spent fuel is storage. This corresponds to the pressure 2 processed for dry storage. boundary for the Dry Shielded Canister (DSC) u::: w shell (including the inner bottom cover plate) C I base metal and associated confinement cc X boundary welds. c 2 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. <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 IC NEI 99-01 Rev. 6 DAEC Change .,_f'" .. , Justification, 1 ".,4 ~.

• Validation

1. . 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 VI inspection to determine if the 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 u: Term not used for BWRs None w generator that has a steam leak on the C I secondary side of sufficient size to cause an cc X uncontrolled drop in steam generator c z pressure or the steam generator to become w C. completely depressurized. 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 offlame 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 NEI 99-01 Rev. 6 DAEC Change: Justification .Validation n* 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 NEI 99-01 Rev. 6 DAEC Change Justification*

• Validation

1. 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 of the 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 V) Note -This term is typically taken to mean control of the licensee. Notes deleted. z 0 the site property owned by, or otherwise E z under the control of, the licensee. In some u: cases, it may be appropriate for a licensee to LI.I C define a smaller area with a perimeter closer I r::c to the plant Protected Area perimeter (e.g., a X 25 site with a large OCA where some portions of z the boundary may be a significant distance LI.I Q. Q. 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 )C NEI 99-0l_Rev. 6 DAEC Ch,ange . .Justifi.catiOIJ . *Validation.#. 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 VI leakage is of sufficient magnitude to require a z safety injection. Developer Note -This term 0 E is applicable to PWRs only. z u::: w SAFETY SYSTEM: A system required for safe SAFETY SYSTEM: A system required for safe Difference Removed developer notes C None I plant operation, cooling down the plant plant operation, cooling down the plant and clarified last sentence. cc X and/or placing it in the cold shutdown and/or placing it in the cold shutdown 25 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.. <C 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 IC NEI 99-01 Rev. 6 DAEC , Change . J1,.1stification

Validation.#*

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 II) change or event may be known or unknown. change or event may be known or unknown. 2 0 E VISIBLE DAMAGE: Damage to a component or VISIBLE DAMAGE: Damage to a component or Deviation Updated to reflect V22/47 2 u: structure that is readily observable without structure that is readily observable without wording and guidance of w measurements, testing, or analysis. The visual measurements, testing, or analysis. The visual C approved EAL FAQ 2016-I impact of the damage is sufficient to cause impact of the damage is sufficient to cause 02. The updated wording CQ X concern regarding the operability or concern regarding the operability or clarifies damage c 2 reliability of the affected component or reliability of the affected component or assessment meriting an w Q. structure. structure. Damage resulting from an ALERT declaration as used Q. <t equipment failure and limited to the failed in I Cs 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 98 > 'P C: Sectior1 Q,J <( E w ......... Q,J Ill C. I "C u~ )( Q,J *-:I "'C .... C: Q,J Q.J C C. C. <( NEI 99-01 Rev. 6 ' #fr ' . _:f Appendix C -Permanently Defueled ICs/EALs DAEC DEVIATIONS AND DIFFERENCES MATRIX DA~C Change ," J ustific~tion . Validation#> ,,,.",: *-* Not used at DAEC Difference Not applicable to DAEC None 99 ATTACHMENT 4 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER LICENSE AMENDMENT REQUEST TSCR-166 SUPPORTING TECHNICAL INFORMATION 503 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 :<:,; 212 NA (b) One or more reactor vessel head closure bolts less than fully tensioned. DAEC 1.1-8 Amendment 223 INTERIM STAFF GUIDANCE EMERGENCY PLANNING FOR NUCLEAR POWER PLANTS licensee to promptly declare the emergency condition as soon as possible following the ident ifi cation of the appropriate ECL. As used here , " promptly" means the next available opportunity unimpeded by activities not related to the emergency declaration , unless such activities are necessary for protecting health and safety. (See Paragraph 8 of this section.) 6. Consistent with the NRC's position that emergency declarations are made promptly , the final rul e states that the 15-minute criterion not be construed as a grace period in which a licensee may attempt to resto r e plant conditions to avoid declaring an EAL that has already been exceeded. This statement does not preclude licensees from acting to correct or mitigate an off-normal condition , but once an EAL has been recognized as being exceeded , the emergency declaration shall be made promptly without waiting for the 15-minute period to elapse. This is particularly the case when the EAL threshold is exceeded based on occurrence of a condition , rather than the duration of a condition. 7. For EAL thresholds that specify a duration of the off-normal condition , the NRC expects that the emergency declaration process run concurrently with the specified threshold dura t ion. Once the off-normal condition has existed for the duration specified in the EAL , no further effort on this declarat i on is necessary-the EAL has been exceeded. Consider as an example , the EAL " fire wh i ch is not extinguished with i n 15 minutes of detection." On receipt of a fire alarm, the plant fire brigade is dispatched to the scene to begin fire suppression efforts.

• If the fire brigade reports that the fire can be extinguished before the specified duration , the emergency declaration is placed on hold while firefighting activities continue. If the fire brigade is successful in extinguishing the fire within the specified duration from detection , no emergency declaration is warranted based on that EAL.
• If the fire is still burning after the specified duration has elapsed , the EAL is exceeded , no further assessment is necessary , and the emergency declaration would be made promptly.

As used here , "promptly" means at the first available opportunity (e.g., if the Shift Manager is receiving an update from the fire brigade at the 15-minute mark, it is expected that the declaration will occur as the next action after the call ends).

• If , for example , the fire brigade notifies the shift supervision 5 minutes after detection that the brigade itself cannot extinguish the fire such that the EAL will be met imm i nently and cannot be avoided , the NRC would not consider it a violation of the licensee's emergency plan to declare the event before the EAL is met (e.g., the 15-minute duration has elapsed).

While a prompt declaration would be beneficial to public health and safety and is encouraged , it is not required by regulation.

• In all of the above , the fire duration is measured from the time the alarm , indication , or report was first received by the plant operators.

Validation or confirmation establishes that the fire started as early as the time of the alarm , indication , or report. NSIR/DPR-ISG-01 Rev. 0 (November 2011) DAEC EOP BASES DOCUMENT EOP 3 -SECONDARY CONTAINMENT CONTROL GUIDELINE SF/L-3 D DISCUSSION SF/L-2 D Spent Fuel Pool level cannot be maintained above 37 ft 1 in. Maintain 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 tne 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 top of the fuel racks. 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 Steo cool down? ' IPOl3, 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. 2nd Stage Reheat can be left in N/A N/A N/A N/A Section 5, approximately 200 MWe, remove the 1 E-service and the turbine can be step (10) 1 BA[B] 2nd Stage Reheat System from tripped if necessary. service 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. will automatically go into the "hold" -step (10) mode as power and flow are lowered. IPOl4, Commence primary containment purge No. This is only necessary if a N/A N/A N/A N/A Section 3 per 01 573. Drywell entry is anticipated. -step (11) IP014, At the refueling bridge, verify that the Main, No. Control rod insertion will not be N/A N/A N/A

• N/A Section 3 Disconnect is closed and that the inhibited.

step (13) SYSTEM START pushbutton has been depressed. IPOl4, Prior to disconnecting the generator from No. Aux Boiler is not required to N/A N/A N/A N/A Section 3 the grid, perform the following: (a) If accomplish shutdown. step (14) needed, start up the Auxiliary Boiler per 01 727. IP014, Following Turbine Trip: (a) Verify that No. These systems can be left in N/A N/A N/A N/A Section 3 Reactor Coolant Chloride and service if necessary. step (22) Conductivity analyses have been performed. (b) Operate the Turbine Lube Oil 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 cool down? IPOl4, Shut down the following generator support No. These systems can be left in N/A N/A N/A N/A Section 3 systems, as desired: Isolated Phase Bus service if necessary.

step (24) Cooling -01 698, Stator Water Cooling -01 697, H2 Seal Oil -01 695.1, H2 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 per 01 563, Hydrogen Water Chemistry System will secure itself step (26) Chemistrv. if left in service. IPOl4, As directed by the CRS, perform the No. The MSIVs can be closed if N/A N/A N/A NIA Section 3 following steps as necessary to limit necessary to limit plant cooldown step (27) reactor vessel depressurization following rate. the reactor scram: (b) Start 1 P32 Mechanical Vacuum Pump per 01 691. (c) Secure the SJAEs and Offgas per 01 691 and 01672. IPOl4, For the remainder 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 depressurize the reactor plant cooldown rate. vessel to maintain a controlled cooldown (b) No -operated from the rate less than the TS Limit of 100°F in any Control Room 1 hour period. (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 with RCIC per 01 150 closed if necessary to limit (preferred method if 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 limit pressure lowers) (d) Control steam flow plant cooldown rate. from the reactor vessel to the main condenser through steam seals and steam drains, (e) Secure steam seals per OI 692 as required .to limit 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 Stec coo Id own? IPOl4, As plant cooldown continues perform the No. The MSIVs can be closed if N/A N/A N/A

• N/A Section 4 following: (NA if MSIVs are closed) (a) necessary to limit plant cooldown step (7) Control steam seal pressure 3 to 4 psig rate. using M0-1169, MAIN STEAM SUPPLY, M0-1170, REGULATOR BYPASS and/or M0-1171, MANUAL UNLOADER on 1C07, (b) Start 1P-32 MECHANICAL VACUUM PUMP per 01 691, (c) When reactor pressure approaches 500 psig or cooldown rate cannot be controlled within the limit, then secure SJAEs and Offgas System per 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 psig above reactor pressure as reactor pressure decreases.

Otherwise, N/A. IP014, At approximately 400 psig, secure the No. The Feed Pumps and N/A N/A N/A N/A Section 4 operating feed pump per 01 644. Condensate Pumps can be tripped step (8) from the Control Room if necessarv. IPOl4, When RHR Shutdown Cooling Isolation No, this system can be placed in NIA NIA N/A N/A Section 4 Interlocks can be reset service from the Control Room if step (9) (approximately 100 psig), reset the necessary. isolation, then initiate Shutdown Cooling per 01149. IPOl4, Perform the following after the turbine trip, No. These systems can be left in N/A N/A N/A N/A Section 4 if needed: (a) Verify that Reactor Coolant service if necessary. step (10) Chloride and Conductivity analysis has been performed, (b) Operate the Turbine Lube Oil and Turning Gear System per OI 693.3, (c) Shutdown the Main Generator per 01 698, (d) Shutdown the Main Turbine per 01 693.1. Procedure Step Action If action not performed, does Building Elevation Room Mode Section this prevent shutdown or and Steo 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) H2 Seal Oil per 01 695.1, (d) H2 and CO2 Gas per 01 695.2, (e) Secure SJAEs per 01 691 and Offgas per 01 672 if not oreviouslv oerformed. IP014, 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. IPOl4, 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. IPOl4, 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. IPOl4, 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 steo (15) Svstem oer 01 692. rate. IP014, 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. necessarv. Conclusion of manual action evaluation for EALs RA3 and HAS is shown below: EALs RA3 and HAS 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. 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) 1G-31 and 1G-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. Th e Cl a ss 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, 1A3 and 1A4, 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 (1X3) provides the normal source of power to the essential buses 1A3 and 1A4. 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 for 4.16 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 BASES BACKGROUND (continued) APPLICABLE SAFETY ANALYSES LCO DAEC AC Sources -Operating B 3.8.1 connected loads needed to recover the unit or maintain it in a safe condition are returned to service. Ratings for the DGs satisfy the intent of Safety Guide 9 as discussed in UFSAR Section 1.8.9 (Ref. 3). DGs 1 G-31 and 1 G-21 have the following ratings: a. 2850 kWOcontinuous, b. 3000 kW02000 hours, and c. 3250 kW 0300 hours. The initial conditions of OBA and transient analyses in the UFSAR, Chapter 15 (Ref. 5), assume ESF Systems are OPERABLE. The AC electrical power sources are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ESFSystems so that the fuel, Reactor Coolant System (RCS), and containment design limits are not exceeded. These limits are discussed in more detail in the Bases for Section 3.2, Power Distribution Limits; Section 3.5, Emergency Core Cooling Systems (ECCS) and Reactor Core Isolation Cooling (RCIC) System; and Section 3.6, Containment Systems. The OPERABILITY of the AC electrical power sources is consistent with the initial assumptions of the accident analyses and is based upon meeting the design basis of the unit. This includes maintaining the onsite or offsite AC sources OPERABLE during accident conditions in the event of: a. An assumed loss of all offsite power or all onsite AC power; and b. A worst case single failure. AC sources satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). Two qualified circuits between the offsite transmission network and the onsite Class 1 E AC Electric Power Distribution System and two separate and independent DGs (continued) B 3.8-3 TSCR-044A BASES LCO (continued) DAEC AC Sources -Operating B 3.8.1 (1 G-31 and 1 G-21) ensure availability of the required power to shut down the reactor and maintain it in a safe shutdown condition after an Abnormal Operational Transient or a postulated OBA. Qualified offsite circuits are those that are described in the UFSAR, and are part of the licensing basis for the unit. Each offsite circuit must be capable of maintaining rated frequency and voltage, and accepting required loads during an accident, while connected to the essential buses. In accordance with commitments made in response to Generic Letter 2006-02 (Ref. 4), Condition C is entered whenever the grid operator* (e.g., Midwest Independent System Operator (MISO)) determines that offsite power grid conditions are such that a trip of the DAEC turbine/generator would lead directly to voltages in the DAEC switchyard below the trip setpoints for Loss of Power (LOP) Instrumentation (LCO 3.3.8.1). The two offsite circuits consist of: 1) the incoming autotransformer (T1) and disconnect (1401, 6782, 2812 or 4731), the incoming circuit breaker (8490) and disconnect (8491), the underground 34.5 kV line, the standby transformer (1X4), the 4160 volt supply line and the two supply circuit breakers (1A301 and 1A401) to essential buses 1A3 and 1A4, respectively, and 2) either the incoming circuit breaker (5550) and disconnects (5551 and 5552) or incoming circuit breaker (5560) and disconnects (5553 and 5555), the overhead 161 kV line, the startup transformer (1X3), the two 4160 volt supply lines and the two supply circuit breakers (1A302 and 1A402) to essential buses 1A3 and 1A4, respectively. Each DG must be capable of starting, accelerating to rated speed and voltage, and connecting to its respective essential bus on detection of bus undervoltage. This sequence must be accomplished within 10 seconds. Each DG must also be capable of accepting required loads within the assumed loading sequence intervals, and must continue to operate until offsite power can be restored to the essential buses. Proper sequencing of loads, including non-essential load shedding capability, is a required function for DG OPERABILITY. The AC sources must be separate and independent (to the extent possible) of other AC sources. For the DGs, the separation and independence are complete. For the offsite AC sources, the separation and independence are to the extent practical. A circuit may be connected to more than (continued) B 3.8-4 TSCR-082 BASES LCO (continued) APPLICABILITY ACTIONS DAEC AC Sources -Operating B 3.8.1 one essential bus, with slow transfer capability to the other circuit OPERABLE, and not violate separation criteria. A circuit that is not connected to either essential bus is required to have OPERABLE slow transfer interlock mechanisms to both essential buses to support OPERABILITY of that circuit. The AC sources are required to be OPERABLE in MODES 1, 2, and 3 to ensure that: a. Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of Abnormal Operational Transients; and b. Adequate core cooling is provided and containment OPERABILITY and other vital functions are maintained in the event of a postulated OBA. The AC power requirements for MODES 4 and 5 and other Conditions in which AC sources are required are covered in LCO 3.8.2, "AC Sources -Shutdown." A Note prohibits the application of LCO 3.0.4.b to an inoperable DG. There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an inoperable DG and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance. To ensure a highly reliable power source remains with one offsite drcuit inoperable, it is necessary to verify the availability of the remaining required offsite circuit on a more fre-quent basis. Since the Required Action only specifies "perform," a failure of SR 3.8.1.1 acceptance criteria does not result in a Required Action not met. However, if a second required circuit fails SR 3.8.1.1, the (continued) B 3.8-5 TSCR-082 BASES ACTIONS DAEC A.1 (continued) AC Sources -Operating B 3.8.1 second offsite circuit is inoperable, and Condition C, for two offsite circuits inoperable, is entered. The power sources for the plant auxiliary power system are sufficient in number and have adequate electrical and physical independence to ensure that no single probable event could interrupt all auxiliary power at one time. In the condition of one inoperable offsite power source, all essential and non-essential buses remain OPERABLE and the remaining offsite power source continues to provide a highly reliable power source. Required Action A.2 requires restoring the inoperable offsite circuit to OPERABLE status, prior to entering MODE 2 from MODE 3 or 4. The inoperable offsite circuit must be restored to OPERABLE status prior to entering MODE 2 from MODE 3 or 4 to ensure that at least two offsite power sources will be available before the reactor is taken beyond just critical. Entry into MODE 1 from MODE 2 with an inoperable offsite circuit is acceptable since LCO 3.0.4 allows continued operation of the unit in a MODE or other specified condition in which operation for an unlimited period of time is allowed. The inoperable offsite circuit only has to be repaired prior to entering MODE 2 from MODE 3 or 4. To ensure a highly reliable power source remains with one DG inoperable, it is necessary to verify the availability of the offsite circuits on a more frequent basis. Since the Required Action only specifies "perform," a failure of SR 3.8.1.1 acceptance criteria does not result in a Required Action being not met. However, if a circuit fails to pass SR 3.8.1.1, it is inoperable. Upon offsite circuit inoperability, additional Conditions must then be entered. (continued) B 3.8-6 TSCR-082 BASES AC Sources -Operating B 3.8.1 ACTIONS B.2 (continued) DAEC Required Action 8.2 is intended to provide assurance that a loss of offsite power, during the period that a DG is inoperable, does not result in a complete loss of safety function of critical systems. These features are designed with redundant safety related divisions (i.e., single division systems are not included). Failures of "redundant required features" refers to inoperable features associated with a division redundant to the division that has an inoperable DG. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowable out of service time "clock." In this Required Action the Completion Time only begins on discovery that both: a. An inoperable DG exists; and b. A required feature on the other division (Division 1 or 2) is inoperable. If, at any time during the existence of this Condition (one DG inoperable), a required feature subsequently becomes inoperable, this Completion Time begins to be tracked. Discovering one required DG inoperable coincident with one or more inoperable required support or supported features, or both, that are associated with the OPERABLE DG results in starting the Completion Time for the Required Action. Four hours from the discovery of these events existing concurrently is acceptable because it minimizes risk while allowing time for restoration before subjecting the unit to transients associated with shutdown. (continued) B 3.8-7 Amendment 223 BASES ACTIONS OAEC 8.2 (continued) AC Sources -Operating B 3.8.1 The remaining OPERABLE OG and offsite circuits are adequate to supply electrical power to the onsite Class 1 E Distribution System. Thus, on a component basis, single failure protection for the required feature's function may have been lost; however, function has not been lost. The 4 hour Completion Time takes into account the component OPERABILITY of the redundant counterpart to the inoperable required feature. Additionally, the 4 hour Completion Time takes into account the capacity and capability of the remaining AC sources, reasonable time for repairs, and low probability of a OBA occurring during this period. Required Action 8.3 requires that the cause of the inoperability be evaluated to ensure a common cause failure does not exist that could render the OPERABLE OG inoperable. This evaluation may be performed by analysis or inspection or by demonstration of OPERABILITY. If the cause of inoperability exists on the other OG, it is declared inoperable upon discovery, and Condition O of LCO 3.8.1 is entered. Once the failure is repaired, and the common cause failure no longer exists, Required Action 8.3 is satisfied. If the cause of the initial inoperable OG cannot be confirmed not to exist on the remaining OG, SR 3.8.1.2 can be performed within the* same Completion Time as Required Action 8.3 to provide assurance of continued OPERABILITY of the remaining OG. Conversely, Required Action 8.3 may be satisfied by a simple review when the cause of the initial inoperability is pre-planned, preventive maintenance and testing. It is also permissible to perform elective maintenance as part of the pre-planned, preventive maintenance. In this case, there is no potential for a common cause failure, as no failure has occurred. At any point during the pre-planned maintenance or testing, if any new failure is detected, which on its own would cause the EOG to be inoperable and was not maintenance induced, the common cause evaluation ml,Jst be re-performed on the Operable OG. If the 24 hour Completion Time for Required Action B.3 has expired at the point of discovery of the failure requiring corrective maintenance, Condition E must be entered until the common cause evaluation or SR 3.8.1.2 is performed. (continued) B 3.8-8 TSCR-101A BASES ACTIONS (continued) OAEC B. 3 (continued) AC Sources -Operating B 3.8.1. In the event the inoperable OG is restored to OPERABLE status prior to completing 8.3, the plant corrective action program will continue to evaluate the common cause possibility. This continued evaluation, however, is no longer under the 24 hour constraint imposed while in Condition B. According to Generic Letter 84-15 (Ref. 7), 24 hours is a reasonable time to confirm that the OPERABLE OG is not affected by the same problem as the inoperable OG To ensure the continued OPERABILITY of the remaining OG . during the 7 day Completion Time of Required Action 8.5, SR 3.8.1.2 must be performed once per 72 hours for the OPERABLE OG. The 72 hour Completion Time is acceptable since it has already been determined that a common cause failure does not exist. Required Action 8.4 is modified by a Note that removes this requirement when the cause of the initial inoperability is planned, preventive maintenance and testing. It is also permissible to perform elective maintenance as part of the planned preventive maintenance. In this case, no actual failure has occurred (i.e., a potential for a common mode failure has not been identified) and the likelihood of the other OG having an undetected failure during this period is low. At any point during the pre-planned maintenance, if any new failure requiring corrective maintenance is detected, i.e., which on its own would cause the EOG to be inoperable and was not maintenance induced, Required Action 8.4 must be entered for the Operable OG. If the 72 hour Completion Time has expired at the point of discovery of the failure requiring corrective maintenance, then Corrective E must be entered until SR 3.8.1.2 is performed. 8.5 . In Gondition 8, the remaining OPERABLE OG_and offsite circuit(s) are adequate to supply electrical power to the onsite Class 1 E Distribution System. The 7 day Completion Time takes into account the capacity and capability of the remaining AC sources, reasonable time for repairs, and low probability of a OBA occurring during this period. (continued) B 3.8-9 TSCR-101A BASES ACTIONS DAEC 8.5 (continued) AC Sources -Operating B 3.8.1 The second Completion Time for Required Action 8.5 establishes a limit based on the maximum time allowed for the combination of one DG and two offsite AC power sources to be inoperable during any single contiguous occurrence of failing to met the LCO except for Action A. If Condition B is entered while, for instance, two offsite circuits are inoperable and one circuit is subsequently restored OPERABLE, the LCO may already have been not met for up to 24 hours. This situation could lead to a total of 8 days, since initial failure of the LCO (except for Condition A), to restore the DG. At this time, the second offsite circuit could again become inoperable, the DG restored OPERABLE, and an additional 24 hours (for a total of 9 days) allowed prior to complete restoration of the LCO (except for Condition A). The 8 day Completion Time provides a limit on the time allowed in a specified condition after discovery of failure to meet the LCO. This limit is considered reasonable for situations in which Conditions Band Care entered concurrently, and when corrective actions are completed prior to completing the shutdown required by LCO 3.0.3 (which is required to be entered by Action F). The "AND" connector between the 7 day and 8 day Completion Times means that both Completion Times apply simultaneously, and the more restrictive must be met. As in Required Action 8.2, the Completion Time allows for an exception to the normal "time zero" for beginning the allowable out of service time "clock." This exception results in establishing the "time zero" at the time that the LCO was initially not met, instead of the time that Condition B was entered. C.1 and C.2 As noted above, this Condition is entered whenever the grid operator informs the DAEC that a unit trip could lead to degraded voltage conditions in the DAEC Switchyard (Ref. 4). Required Action C.1 addresses actions to be taken in the event of inoperability of redundant required features concurrent with inoperability of two offsite circuits. Required Action C.1 reduces the vulnerability to a loss of function. The Completion Time for taking these actions is 12 hours. When a concurrent redundant required feature failure exists, a shorter Completion Time of 12 hours is appropriate. These features are designed with redundant safety related divisions, (i.e., single division systems are not included in the list). Redundant required features failures consist of any of these features that are inoperable because any (continued) B 3.8-10 TSCR-101A BASES ACTIONS DAEC AC Sources -Operating B 3.8.1 C.1 and C.2 (continued) inoperability is on a division redundant to a division with inoperable offsite circuits. The Completion Time for Required Action C.1 is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowable out of service time "clock." In this Required Action, the Completion Time only begins on discovery that both: a. All offsite circuits are inoperable; and b. A required feature is inoperable. If, at any time during the existence of this Condition (two offsite circuits inoperable), a required feature subsequently becomes inoperable, this Completion Time begins to be tracked. According to the recommendations contained in Regulatory Guide 1.93 (Ref. 6), operation may continue in Condition C for a period that should not exceed 24 hours. This level of degradation means that the Offsite Electrical Power System does not have the capability to effect a safe shutdown and to mitigate the effects of an accident; however, the onsite AC sources have not been degraded. This level of degradation generally corresponds to a total loss of the immediately accessible offsite power sources. Because of the normally high availability of the offsite sources, this level of degradation may appear to be more severe than other combinations of two AC sources inoperable that involve one or more DGs inoperable. However, two factors tend to decrease the severity of this degradation level: a. The configuration of the r_edundant AC Electrical Power System that remains available is not susceptible to a single bus or switching failure; and b. The time required to detect and restore an unavailable offsite power source is generally much less than that required to detect and restore an unavailable onsite AC source. (continued) B 3.8-11 TSCR-101A BASES ACTIONS DAEC C.1 and C.2 (continued) AC Sources -Operating B 3.8.1 With both of the offsite circuits inoperable, sufficient onsite AC sources are available to maintain the unit in a safe shutdown condition in the event of a OBA or transient. In fact, a simultaneous loss of offsite AC sources, a LOCA, and a worst case single failure were postulated as a part of the design basis in the safety analysis. Thus, the 24 hour Completion Time in Required Action C.2 provides a period of time to effect restoration of one of the offsite circuits commensurate with the importance of maintaining an AC Electrical Power System capable of meeting its design criteria. According to the recommendations contained in Regulatory Guide 1.93 (Ref. 6), with the available offsite AC sources two less than required by the LCO, operation may continue for 24 hours. If two offsite sources are restored within 24 hours, unrestricted operation may continue. If only one offsite source is restored within 24 hours, power operation continues in accordance with Condition A. The second Completion Time for Required Action C.2 establishes a limit based on the maximum time allowed for the combination of one DG and two offsite AC power sources to be inoperable during any single contiguous occurrence of failing to meet the LCO except for Condition A. If Condition C is entered while, for instance, one DG is inoperable and the DG is subsequently restored OPERABLE, the LCO may already have been not met for up to 7 days. This situation could lead to a total of 8 days, since initial failure of the LCO (except for Condition A), to restore one of the two inoperable offsite circuits. At this time, a DG could again become inoperable, one offsite circuit restored OPERABLE, and an additional 7 days (for a total of 15 days) allowed prior to complete restoration of the LCO (except for Condition A). The 8 day Completion Time provides a limit on the time allowed in a specified condition after discovery of failure to meet the LCO. This limit is considered reasonable for situations in which Conditions B and C are entered concurrently, and when corrective actions are completed prior to completing the shutdown required by LCO 3.0.3 (which is required to be entered by Action F). The "AND" connector between the 24 hours and 8 day Completion Times means that both Completion Times apply simultaneously, and the more restrictive must be met. (continued) B 3.8-12 TSCR-101A BASES ACTIONS (continued) DAEC C. 1 and C.2 (continued) AC Sources -Operating B 3.8.1 This Completion Time allows for an exception to the normal "time zero" for beginning the allowable out of service time "clock." This exception results in establishing the "time zero" at the time that the LCO was initially not met (except for condition A), instead of the time that Condition C was entered. With two DGs inoperable, there is no remaining standby AC source. Thus, with an assumed loss of offsite electrical power, insufficient standby AC sources are available to power the minimum required ESF functions. Since the offsite electrical power system is the only source of AC power for the majority of . ESF equipment at this level of degradation, the risk associated

• with continued operation for a very short time could be less than that associated with an immediate controlled shutdown. (The immediate shutdown could cause grid instability, which could result in a total loss of AC power.) Since any inadvertent unit generator trip could also result in a total loss of offsite AC power, however, the time allowed for continued operation is severely restricted.

The intent here is to avoid the risk associated with an immediate controlled shutdown and to minimize the risk associated with this level of degradation. According to.the recommendations contained in Regulatory Guide 1.93 (Ref. 6), with both DGs inoperable, operation may continue for a period that should not exceed 2 hours. E.1 and E.2 If the inoperable AC electrical power sources cannot be restored to OPERABLE status within the associated Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. (continued) B 3.8-13 TSCR-101A BASES AC Sources -Operating B 3.8.1 ACTIONS F.1 (continued) SURVEILLANCE REQUIREMENTS DAEC Condition F corresponds to a level of degradation in which all redundancy in the AC electrical power supplies has been lost. At this severely degraded level, any further losses in the AC Electrical Power System will cause a loss of function. Therefore, no additional time is justified for continued operation. The unit is required by LCO 3.0.3 to commence a controlled shutdown. The AC sources are designed to permit inspection and testing of all important areas and features, especially those that have a standby function, in accordance with UFSAR Section 3.1.2.2.9 (Ref. 8). Periodic component tests are supplemented by extensive functional tests during refueling outages (under simulated accident conditions). The SRs for demonstrating the OPERABILITY of the DGs are largely in accordance with the recommendations of Safety Guide 9 as discussed in UFSAR Section 1.8.9 (Ref. 3), Regulatory Guide 1.108 (Ref. 9), and Regulatory Guide 1.137 (Ref. 10) or as addressed in the UFSAR. Where the SRs discussed herein specify voltage and frequency tolerances, the following summary is generally applicable. The minimum steady state output voltage of 37 44 V is approximately 90% of the nominal 4160 V output voltage. This value, which is specified in ANSI C84.1 (Ref. 11), allows for voltage drop to the terminals of 4000 V motors whose minimum operating voltage is specified as 90% or 3600 V. It also allows for voltage drops to motors and other equipment down through the 120 V level where minimum operating voltage is also usually specified as 90% of name plate rating. This value also provides a large margin of safety, since safety related motors are capable of accelerating their loads at 70% of rated voltage (2912 V or 322 V). The specified maximum steady state output voltage of 4576 V or 110% of 4160 V is less than the maximum operating voltage specified for 4000 V motors (4756 V), and therefore also provides ample margin. It ensures that for a lightly loaded distribution system, the voltage at the terminals of 4000 V motors is no more than the maximum rated operating voltages. The ~pecified minimum and maximum frequencies of the DG are 59.5 Hz and 60.5 Hz, respectively. These values are approximately equal to +/- 1 % of the 60 Hz nominal frequency and are conservative with respect to the recommendations found in Regulatory Guide 1.9 (Ref. 17). (continued) B 3.8-14 TSCR-101A BASES SURVEILLANCE REQUIREMENTS (continued) DAEC SR 3.8.1.1 AC Sources -Operating B 3.8.1 This SR ensures proper circuit continuity for the offsite AC electrical power supply to the onsite distribution network and availability of offsite AC electrical power. The breaker alignment verifies that at least the minimum required offsite power supply breakers are in their correct position to ensure that distribution buses and loads are connected to either the preferred power source or the alternate preferred power source and that appropriate independence of offsite circuits is maintained. This can be accomplished by verifying that an essential bus is energized, and that the status of offsite supply breakers that are displayed in the control room are correct. The status of manual disconnects is verified administratively. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Frequency is adequate since breaker position is not likely to change without the operator being aware of it and because its status is displayed in the control room. SR 3.8.1.2 and SR 3.8.1.7 These SRs help to ensure the availability of the standby electrical power supply to mitigate DBAs and transients and maintain the unit in a safe shutdown condition. To minimize the wear on moving parts that do not get lubricated when the engine is not running, these SRs have been modified by a Note (Note 2 for SR 3.8.1.2 and Note 1 for SR 3.8.1.7) to indicate that all DG starts for these Surveillances may be preceded by an engine prelube period and (for SR 3.8.1.2 only) followed by a warmup prior to loading. Note 3 to SR 3.8.1.2 allows delaying the entry into associated Conditions and Required Actions for up to two hours during the performance of the conditional surveillance required by Required Actions 8.3 or 8.4. This Note is necessary because to perform a slow start and warm up of the DG requires reducing the governor control setting to minimum and securing the generator field excitation. The governor control setting is gradually increased to bring the DG to synchronous speed and to allow for warmup. Once the DG is at synchronous speed, the generator field excitation is enabled and the DG is again capable of supplying the essential bus. During this warm up portion of the surveillance test, the DG is incapable of supplying the essential bus and is considered inoperable. (continued) B 3.8-15 TSCR-120 BASES AC Sources -Operating B 3.8.1 SURVEILLANCE SR 3.8.1.2 and SR 3.8.1.7 (continued) REQUIREMENTS DAEC After completion of the SR, the fuel racks to the DG are disabled to allow purging of any residual fuel oil from the cylinders. This also renders the DG inoperable. The two hours allowed by the Note minimizes the amount of time a DG is inoperable while providing enough time to perform the required Conditional Surveillance and avoids entering the shutdown actions of Condition E or F unnecessarily. For the purposes of this testing, the DGs are manually started from standby conditions. Standby conditions for a DG mean that the diesel engine coolant and oil are being continuously circulated and temperature is being maintained consistent with manufacturer recommendations. In order to reduce stress and wear on diesel engines during testing, the manufacturer of the DGs installed at the DAEC recommends a modified start in which the starting speed of the DG is limited, warmup is limited to this lower speed, and the DGs are gradually accelerated to synchronous speed prior to loading. These start procedures are the intent of Note 2 (SR 3.8.1.2). SR 3.8.1.7 requires that DG starts from standby conditions and achieves required voltage and frequency (i.e. -voltage 2:: 3744 V and frequency 2:: 59.5 Hz) within 10 seconds; and achieves steady state voltage 2:: 3744 V and ::s; 4576 V and frequency 2:: 59.5 Hz and ::s; 60.5 Hz. The 10 second start requirement supports the assumptions in the design basis LOCA analysis of UFSAR, Section 15.2.1 (Ref. 12). The 10 second start requirement is not applicable to SR 3.8.1.2 (see Note 3 of SR 3.8.1.2), when a modified start procedure as described above is used. If a modified start is not used, the 10 second start requirement of SR 3.8.1.7 applies. In addition to the SR requirements, the time for the DG to reach steady state operation, unless the modified DG start method is employed, is periodically monitored and the trend evaluated to identify degradation of governor and voltage regulator performance. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The normal Frequency for SR 3.8.1.2 is consistent with Safety Guide 9. The Frequency for SR 3.8.1.7 is a reduction in cold testing consistent with Generic Letter 84-15 (Ref. 7). These Frequencies provide adequate assurance of DG OPERABILITY, while minimizing degradation resulting from testing. (continued) B 3.8-16 TSCR-120 BASES SURVEILLANCE REQUIREMENTS (continued) DAEC SR 3.8.1.3 AC Sources -Operating B 3.8.1 This Surveillance verifies that the DGs are capable of synchronizing and can be manually loaded to 2750 kW and 2950 kW, providing a 200 kW range centered on the continuous duty rating of the DGs of 2850 kW. This range ensures that the DGs are tested at a load above the maximum expected accident load. A minimum run time of 60 minutes is required to stabilize engine temperatures, while minimizing the time that the DG is connected to the offsite source. Although no power factor requirements are established by this SR, the DG is normally operated at a power factor greater than 0.9 lagging. While a value of 0.8 is the design rating of the machine, the machine is operated at power factors greater than 0.9 for normal operations and greater than 0.8 for surveillance testing. The load limit is provided to avoid routine overloading of the DG. Routine overloading may result in more frequent teardown inspections in accordance with vendor recommendations in order to maintain DG OPERABILITY. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The normal Frequency for this Surveillance is consistent with Safety Guide 9. Note 1 modifies this Surveillance to indicate that diesel engine runs for this Surveillance may include gradual loading, as recommended by the manufacturer, so that mechanical stress and wear on the diesel engine are minimized. Note 2 modifies this Surveillance by stating that momentary transients because of changing bus loads do not invalidate this test. Similarly, momentary power factor transients above the limit do not invalidate the test. Note 3 indicates that this Surveillance should be conducted on only one DG at a time in order to avoid common cause failures that might result from offsite circuit or grid perturbations. Note 4 stipulates a prerequisite requirement for performance of this SR. A successful DG start must precede this test to credit satisfactory performance. (continued) B 3.8-17 TSCR-120 BASES SURVEILLANCE REQUIREMENTS (continued) DAEC SR 3.8.1.4 AC Sources -Operating B 3.8.1 This SR provides verification that the level of fuel oil in the day tank is at or above the level at which the day tank low level alarm is annunciated. This low level alarm should only be received if the automatic fuel oil transfer instrumentation is not functioning properly. The level is expressed as an equivalent volume in gallons, and is selected to ensure adequate fuel oil for a minimum of approximately one hour of DG operation at full load, considering a conservative fuel consumption rate. Verification that at least a one hour supply of fuel oil exists in a day tank provides assurance that a DG can operate continuously, and also allows the operating crew sufficient time to take corrective action should the automatic fuel oil transfer system not function properly. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Frequency is adequate to ensure that a sufficient supply of fuel oil is available, since low level alarms are provided and facility operators would be aware of any large uses of fuel oil during this period. SR 3.8.1.5 Microbiological fouling is a major cause of fuel oil degradation. There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Testing for water content and removal of water from the fuel oil day tanks as necessary, eliminates the necessary environment for bacterial survival. This is the most effective means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation. Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, and breakdown of the fuel oil by bacteria. Frequent checking for and removal of accumulated water, as necessary, minimizes fouling and provides data regarding the watertight integrity of the fuel oil system. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program, The Surveillance Frequencies meet the intent of Regulatory Guide 1.137 (Ref. 10). This SR is for preventive maintenance. The presence of water does not necessarily represent a failure of this SR provided that accumulated water is removed during performance of this Surveillance. (continued) B 3.8-18 TSCR-120 BASES SURVEILLANCE REQUIREMENTS (continued) DAEC SR 3.8.1.6 AC Sources -Operating B 3.8.1 This Surveillance demonstrates that each required fuel oil transfer pump operates and transfers fuel oil from its associated storage tank to its associated day tank. It is required to support continuous operation of standby power sources. This Surveillance provides assurance that the fuel oil transfer pump is OPERABLE, the fuel oil piping system is intact, the fuel delivery piping is not obstructed, and the controls and control systems for manual fuel transfer systems are OPERABLE. Additional assurance of fuel oil transfer pump OPERABILITY is provided by meeting the testing requirements for pumps that are contained in the ASME Boiler and Pressure Vessel Code, Section XI (Ref. 13). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. SR 3.8.1.7 See SR 3.8.1.2. SR 3.8.1.8 The slow transfer of each 4.16 kV essential bus power supply from the preferred offsite circuit (i.e. -the startup transformer) to the alternate preferred offsite circuit (i.e .. the standby transformer) demonstrates the OPERABILITY of the alternate preferred circuit distribution network to power the shutdown loads. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Frequency of the Surveillance is based on engineering judgment taking into consideration the plant conditions required to perform the Surveillance, and is intended to be consistent with expected fuel cycle lengths. Operating experience has shown that these components usually pass the SR when performed on this Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. This SR is modified by a Note. The reason for the Note is that, during operation with the reactor critical, performance of this SR could cause perturbations to the Electrical Distribution Systems that could challenge continued steady state operation and, as a result, plant safety systems. Credit may be taken for unplanned events that satisfy this SR. (continued) B 3.8-19 TSCR-120 BASES SURVEILLANCE REQUIREMENTS (continued) DAEC SR 3.8.1.9 AC Sources -Operating B 3.8.1 Each DG is provided with an engine overspeed trip to prevent damage to the engine. Recovery from the transient caused by the loss of a large load could cause diesel engine overspeed, which, if excessive, might result in a trip of the engine. This Surveillance demonstrates the DG load response characteristics and the

• capability to reject the largest single load and return to the required voltage and frequency (i.e. -voltage 2 3744 V and~ 4576 V and frequency 2 59.5 Hz and 60.5 Hz) within predetermined periods of time (i.e., 1.3 seconds for voltage and 3.9 seconds for frequency) while maintaining an acceptable margin to the overspeed trip. The largest single load for each DG is a core spray pump motor (700 hp). This Surveillance may be accomplished by tripping its associated single largest accident load with the DG solely supplying the bus. As specified by IEEE-308 (Ref. 14), the load rejection test is acceptable if the increase in diesel speed does not exceed 75% of the difference between synchronous speed and the overspeed trip setpoint, or 15% above synchronous speed, whichever is lower. For both DGs, this represents 64.5 Hz, equivalent to 75% of the difference between nominal speed and the overspeed trip setpoint.

The time, voltage, and frequency tolerances specified in the Bases for this SR are derived from UFSAR Table 8.3-1 (Ref. 16) recommendations for response during load sequence intervals. The voltage and frequency are consistent with the design range of the equipment powered by the DG. SR 3.8.1.9.a corresponds to the frequency excursion, while SR 3.8.1.9.b and SR 3.8.1.9.c are the steady state voltage and frequency to which the system must recover following load rejection within a predetermined time period. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Frequency is consistent with the recommendation of Regulatory Guide 1.108 (Ref. 9). This SR is modified by a Note. The reason for the Note is that, during operation with the reactor critical, performance of this SR could cause perturbations to the Electrical Distribution Systems that could challenge continued steady state operation and, as a result, plant safety systems. Credit may be taken for unplanned events that satisfy this SR. (continued) B 3.8-20 TSCR-120 BASES SURVEILLANCE REQUIREMENTS (continued) DAEC SR 3.8.1.10 AC Sources -Operating B 3.8.1 This Surveillance demonstrates that DG non-critical protective functions (e.g., high jacket water temperature and low lubricating oil pressure) are. bypassed on either an ECCS initiation test signal or a LOOP test signal and critical protective functions (engine overspeed and generator differential current) trip the DG to avert substantial damage to the DG unit. The non-critical trips are bypassed during DBAs and LOOPs and provide an alarm on an abnormal engine condition. This alarm provides the operator with sufficient time to react appropriately. The DG availability to mitigate the OBA is more critical than protecting the engine against minor problems that are not immediately detrimental to emergency operation of the DG. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Frequency is based on engineering judgment, takes into consideration plant conditions required to perform the Surveillance, and is intended to be consistent with expected fuel cycle lengths. Operating experience has shown that these components usually pass the SR when performed at this Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required DG from service. Credit may be taken for unplanned events that satisfy this SR. SR3.8.1.11 As specified by Regulatory Guide 1.108 (Ref. 9), paragraph 2.a.(6), this Surveillance ensures that the manual synchronization and load transfer from the DG to the offsite source can be made and that the DG can be returned to ready-to-load status when offsite power is restored. The DG is considered to be in load status when the DG is at rated speed and voltage, the output breaker is open and can receive an auto-close signal on bus undervoltage, and the individual pump timers are reset. (continued) B 3.8-21 TSCR-120 BASES AC Sources -Operating B 3.8.1 SURVEILLANCE SR 3.8.1.11 (continued) REQUIREMENTS DAEC The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Frequency is consistent with the recommendations of Regulatory Guide 1.108 (Ref. 9), paragraph 2.a.(6), and takes into consideration plant conditions required to perform the Surveillance. This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems. Credit may be taken for unplanned events that satisfy this SR. SR 3.8.1.12 Under either LOCA conditions or during a loss of offsite power, loads are sequentially connected to the bus by a timed logic sequence using individual time delay relays. The sequencing logic controls the permissive and starting signals to motor breakers to prevent overloading of the DGs due to high motor starting currents. Verifying the load sequence time interval is* greater than or equal to 2 seconds ensures that sufficient time exists for the DG to restore frequency and voltage prior to applying the next load. The Allowable Values for the Core Spray and Low Pressure Coolant Injection Pump Start -Time Delay Relays, Table 3.3.5.1-1, Functions 1.e and 2.e, ensure this time interval is maintained as well as ensuring that safety analysis assumptions regarding ESF equipment time delays are not violated. Allowances for instrument inaccuracies in the load sequence time interval are also accounted for by the Pump Start -Time Delay Relay Allowable Value. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Frequency is consistent with the recommendations of Regulatory Guide 1.108 (Ref. 9), paragraph 2.a.(2); takes into consideration plant conditions required to perform the Surveillance; and is intended to be consistent with expected fuel cycle lengths. This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems. Credit may be taken for unplanned events that satisfy this SR. (continued) B 3.8-22 TSCR-120 BASES SURVEILLANCE REQUIREMENTS (continued) DAEC SR 3.8.1.13 AC Sources -Operating B 3.8.1 In the event of a OBA coincident with a loss of offsite power, the DGs are required to supply the necessary power to ESF Systems so that the fuel, RCS, and containment design limits are not exceeded. This Surveillance demonstrates DG operation during a Loss of Offsite Power actuation test signal (LOOP signal) in conjunction with an ECCS initiation signal (LOCA signal). This test verifies all actions encountered from the LOOP/LOCA, including the LOOP/LOCA load shedding function and energization of the essential buses and respective loads from the DG. This Surveillance also demonstrates the as-designed operation of the standby power sources during a LOOP, including:

1) de-energization of the essential buses, 2) the dead bus load shedding function, and 3) that the DG receives a start signal. This surveillance also demonstrates that the DG automatically starts from the design basis actuation signal (LOCA signal). It further demonstrates the capability of the DG to automatically achieve the required voltage and frequency (i.e., voltage~ 3744 V and~ 4576 V and frequency~

59.5 Hz and 60.5 Hz) within the specified time (10 seconds). In lieu of multiple demonstrations of DG starting and achieving the required voltage and frequency in the specified time from each of the various start signals (LOOP, LOCA and LOOP/LOCA), and operation for~ 5 minutes, testing that adequately shows the capability of the DG to start from each of the signals is acceptable. The DG auto-start time of 10 seconds is derived from requirements of the accident analysis, (Ref. 12), for responding to a design basis large break LOCA. The I Surveillance should be continued for a minimum of 5 minutes (with a LOOP signal in conjunction with a LOCA signal present) in order to demonstrate that all of the starting transients have decayed and stability has been achieved. The requirement to verify the connection and power supply of permanent and auto-connected loads is intended to satisfactorily show the relationship of these loads to the DG loading logic. In certain circumstances, many of these (continued) B 3.8-23 TSCR-044 BASES AC Sources -Operating B 3.8.1 SURVEILLANCE SR 3.8.1.13 (continued) REQUIREMENTS REFERENCES DAEC loads cannot actually be connected or loaded without undue hardship or potential for undesired operation. For instance, ECCS injection valves are not desired to be stroked open, or systems are not capable of being operated at full flow. In lieu of actual demonstration of connection and loading of these loads, testing that adequately shows the capability of the DG system to perform these functions is acceptable. This testing may include any series of sequential, overlapping, or total steps so that proper operation with each of the various signals present is verified. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Frequency takes into consideration plant conditions required to perform the Surveillance and is intended to be consistent with an expected fuel cycle length. Operating experience has shown that these components usually pass the SR when performed at this Frequency. Therefore, the Frequency is acceptable from a reliability standpoint.

• This SR is modified by two Notes. The reason for Note 1 is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil being continuously circulated and temperature maintained consistent with manufacturer recommendations.

The reason for Note 2 is that performing the Surveillance would remove the required offsite circuit from service, perturb the Electrical Distribution System, and challenge safety systems. Credit may be taken for unplanned events that satisfy this SR. 1. UFSAR, Section 3.1.2.2.8.

2. UFSAR, Section 8.2.1.3 and Section 8.3.1.1.5

3. UFSAR, Section 1.8.9. 4. FPL letter, L-2006-073, dated April 3, 2006, Response to NRC Generic Letter 2006-02, Grid Reliability and the Impact on Plant Risk and the Operability of Offsite Power. 5. UFSAR, Chapter 15. (continued)

B 3.8-24 TSCR-120 BASES REFERENCES

6. (continued)

7. 8. 9. 10. 11. 12. 13. 14. 15. 16. 17. DAEC Regulatory Guide 1.93. Generic Letter 84-15. UFSAR, Section 3.1.2.2.9 Regulatory Guide 1.108. Regulatory Guide 1.137. [Deleted]

UFSAR, Section 15.2.1 AC Sources -Operating B 3.8.1 ASME Boiler and Pressure Vessel Code, Section XI. IEEE Standard 308. [Deleted] UFSAR, Table 8.3-1. Regulatory Guide 1.9. B 3.8-25 TSCR-082 AOP 301.1 STATION BLACKOUT STATION BLACKOUT FOLLOW-UP ACTIONS (continued)

18. Direct an operator and electricians (if available) to perform Attachment 10 Alternate AC power to 125VDC and 250VDC chargers. The inverters will automatically trip at 105 VDC decreasing. 19. Afte r one hour has elapsed , dispatch an operator to implement Attachment 13 , Load Shedding to Preserve Station Batteries (from AOP 301.1 hanging file). Attachment 13 is to be completed within two hours of the SBO event. 20. IF a SBDG is available for operation , THEN restore power to its essential bus pe r Restoration of Standby Diesel Generator Power section. 21. WHEN the DAEC Switchyard inspection has been completed , THEN restore power to the switchyard per Restoration of Offsite Power section. 22. WHEN sufficient offsite power becomes available , THEN restore power to the non-essential buses per AOP 304.1. I AOP 301.1 Page 7 of 45 Rev. 61 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 1A3 Essential Bus 1A4 480V Load Centers Load Centers 183 , 189 184, 1820 480V Motor Control Motor Control Centers Centers 1832 , 1834 1842 , 1844 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 1041 and 1042 (a) Each division of the AC and DC electrical power distribution systems is a subsystem. DAEC B 3.8-73 Amendment 223 DAEC EMERGENCY PLAN EMERGENCY COMMUNICATIONS FIGURE F-5 DAEC TELEPHONE SYSTEMS -------------1 Ir----, I I I I Operational Sup p ort Cente r (Access C on trol) Satell it e Communicat i ons Offsite Relocation a nd Assem b l Area (f)*EOF (1)

• FPLE Duane Amold COIJ)orate Offices
• County Sherffl's Offices
• Palo Fre Departmenl
• Mercy Hospbl
• State Highway Potrol
• State E mergency Ma n agement Division
• University of Iowa 'NRC 'DOE 'FEMA
• Lin n County Emergency Manageme n t
• Benton County Emergency Management 4 emergency unlisted Ines (Blue Phones) i n Conttol Room. TSC , CAS , SAS Local Telephone Co mp any Central Office Normal Telephone Services Qwest (Cedar Rapids) To other Bel Central Offices DAEC Microwave to Alliant Tower j Alliant Tower Microwave to DA E C Offsite Laboratory and Decontamination Ce n ter Emergency Operations Facility (1) Denotes a Ded i ca t ed Line Joint Public Info r mation C enter SECTION 'F' Rev. 29 Page 15 of 17 DAECEMERGENCYPLAN SECTION 'F' EMERGENCY COMMUNICATIONS Rev. 29 Page 16 of 17 FIGURE F-6 FEDERAL TELEPHONE SYSTEM (FTS-2001)

NRC EROS EOF Local TSC © ENS HPN @ RSCL © PMCL MCL DAEC EMERGENCY PLAN SECTION 'F' Rev. 29 EMERGENCY COMMUNICATIONS Page 17 of 17 FIGURE F-7 ALL-C ALL TELEPHONE SYSTEM LINN COUNTY Internet E OC Sheriff's Office 8.ckup F adlity MalnFadtlty '----------~,--~ - DAEC EMERGENCY PLAN SECTION 'E' Rev. 23 NOTIFICATION METHODS AND PROCEDURES Page 3 of 7 1.0 PURPOSE (1) This section describes 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 numbe r s 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 NRG Operations Center
• Emergency Coordinator
• Emergency Response and Recovery Director
• NRG Resident 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 verification 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
• 41 19.5 High Pressure Injection,
• HPCI/RCIC Auto Initiation PCIS Group 5 Isolation ,
• RWCU Isolation ARI ARI Initiation

& Recirc Pump A TWS Trip * +87 Two Feet Below During A TWS 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.

ACUREN November 17, 2012 RE: Radiation Incident, License No. 42-32443-01 Acuren USA BPXGPBU Prudhoe Bay, Alaska 99734-0122 Phone: (907) 659-8249 Fax: (907) 659-5936 Materials Engineering and Testing A Rockwood Company Reported to NRC by Telephone per 10 CFR, §30.50(b)(2)(ii), 1050hrs, November 17,2012, Report #48516 Equipment Problem: Source could not be fully retracted into shielded position. Cause of incident: The control cable ofa 35' control assembly broke approximately three (3) inches from the source connector. Equipment Involved: 7' flexible guide tube serial number GTl 211 Exposure device model: Sentinel Delta 8800, serial# D3503 Isotope : Ir-192, source serial # 87400B Source Activity : 60.7ci 35' Control Assembly s/n 11645 Place, Date, Time: Place: Drill Site 6 to Flow Station 3 Access Road, Eastern Operating Area, Prudhoe Bay, Alaska Date: November 16, 2012 Time: 1400 hrs AKT Actions taken to establish normal operations: The source assem~ly was gravity fed from the guide tube onto the ground and then shielded by reverse placement into an 880D exposure device, s/n D3926. A serviceable control assembly was then connected to exposure device D3503 and control cable routed through the device and guide fube, then connected to the source. The source was then retracted into device D3503 without incident. Corrective actions taken and planned to prevent reoccurrence: Remove 100% of control assemblies from service and complete a thorough inspection. Perform a safety stand down with all radiographers and assistants for review of daily equipment inspections as required in the Acuren O&E manual. Qualifications of personnel involved in incident: (1) Texas Industrial Radiographer Certification holder, dose received, 20mR (2) IRRSP card holder, dose received, 20mR (3) IRRSP card holder, Source Retrieval qualified, dose received, 48mR (4) IRRSP card holder, Source Retrieval qualified, dose received, 95rnR ~* Robert L. Jefferson Radiation Safety Officer, Alaska Acuren USA Inc. PO Box 340122 Prudhoe Bay, AK 99734 direct: 907-659-8249 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH ABNORMAL OPERATING PROCEDURE AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH Usage Level REFERENCE 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 RIVER WATER SUPPLY HIGH RIVER BED ELEVATION LOW RIVER WATER DEPTH NOTE Refer to EPIP 1.1 for EAL ASSESSMENT. I AOP 410 Page 1 of 24 PAGE PAGE PAGE 2 9 16 Rev. 2a j AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOSS OF RIVER WATER SUPPLY NOTE The portions of this procedure that minimize river make-up flow and maximizes make-up flow from other sources may be used as necessary in the event of an intrusion of excess foreign material from the Cedar River and/or Intake Structure without a Loss of River Water Supply. IMMEDIATE ACTIONS 1. None AUTOMATIC ACTIONS -CV-4914 and CV-4915 open and CV-4910A and CV-49108 close on low_ level in the ESW/RHRSW wet pits I AOP 410 Page 2 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOSS OF RIVER WATER SUPPLY FOLLOW-UP ACTIONS 1. Establish critical parameter monitoring of Circ Pit Level and ESW/RHRSW Pit Level, as pri.orities allow. 2. IF power is available AND annunciator 1 C06A (A-1 [21) "A"["B"] RWS PIT LO LEVEL is not on 3. IF standby pumps did not start 4. IF power is not available

5. IF no RWS pumps can be started ~ij (9 ~IANUAL SCRAIII 6. IF offsite power was lost AND River Water Supply pumps fail to start from the diesel THEN attempt to start standby pumps in both RWS Subsystems as necessary to restore needed makeup flow. THEN attempt to restart the tripped River Water Pumps. THEN attempt to restore power from 1 C08 AND attempt to restart pumps. THEN Reduce recirc flow to 39 Mlbm/hr in accordance with IPOI 4 AND manually scram the reactor. THEN Attempt to start pumps manually, that have their start permissive light on. AND restore RWS makeup to the stilling basin. 7. Update the Online Risk Monitor for the status of River Water Supply Pumps. 8. Maximize Well Water flow for makeup to the Circ Pit, while maintaining s 170 psig well water system pressure at the main plant BEECO Backflow Preventer.

9. Close the Circ Water Inlet to Slowdown Line valve M0-4253 to maintain Circ Water Pit inventory.

10. Secure Cooling Tower Fans as allowable

11. Secure one circ water pump and one cooling tower as soon as possible.

I AOP 410 Page 3 of 24 Rev. 2a j AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH . LOSS OF RIVER WATER SUPPLY FOLLOW-UP ACTIONS (continued)

12. Send an operator to the Intake Structure to verify alarms and check RWS pump breaker condition.

NOTE The RHRSW Pumps should be kept in operation to support use of the RHR System as directed by EOPs (Torus Cooling/Shutdown Cooling). The RHRSW Pumps may be secured when it is determined that ESW/RHRSW pit level cannot be maintained above 4 feet. Securing RHRSW Pumps prior to reaching 4 feet in the pits will preserve the ESW Supply to the SBDGs. 13. Minimize use of water from the RHRSW/ESW pit as follows: a. Secure RHRSW pumps unless required to support operation of the RHR System b. Shutdown any SBDG not required to ensure one Essential Bus is energized and/or required to ensure adequate core cooling. (1) Verify SBDG Cooling Valves CV-2080 and CV-2081 close when the respective SBDG is secured. c. Verify Well Water is available for cooling the operating Control Building Chiller and then secure ESW to the Control Building Chillers by unlocking and closing V-13-122 and V-13-125 on the Reactor Building 812' level. 14. Minimize heat addition to the Torus (Reliefs, HPCI, RCIC). 15. Use the Tu.rbine Bypass valves and/or Bypass Jack for Reactor pressure control and Reactor cooldown and continue to bleed steam to the Main Condenser for as long as possible.

16. At 1C15 and 1C17, place the HI COND BACKPRESS BYPASS switches in BYPASS. 17. Notify Security at 7254 prior to opening Pumphouse doors to arrange for required Security compensatory measures.

I AOP 410 Page 4 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOSS OF RIVER WATER SUPPLY FOLLOW-UP ACTIONS (continued)

18. Establish makeup to the ESW/RHRSW Wet Pits from the Fire System using 2-1/2" !3nd/or 5" hoses as follows: {C001} a. Using 2-1/2" hoses (N/A if not used): (1) Notify Mechanical Maintenance to remove the cover from outside hose head (octopus head). (2) Obtain 2-1/2" hoses from the warehouse and fire brigade trailer, an.d rig as many as possible (8 preferred) from the octopus head to the stilling basin. (3) Lash the hoses together with rope to prevent hose whip when the lines are charged. b. Using 5" hoses (N/A if not used): (1) Obtain 5" hoses from the 85b hose trailer. (2) Connect a 5" hose to any of the following fire hydrants as required:

• FH-1 located east of the Turbine Building
• FH-2 located southeast of the Turbine Building
• FH-7 located northeast of the Turbine Building (3) Rig 5" hoses as needed to the Stilling Basin. (4) Lash the hoses together with rope to prevent hose whip when the lines are charged. c. When directed by the CRS, start 1 P-48 or 1 P-49 and valve in hoses as necessary to maintain RHRSW/ESW pit level. d. Monitor RHRSW/ESW pit level at 1 C29, computer points 8279 and 8280, or on group display AOP 410. I AOP 410 Page 5 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOSS OF RIVER WATER SUPPLY FOLLOW-UP ACTIONS (continued)

19. Establish makeup to the RHRSW/ESW pits from GSW as follows: a. Verify at least one GSW pump is running. b. At 1 C452 (located in the B RHRSW/ESW pump room) open the following valves with the appropriate handswitch:

CV-8035A A RHRSW/ESW WET PIT CHLORINE HS-8035A INJECTION ISOLATION CV-8035B B RHRSW/ESW WET PIT CHLORINE HS-8035B INJECTION ISOLATION CV-8034 RHRSW/ESW DILUTION WATER HS-8034 SUPPLY VALVE c. In the CHLORINE BOOSTER PUMP ROOM, note and record the position of V-80-154, then fully open V-80-154 GSW Dilution Water Supply Balancing valve using a wrench from the tool board. NOTE NPSH requirement is 8 feet for the Gire Water Pumps. NPSH requirement is 4 feet for the ESW, RHRSW, and GSW pumps. 20. Monitor the Gire Water Pit level at Computer Point F092 and secure Circ Water Pumps if level cannot be maintained or restored greater than 8 feet and GSW pumps as necessary to prevent cavitation.

21. Monitor RHRSW/ESW Pit level and secure RHRSW and ESW pumps as necessary to prevent cavitation.

22. When the status of each River Water Supply pump is known and there is at least one operable pump in each RWS loop, select the operable pumps on HSS-2911A and B to be the pumps to auto restart on the diesel. 23. Comply with Technical Specifications for River Water Supply. 24. When River Water Supply pump operation is restored, return system to normal operation per 01 410. 25. Update the Online Risk Monitor for the status of River Water Supply Pumps. I AOP 410 Page 6 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOSS OF RIVER WATER SUPPLY FOLLOW-UP ACTIONS (continued)

26. Open and Lock open the following valves:

• V-13-122 ESW Loop A Return Header Isolation
• V-13-125 ESW Loop B Return Header Isolation

27. Independently verify the following valves are locked open:

• V-13-122 ESW Loop A Return Header Isolation IV
• V-13-125 ESW Loop B Return Header Isolation IV I AOP 410 Page 7 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOSS OF RIVER WATER SUPPLY PROBABLE ANNUNCIATORS 1 C06A, A-1 "A" RWS PIT LO LEVEL A-2 "B" RWS PIT LO LEVEL A-3 "B" RWS PUMP 1P-117B TRIP A-4 "D" RWS PUMP 1 P-11 ?D TRIP B-1 "A" RWS PUMP 1P-117A TRIP B-2 "C" RWS PUMP 1P-117C TRIP B-5 "A" COOLING TOWER BASIN HI/LO LEVEL B-6 "B" COOLING TOWER BASIN HI/LO LEVEL D-1 "A" RHRSW/ESW PIT LO LEVEL D-2 "B" RHRSW/ESW PIT LO LEVEL
• D-11 CIRC WATER PIT LO LEVEL PROBABLE INDICATIONS 1C06 -River Water makeup flow stopped at FR-4916 and FR-4917 or Fl-4916 and Fl-4917 I AOP 410 Page 8 of 24 Rev. 2a j AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH --1. None AUTOMATIC ACTIONS -None I AOP 410 HIGH RIVER BED ELEVATION IMMEDIATE ACTIONS Page 9 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY /HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH HIGH RIVER BED ELEVATION FOLLOW-UP ACTIONS NOTE River water elevation is measured via LR-2901, Computer Point M01 O.V, and/or STP 3.0.0-01.

River water depth is determined utilizing 1 H239 River Depth Measuring Crane per 01 41 O or during surveillance testing. River bed elevation is determined by subtracting the 1 H239 measured river water depth from river water elevation. River bedrock elevation ranges from 722' to 723'. The difference between river bed elevation and river bedrock elevation is considered the sand bed height. River bed elevation and sand bed elevation are synonymous with each other. Current sand gate elevation is listed in STP 3.0.0-01. Actions taken in this section are to be coordinated with the requirements of the Low River Water Depth section of this AOP. 1. IF River bed elevation is determined to be greater than (>) 726' utilizing 1 H239. THEN Coordinate with the Work Week Manager, System Engineering, and Maintenance to perform the following as applicable.

a. Task and schedule Iowa Vane Field Mapping (Model WO 40105684).

b. IF following completion of Iowa Vane Field Mapping it is determined river bed elevation is less than ( <) 727' within 1 O' of the intake, THEN continue performing Iowa Vane Field Mapping every two weeks. c. IF following completion of Iowa Vane Field Mapping it is determined river bed elevation is greater than or equal to (;::) 727' within 1 O' of the intake, THEN perform Iowa Vane Field Mapping weekly. d. IF following completion of Iowa Vane Field Mapping it is determined river bed elevation is greater than or equal to (;::) 728' within 1 O' of the intake, I AOP 410 OR IF River bed elevation is determined to be greater than (>) 727' utilizing 1H239, THEN perform the following:

(1) Determine river water depth and river bed elevation daily utilizing 1 H239 per 01 410, River Water Depth Section. Log daily river water depth and river bed elevation results in the shift log. (2) Task and schedule River Channel Mapping (Model WO 1374301). (3) OSM and System Engineering shall evaluate raising the sand gate elevation to minimize sand ingestion based on current conditions and trend. Sand gates may be raised per Step 3. Page 10 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH HIGH RIVER BED ELEVATION FOLLOW-UP ACTIONS (continued)

2. IF River bed elevation is determined utilizing 1 H239 to be less than or equal to THEN work with the Work Week Manager, System Engineer, and Maintenance to perform the following as applicable.

(::::) 0.5 feet from the current sand gate elevation.

a. IF River Bed Elevation is less than (<) 0.5 feet below the top of the Sand Gate, THEN update the Online Risk Monitor by using the Review/Change Environmental Variables button to adjust Ultimate heat Sink to High Risk. b. Determine river water depth and river bed elevation daily utilizing 1 H239 per 01 410, River Water Depth Section. Log daily river water clepth and river bed elevation results in the shift log. -c. Task and schedule the following:

(1) Iowa Vane Excavation (Model WO 40158555):

• Schedule excavation to begin prior to river bed elevation reaching the top elevation of the sand gates.
• Excavate Iowa Vane field and river channel downstream of intake structure for a minimum of 100 yards.
• Excavate Iowa Vane field to river bottom (722' to 723') per mapping.
• Monitor screen differential pressure during excavation.

NOTE Dive Work Orders are to be performed following completion of excavation. Inspection/cleaning includes Forebay up to the river side of Lite Locs. Removal of Lite Locs and inspection/cleaning to Traveling Screen and RWS pump pits to be completed based on Forebay results and OSM direction. I AOP 410 (2) Diving inspection and cleaning work orders: * 'A' side Intake structure forebay (Model WO 1374681) or 'B' side Intake structure forebay (Model WO 1375828) as applicable.

• 'A' side RWS pump pit (Model WO 1374679) and 'B' side RWS pump pit (Model WO 1374680) per OSM direction.

(3) Following excavation and diving, determine river water depth and river bed elevation using 1 H239 per 01 410 River Water Depth section. Log river watE:r depth and river bed elevation results in the shift log. (4) Following excavation and diving, return to Step(s) 1 and/or 2 and take actions as applicable. Page 11 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH HIGH RIVER BED ELEVATION FOLLOW-UP ACTIONS (continued) NOTE Base sand gate elevation is 728.0'. Fully raised sand gate elevation is 730'. Fully lowered elevation is 723.5'. Both sand gates are to be maintained at the same elevation. Raising sand gates will impact water level at the intake structure. Divers must inspect and clean (if necessary) the intake structure forebay in the vicinity of the Sand Gates to verify the gates are clear prior to lowering the gates back to 'base' elevation. Sand Gate electrical controls are non-functional. Sand Gates must be raised and lowered manually using manual ratchet tools. Sand loading on the river side of the Sand Gates may impede or prevent gate movement. The OSM and System Engineering shall consider the requirements of SR 3.7.2.3 and SR 3.7.2.5 for maintaining greater than 1' water level above the sand gate elevation when* proposing to raise the sand gates. 3. IF desired to raise sand gate elevation, THEN coordinate with the Work Week Manager, System Engineering, and Maintenance to perform the following as applicable.

a. OSM consult with System Engineering to determine the newly proposed sand gate elevation.

I AOP 410 (1) IF it is expected that river water level will remain greater than or equal (~) to 2.5' above the newly proposed sand gate elevation, THEN the sand gates may be raised. (2) IF it is expected that river water level will not remain greater than or equal (~) to 2.5' above the newly proposed sand gate elevation, THEN the sand gates may be raised only after the OSM verifies that the requirements of SR 3.7.2.3 and SR 3.7.2.5 will remain met. (3) IF sand gates are to be raised, THEN perform the remaining steps, otherwise N/A the remaining Step 3 substeps. Page 12 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH HIGH RIVER BED ELEVATION FOLLOW-UP ACTIONS (continued) NOTE The area beneath the sand gates should be inspected and cleaned as necessary to remove sand accumulation that may prevent sand gate movement prior to attempting to lower the sand ates. CAUTION. *----~------~----'---II If sand gates are raised, sand may build up underneath the gates preventing the gates from being lowered. This could result in a loss of river water supply if river level drops below the top of the gates. To preclude this, the sand gates should be kept at least 1 foot below river water level as monitored in STP 3.0.0-01. {C002} b. IF not already completed, THEN task and schedule Model WO 1374681 OR Model WO 1375828 for diver inspection/cleaning of the Intake Structure forebay. c. Manually raise gates to the proposed elevation per 01410 (maximum elevation is 730.0'). d. Verify final sand gate elevation using gate dial indicator, gate height measuring device, or divers. e. Determine river water depth and river bed elevation utilizing 1 H239 per 01410, River Water Depth Section. Log river water depth and river bed elevation results in the shift log. f. Revise STP 3.0.0-01, Instrument Checks, to document final current sand gate elevation.

g. Continue to perform applicable actions per Step(s) 1 and/or 2 until the conditions of 3.i are met. h. WHEN River Bed Elevation is more than (>) 0.5 feet below the top of the Sand Gate, THEN update the Online Risk Monitor by using the Review/Change Environmental Variables button to adjust Ultimate heat Sink to Normal Risk. I AOP 410 Page 13 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH HIGH RIVER BED ELEVATION FOLLOW-UP ACTIONS (continued)

i. IF river bed elevation is determined to be less than (<) 726' utilizing 1 H239. I AOP 410 IF river bed elevation is determined utilizing 1 H239 to be greater than (>) 0.5' below current sand gate elevation.

THEN the sand gates may be returned to 728' by performing the following (1) Verify all excavation, inspection and or cleaning as required by Step 2 and or Step 3.b is completed as necessary. (2) Manually lower sand gates to base elevation of 728' or as determined by System Engineering per 01 410. (3) Verify final sand gate elevation using gate dial indicator, gate height measuring device, or divers. (4) Determine river water depth and river bed elevation utilizing 1 H239 per 01 410, River Water Depth Section. Log river water depth and river bed elevation results in the shift log. (5) Revise STP 3.0.0-01, Instrument Checks, to document final current sand gate elevation. Page 14 of 24 Rev. 28 j AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH HIGH RIVER BED ELEVATION PROBABLE ANNUNCIATORS None PROBABLE INDICATIONS River bed elevation > 726 feet River bed elevation is determined to be less than or equal to (:5) 0.5 feet from the current sand gate elevation I AOP 410 Page 15 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY /HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH 1. None AUTOMATIC ACTIONS -None I AOP 410 LOW RIVER WATER DEPTH IMMEDiATE ACTIONS Page 16 of 24 Rev. 2a j AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOW RIVER WATER DEPTH FOLLOW-UP ACTIONS NOTE River water elevation is measured via LR-2901, Computer Point M01 O.V, and/or STP 3.0.0-01. River water depth is determined utilizing 1 H239 River Depth Measuring Crane per 01410 or during surveillance testing. River bed elevation is determined by subtracting the 1 H239 measured river water depth from river water elevation. River bedrock elevation ranges from 722' to 723'. The difference between river bed elevation and river bedrock elevation is considered the sand bed height. River bed elevation and sand bed elevation are synonymous with each other. Current sand gate elevation is listed in STP 3.0.0-01. Actions taken in this section are to be coordinated with the requirements of the High River Bed Elevation section of this AOP. 1. IF River Water Depth is less than THEN (<) 1.5' above the Sand Gate elevation, update the Online Risk Monitor by using the Review/Change Environmental Variables button to adjust Ultimate heat Sink to High Risk. 2. IF river water depth is THEN comply with Technical Specification determined to be less than or SR 3.7.2.3 and SR 3.7.2.5. equal to (S) 2' utilizing 1 H239, 3. IF river water depth is THEN perform the following determined to be less(<) 2.5'

• utilizing 1 H239, river water elevation is determined to be less than (<) 2.5' above the current sand gate elevation

a. Determine river water depth and river bed elevation daily using 1 H239 per 01 410 River Water Depth section. Log river water depth and river bed elevation results in the shift log. b. IF river bed elevation is less than or equal to (S) 725' elevation, THEN the OSM and System Engineering shall evaluate lowering the sand gate elevation to maintain river water elevation greater than or equal to (.::) 2.5' above the sand gate elevation.

Sand gates may be lowered per Step 4. I AOP 410 Page 17 of 24 Rev. 2a j AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOW RIVER WATER DEPTH FOLLOW-UP ACTIONS (continued)

c. IF river bed elevation is greater than(>) 725' elevation, THEN task and schedule the following prior to lowering the sand gates: (1) Iowa Vane Excavation (Model WO 40158555)

• Schedule excavation to begin prior to river depth less than (<) 2.5' OR river level less than (<) 2.5' from the top of the gate.
• Excavate Iowa Vane field and river channel downstream of intake structure for a minimum of 100 yards.
• Excavate Iowa Vane field to river bottom (722' to 723') per mapping. * . Monitor screen differential pressure during excavation.

NOTE Dive Work Orders are to be performed following completion of excavation. Inspection/cleaning includes the Forebay up to the river side of Lite Locs. Removal of Lite Locs and inspection/cleaning to Traveling Screen and RWS pump pits to be completed based on Forebay results and OSM direction. I AOP 410 (2) Diving inspection and cleaning work orders: * 'A' side Intake structure forebay (Model WO 1374681) or 'B' side Intake structure forebay (Model WO 1375828) as applicable.

• 'A' side RWS pump pit (Model WO 1374679) and 'B' side RWS pump pit (Model WO 1374680) per OSM direction.

(3) Following excavation and diving, determine river water depth and river bed elevation using 1 H239 per 01 410 River Water Depth section. Log river water depth and river bed elevation results in the shift log. (4) Following excavation and diving, return to Step(s) 2 and/or 3 and take actions as applicable. Page 18 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOW RIVER WATER DEPTH FOLLOW-UP ACTIONS (continued) NOTE Base sand gate elevation is 728.0'. Fully raised sand gate elevation is 730'. Fully lowered elevation is 723.5'. Both sand gates are to be maintained at the same elevation. Raising sand gates will impact water level at the intake structure. Divers must inspect and clean (if necessary) the intake structure forebay in the vicinity of the Sand Gates to verify the gates are clear prior to lowering the gates back to 'base' elevation. Sand Gate electrical controls are non-functional. Sand Gates must be raised and lowered manually using manual ratchet tools. Sand loading on the river side of the Sand Gates may impede or prevent gate movement. The OSM and System Engineering shall consider the requirements of SR 3.7.2.3 and SR 3.7.2.5 for maintaining greater than 1' water level above the sand gate elevation when proposing to lower the sand gates. 4. IF desired to lower sand gate elevation per Step 3.b, THEN coordinate with the Work Week Manager, System Engineering, and Maintenance to perform the following as applicable: NOTE The area beneath the sand gates should be inspected and cleaned as necessary to remove sand accumulation that may prevent sand gate movement prior to attempting to lower the sand ates. If sand gates are raised, sand may build up underneath the gates preventing the gates from being lowered. This could result in a loss of river water supply if river level drops below the top of the gates. To preclude this, the sand gates should be kept at least 1 foot below river water level as monitored in STP 3.0.0-01. {C002} a. IF not already completed, THEN task and schedule Model WO 1374681 OR Model WO 1375828 for diver inspection/cleaning of the Intake Structure forebay. b. Manually lower gates to desired elevation per 01410 maintaining greater than or equal to (;::) 1' above river bed elevation (minimum sand gate elevation is 723.5'). I AOP 410 Page 19 of 24 Rev. 28 j AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOW RIVER WATER DEPTH FOLLOW-UP ACTIONS (continued)

c. Verify sand gate elevation using gate dial indicator, gate height measuring device, or divers. d. Determine river water depth and river bed elevation utilizing 1 H239 per 01 410, River Water Depth Section. Log river water depth and river bed elevation results in the shift log. e. Revise STP 3.0.0-01, Instrument Checks, to document final current sand gate elevation.

f. Continue to perform applicable actions per Step(s) 2 and/or 3 until the conditions of 4.g are met. g. IF river water depth is determined to be greater than or equal to (.::) 2.5' utilizing 1 H239, IF river water elevation is determined to be greater than-or equal to (.::) 732.5' to ensure greater than 2.5' is maintained above the current sand gate elevation THEN the sand gates may be returned to 728' by performing the following:

(1) Verify all excavation, inspection and or cleaning as required by Step 3 and or Step 4.a is completed as necessary. (2) Manually raise sand gates per 01410 to base elevation of 728' or as determined by System Engineering. (3) Verify final sand gate elevation using gate dial indicator, gate height measuring device, or divers. (4) Determine river water depth and river bed elevation utilizing 1 H239 per 01 410, River Water Depth Section. Log river water depth and river bed elevation results in the shift log. (5) Revise STP 3.0.0-01, Instrument Checks, to document final current sand gate elevation.

5. WHEN River Water Depth is more than(>) 1.5' above the Sand Gate elevation, THEN update the Online Risk Monitor by using the Review/Change Environmental Variables button to adjust Ultimate heat Sink to Normal Risk. I AOP 410 Page 20 of 24 Rev. 2a j L__ _____ _

AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOW RIVER WATER DEPTH PROBABLE ANNUNCIATORS None PROBABLE INDICATIONS River water depth 2.5 feet I AOP 410 Page 21 of 24 Rev. 2a j AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH APPENDIX 1 INFORMATION Below is a table of approximate times before the ESW/RHRSW Wet Pits and Stilling Basin empty on a complete loss of River Water supply. Times are based on an initial pit level of 28 feet (Reference CAL-M93-078). Pumps 1 RHRSW 1 RHRSW 2 RHRSW Running 1 ESW 2ESW 1 ESW 2ESW 2ESW GPM Flow 1200 2400 3600 4800 7200 Minutes 92 46 31 23 15 River Water Supply Pumps can be started/restarted when the pump's respective white "START PERMISSIVE" indicating light is on. This light will be on when the following conditions are met: Control Power is available No existing undervoltage condition on respective Essential Bus Pump restart timer has timed out (approximately 2 minutes) Pump control switch is positioned to either AUTO or START I AOP 410 Page 22 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH References

6. Technical Specifications

7. 01410, River Water Supply System 8. 01 408, Well Water System 9. 01 411, General Service Water System 10. 01416, RHR Service Water System 11. 01 442, Circ Water System 12. 01454, Emergency Service Water System 13. P&ID M129, River Water Supply System Intake Structure

14. P&ID M146, Service Water System Pumphouse

15. Service Water Systems, Bechtel Drawing No. 7884-E-111

<13, 13A> 16. 4160V and 480V System Control and Protection, Bechtel Drawing No. 7884-E-104<25, 26> 17. Bechtel Drawing No. 7884-APED-B21 (3A) 18. Bechtel Drawing No. 7884-APED-B21-18(3A)NI

19. P&ID M15, Equipment Location Pumphouse Plans and Elevations

20. Architectural No. A-78, Pumphouse Plans and Elevations

21. Civil No. C-684, Pumphouse Cone. Floor Plans at 761'-0" and 747'-6" 22. DCP 1496, River Water Pump Restart Logic Mod 23. CAL-M93-078, RHRSW/ESW Pit Pumpdown Times 24. AR 95-1478, AR 95-2070-15, AR 18569, AR 23595 25. NG-94-4671, NG-95-2864 Commitment Items 26. C001 -INPO SOER 07-2 (recommendation 5a) regarding

'Intake Cooling Water Blockage'

27. C002 -Licensing CTS No. 199005110204, NRC Inspection Report 90003 I AOP 410 Page 23 of 24 Rev. 281 ABNORMAL OPERATING PROCEDURE AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH Usage Level Reference Use Record the following:

DatefTime: ______ / ____ 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 Da~: -------------------

---

Print Signature CROSS~DISCIPLINE REVIEW (AS REQUIRED) " / Reviewed By: I Date: Print Signature Reviewed By: I Date: Print Signature PROCEDURE APPROVAL " Approved By _________

! ___________ Date: -------.1 Print Signature I AOP 410 Page 24 of 24 Rev. 281 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, Meth.odo/ogy 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 NRG 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 NRG 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)? ~1 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) is 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 equipment, 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 reduction 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 within 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 response 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 ICs, 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 reliability of the second train is a concern. In addition, escalation to the Alert classification should not occur if the damage from the hazardous event is limited to a SAFETY SYSTEM that was inoperable, or out of service, prior to the event occurring. As such, the proposed 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 Fission Product Barrier and/or Abnormal Radiation Levels/Radiological Effluent recognition categories would provide an escalation path to a Site Area Emergency or a General Emergency. The 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 NRG-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 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.

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 NRG-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) a. The occurrence of ANY of the following hazardous events: Basis:

• 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 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 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/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. Operators will make a determination of VISIBLE DAMAGE based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to 'cause concern regarding the operability or reliability of the SAFETY SYSTEM train. Escalation of the emergency classification level would be via IC 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 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) a. The occurrence of ANY of the following hazardous events: Basis:

• 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/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 .. 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 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 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 inc l uding LPCI flow will cause the Fuel Zone instruments to read high.

I i I H-2 <6% 6% or unknown CD IF ........ offsite release rate is expected to stay below normal limits (Detail D), THEN ... vent and purge the primary containment: H-3 ..-OK to defeat isolations except high radiation (Defeat 9). ..-If pneumatic supplies are unavailable, use SAMP 706 , Venting the Primary Containment Following Loss of Pneumatic Supply. 1. Vent as follows:

• IF ........ torus water level is below 16 ft , THEN ... vent the drywell through the torus (SEP 301.1 ).
• IF ........ torus water level is at or above 16 ft , OR ...... the torus cannot be vented, THEN ... vent directly from the drywell (SEP 301.2). 2. IF ........ the primary containment can be vented , THEN ... purge the drywell with nitrogen using N 2 purge (SEP 303.2). 3. Stop the vent and purge (if not required by other SAG steps) when:
• Hydrogen is no longer detected in the drywell , OR
• Offsite release rate reaches normal limits (Detail D). IF ........ offsite release rate is expected to stay below General Emergency Levels (EAL RG1 ), OR ...... RPV water level cannot be maintained above +15 in. (TAF), THEN ... vent and purge the primary containment:

..-OK to defeat .§.ll isolations (Defeat 10). ..-If pneumatic supplies are unavailable , use SAMP 706, Venting the Primary Containment Following Loss of Pneumatic Supply. 1. Vent as follows:

• IF ........ torus water level is below 16 ft , THEN ... vent the drywell through the torus (SEP 301.1 ).
• IF ........ torus water level is at or above 16 ft , OR ...... the torus cannot be vented , THEN ... vent directly from the drywell (SEP 301.2). 2. IF ........ the primary containment can be vented , THEN ... purge the drywell with nitrogen at max flow using N 2 purge (SEP 303.2). 3. Stop the vent and purge (if not required by other SAG steps) when:
• Hydrogen is no longer detected in either the drywell or the torus, OR UI :I ... 0 I-H-6 <6% 6% or unknown IF ........ offsite release r: (Detail D), THEN ... vent and purge H-7 .... OK to defE ..-If pneuma Venting th Pneumatic

1. Vent as folio*

• IF ........ 1 THEN ... ,
• IF ........ 1 AND .... 1 THEN ... , 2. IF ........ the~ THEN ... purg (SEF 3. Stop the ven steps) when:
• Hydroge OR
• Offsite re IF ........ offsite release r: Emergency Lev OR. ..... RPV water leve THEN ... vent and purge ..-OK to defE ..-If pneuma Venting th Pneumatic

1. Vent as folio*

• IF ........ 1 THEN ... ,
• IF ........ 1 AND .... ! THEN ... , 2. IF ........ the~ THEN ... purg usin! 3. Stop the ven steps) when:
• Hydroge the torus OR

,J ~----10 )umps to =r level lperating Q) ... ::::, -cu ... Q) a. E Q) I-f"" HPCI Room Area HPCI EMER COOLER AMBIENT HPCI ROOM AMBIENT HPCI ROOM DIFFERENTIAL RCIC Room Area RCIC EMER COOLER AMBIENT RCIC ROOM AMBIENT RCIC ROOM DIFFERENTIAL Torus Area TORUS CATWALK NORTH AMBIENT TORUS CATWALK WEST AMBIENT TORUS CATWALK SOUTH AMBIENT TORUS CATWALK EAST AMBIENT TORUS CATWALK EAST DIFF TORUS CATWALK WEST DIFF TORUS CATWALK SOUTHWEST DIFF TORUS CATWALK SOUTH DIFF RB 786' South Area RWCU PUMP ROOM AMBIENT RWCU HX ROOM AMBIENT RB 757' South Area RWCU ABOVE TIP ROOM AMBIENT Steam Tunnel Area STEAM TUNNEL AMBIENT STEAM TUNNEL DIFFERENTIAL Area/Location RB 757' South Area RB RAILROAD ACCESS AREA SOUTH CRD MODULE AREA TIP ROOM RB 757' North Area NORTH CRD MODULE CRD REPAIR ROOM RB 786' North Area ---*. .. TR/TOR 2225A[B] Ch 1 175 310 TR/TOR 2225A Ch 2 175 310 TR/TOR 2225A[B] Ch 4[3] 50 N/A TR/TOR 2425A[B] Ch 1 175 300 TR/TOR 2425A Ch 2 175 300 TR/TOR 2425A[B] Ch 4 50 N/A TR/TDR 2425A Ch 3 150 165 TR/TOR 2425B Ch 2 150 165 TR/TDR 2225A Ch 3 150 165 TR/TOR 2225B Ch 2 150 165 TR/TDR 2425A Ch 5 50 N/A TR/TOR 2425B Ch 5 50 N/A TR/TOR 2225A Ch 5 50 N/A TR/TDR 2225B Ch 4 50 N/A TR/TDR 2700A[B] Ch 1 130 212 TR/TDR 2700A[B] Ch 2 , 3 130 212 TR/TOR 2700A[B] Ch 4 , 5 111.5 150 TR/TOR 2425B Ch 3 160 300 TR/TOR 2225B Ch 5 70 NIA Indicator mR/hr mR/hr R19167 10 100 Rl9169 10 100 Rl9176 60 600 Rl9168 10 100 Rl9170 15 150 - 1.2.7 HSM Dose Rates with a Loaded 24P , 528 or 61 BT DSC Limit/Specification: Applicability

Objective:

Action: 6IBT 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 , 528 or 61 BT 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 site exposures during storage. a. If specified dose rates are exceeded , the following actions should be taken: 1. Ensure that the DSC i s 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 BASES-DAEC EOP BASES DOCUMENT BREA K POINTS R ev. 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 Coo l ing 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 A TWS 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.

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 J ::=J-----10 )umps to =r level lperating G) ... ::::, ... n:s ... G) C. E G) I-f"" HPCI Room Area HPCI EMER COOLER AMBIENT HPCI ROOM AMBIENT HPCI ROOM DIFFERENTIAL RCIC Room Area RCIC EMER COOLER AMBIENT RCIC ROOM AMBIENT RCIC ROOM DIFFERENTIAL Torus Area TORUS CATWALK NORTH AMBIENT TORUS CATWALK WEST AMBIENT TORUS CATWALK SOUTH AMBIENT TORUS CATWALK EAST AMBIENT TORUS CATWALK EAST DIFF TORUS CATWALK WEST DIFF TORUS CATWALK SOUTHWEST DIFF TORUS CATWALK SOUTH DIFF RB 786' South Area RWCU PUMP ROOM AMBIENT RWCU HX ROOM AMBIENT RB 757' South Area RWCU ABOVE TIP ROOM AMBIENT Steam Tunnel Area STEAM TUNNEL AMBIENT STEAM TUNNEL DIFFERENTIAL A r ea/Location RB 757' South Area RB RAILROAD ACCESS AREA SOUTH CRD MODULE AREA TIP ROOM RB 757' North Area NORTH CRD MODULE CRD REPAIR ROOM RB 786' North Area ----* -.. TR/TOR 2225A[BJ Ch 1 175 310 TR/TOR 2225A Ch 2 175 310 TR/TOR 2225A[BJ Ch 4[3] 50 N/A TR/TOR 2425A[B] Ch 1 175 300 TR/TOR 2425A Ch 2 175 300 TR/TOR 2425A[BJ Ch 4 50 N/A TR/TOR 2425A Ch 3 150 165 TR/TOR 2425B Ch 2 150 165 TR/TOR 2225A Ch 3 150 165 TR/TOR 2225B Ch 2 150 165 TR/TOR 2425A Ch 5 50 N/A TR/TOR 2425B Ch 5 50 N/A TR/TOR 2225A Ch 5 50 N/A TR/TOR 2225B Ch 4 50 N/A TR/TDR 2700A[B] Ch 1 130 212 TR/TOR 2700A[B] Ch 2 , 3 130 212 TR/TOR 2700A[B] Ch 4 , 5 111.5 150 TR/TDR 2425B Ch 3 160 300 TR/TOR 2225B Ch 5 70 N/A Ind i cator mR/hr mR/hr RI 9167 10 100 RI 9169 10 100 RI 9176 60 600 RI 9168 10 100 RI 9170 15 150 - 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, containmen 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 DWff-3 D Maximize drywell cooling. IF ..-If necessary , bypass drywell cooling isolation and fan speed interlock (Defeat 4). THEN Drywell pressure drops below 2.0 psig Verify containment sprays isolate. DWff-4 drywell temperature reaches 280°F owrr-s D D IF torus water level is below 13.5 ft , AND drywell temperature is below the Drywell Spray Initiation Limit (Graph 7), THEN 1. Shut down recirc pumps. 2. In i tiate drywell sprays using only pumps not required for adequate core cooling (Table 9). DW/T-6 drywell temperature cannot be restored and maintained below 340°F LJ IF Torus pressure drops below 2.0 psif Drywell pressure drops below 2.0 ps Primary containment pressure reduc required to:

• Restore and maintain adequate c OR
• Reduce the total offsite radiation PC/P-3 D PC/P-4 D Initiate to1 required f ..-OK con leve belc PC/P-5 D *~*

)umps to =r level 1perating ,J C) L. :l ... C'G L. C) C. E C) I-f"" HPCI Room Area HPCI EMER COOLER AMBIENT HPCI ROOM AMBIENT HPCI ROOM DIFFERENTIAL RCIC Room Area RCIC EMER COOLER AMBIENT RCIC ROOM AMBIENT RCIC ROOM DIFFERENTIAL Torus Area TORUS CATWALK NORTH AMBIENT TORUS CATWALK WEST AMBIENT TORUS CATWALK SOUTH AMBIENT TORUS CATWALK EAST AMBIENT TORUS CATWALK EAST DIFF TORUS CATWALK WEST DIFF TORUS CATWALK SOUTHWEST DIFF TORUS CATWALK SOUTH DIFF RB 786' South Area RWCU PUMP ROOM AMBIENT RWCU HX ROOM AMBIENT RB 757' South Area RWCU ABOVE TIP ROOM AMBIENT Steam Tunnel Area STEAM TUNNEL AMBIENT STEAM TUNNEL DIFFERENTIAL Area/Location RB 757' South Area RB RAILROAD ACCESS AREA SOUTH CRD MODULE AREA TIP ROOM RB 757' North Area NORTH CRD MODULE CRD REPAIR ROOM RB 786' North Area . ---* -.. TR/TOR 2225A[B] Ch 1 175 310 TR/TOR 2225A Ch 2 175 310 TR/TOR 2225A[B] Ch 4[3] 50 N/A TR/TOR 2425A[B] Ch 1 175 300 TR/TOR 2425A Ch 2 175 300 TR/TOR 2425A[B] Ch 4 50 N/A TR/TOR 2425A Ch 3 150 165 TR/TOR 24258 Ch 2 150 165 TR/TOR 2225A Ch 3 150 165 TR/TOR 22258 Ch 2 150 165 TR/TOR 2425A Ch 5 50 N/A TR/TOR 24258 Ch 5 50 N/A TR/TOR 2225A Ch 5 50 N/A TR/TOR 22258 Ch 4 50 N/A TR/TOR 2700A[B] Ch 1 130 212 TR/TOR 2700A[BJ Ch 2 , 3 130 212 TR/TOR 2700A[BJ Ch 4 , 5 111.5 150 TR/TOR 24258 Ch 3 160 300 TR/TOR 22258 Ch 5 70 N/A Indicator mR/hr mR/hr Rl9167 10 100 Rl9169 10 100 Rl9176 60 600 Rl9 1 68 10 100 Rl9170 15 150 - I AOP 901 PROBABLE ANNUNCIATORS None PROBABLE INDICATIONS 1C35 EARTHQUAKE -The amber DESIGN BASIS EARTHQUAKE (DBE) light is ON. -The amber OPERATING BASIS EARTHQUAKE (OBE) light is ON. -The amber .01 G RECORDERS RUNNING light is ON. -The white CONTINUITY light is OFF. -The Seismic Wailing Alarm is sounding. -Building vibration. A Cooling Tower Valve House -No power i ndicating light 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 I AOP 902 FLOOD FOLLOW: UP ACTIONS (continued)

NOTE River Water Level is required to be r ecorded hourly and verified to be < 757 feet whenever river level is >753 feet, IAW TLCO 3.7.1. 9. IF rive r water level THEN Commence r ecording river water level hourly. reaches 753' 10. Pre-stage equipment in the location per Attachment 3 for the Pump House , Turbine Building , Control Building, Reactor and Recombiner Building , and Radwaste/LLRPSF CAUTION --------------

---

u D e energi z ing MCC 189106 [182106] will cause RWS Screen Wash Pump 1 P-112A[B] to be inoperable. Refer to applicable Tech. Spec. Sections. NOTE Once level is greater than 754', the only method to get into the Intake Structure w i thout walking through water is to remove the louver on the Intake Air Supply and use a boat to get to the building. 11. Prior to river level reach i ng 754', open the following breakers to prevent energizing electrical equipment in the Intake Structure that may become submerged. Other loads at the Intake are on the 2nd floor and will be deenergized later at higher river levels. I AOP 902 Breaker Load 1 89106 Screen Wash Pump 1 P-112A 189107 189108 189109 189111 189112 182106 182107 182108 182109 182110 182111 182112 182114 Travel i ng Screen Drive 1 F-36A Screen Wash Control Panel 1C-154A Radial Sand and Side Gate Hoist 1 H-26A Intake Structure 480VAC Power Receptacles Screen Wash Nozzle Shutoff M0-2902 Screen Wash Pump 1 P-1128 Traveling Screen Drive 1 F-368 Screen Wash Control Panel 1 C-1548 Intake Structure Trash Rake 1 S-83 Intake Structure 480VAC Power Receptacles Radial Sand and Side Gate Hoist 1 H-268 Screen Wash Nozzle Shutoff M0-2903 Instrument Enclosure 1C412 Page 10 of 49 Rev. 541 I AOP 902 FLOOD FOLLOW-UP ACTIONS (continued)

12. When river level reaches 756', at the Pump House , Turbine Building , Control Building , Reactor and Recombiner Building , and Radwaste/LLRPSF implement action per Attachment 3 to pump water/monitor levels that may leak into these buildings in the future if river level continues to rise. 13. IF it is determined that the THEN a. flood level will reach Shut down the plant to Cold Shutdown 757' OR ----does reach 757' u b. Deenergize equipment no t r equired to shutdown the plant or for personal safety or for security or fuel pool cooling, at the following: (1) Cooling towers (2) Administration Bldg. (3) Pump House (4) Low Level Radwaste Process Storage Fac i lity (LLRPSF) Building (5) Machine Shop (6) Offgas Retention Facility (7) Security Bldg (8) TSC Bldg (9) Data Acquisition Center (10) Badging Center (11) Electric Shop (12) Construction support Center (13) West/East Warehouse (14) Fabrication Shop (15) Sewage Treatment Plant (16) Shooting Range (17) Barn c. Refer to EPIP 1.1 for EAL assessment.

CAUTION Major 4160V loads should be started at approximately 10 second intervals to avoid overloading the diesel generator. Diesel generator load should be monitored at 1 C08 as bus loads are added. I AOP 902 d. Start both SBDGs and transfer the 4160VAC Essential Susses 1 A3 and 1 A4 to the SBDGs per 01 304.2 , Section 7.6. Page 11 of 49 Rev. 541

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AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH ABNORMAL OPERATING PROCEDURE AOP 410 LOSS OF RIVE R W A TER SUPPLY/HIGH RIVER BED E L EV A T ION/L O W RIVER WATER DEPTH Usage Level REFERENCE NOTE This AOP is normally coordinated by the Reactor Operator.

Record the following: Date/Time: ______ / ____ 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 RIVER WATER SUPPLY HIGH RIVER BED ELEVATION LOW RIVER WATER DEPTH NOTE Refer to EPIP 1.1 for EAL ASSESSMENT.

I AOP 410 Page 1 of 24 PAGE PAGE PAGE 2 9 16 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOSS OF RIVER WATER SUPPLY NOTE The portions of this procedure that minimize river make-up flow and maximizes make-up flow from other sources may be used as necessary in the event of an intrusion of excess foreign material from the Cedar River and/or Intake Structure without a Loss of River Water Supply. IMMEDIATE ACTIONS 1. None AUTOMATIC ACTIONS -CV-4914 and CV-49 15 open and CV-4910A and CV-49108 close on low level in the ESW/RHRSW wet pits I AOP 410 Page 2 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOSS OF RIVER WATER SUPPLY FOLLOW-UP ACTIONS 1. Establish critical parameter monitoring of Circ Pit Level and ESW/RHRSW Pit Level , as priorities allow. 2. IF power is available AND annunciator 1 C06A (A-1 [2]) " A"[" B"] RWS PIT LO LEVEL is not on THEN attempt to start standby pumps in both RWS Subsystems as necessary to restore needed makeup flow. 3. IF standby pumps did not start THEN attempt to restart the tripped River Water Pumps. 4. IF power is not available THEN attempt to restore power from 1 COB 5. IF no RWS pumps can be started 6. IF offsite power was lost AND River Water Supply pumps fail to start from the diesel AND attempt to restart pumps. THEN Reduce recirc flow to 39 Mlbm/hr in accordance with IPOI 4 AND manually scram the reactor. THEN Attempt to start pumps manually , that have their start permissive light on. AND restore RWS makeup to the stilling basin. 7. Update the Online Risk Monitor for the status of River Water Supply Pumps. 8. Maximize Well Water flow for makeup to the Circ Pit , while maintaining

170 psig well water system pressure at the main plant BEECO Backflow Preventer.

9. Close the Circ Water Inlet to Slowdown Line valve M0-4253 to maintain Circ Water Pit inventory.

10. Secure Cooling Tower Fans as allowable

11. Secure one circ water pump and one cooling tower as soon as possible.

I AOP 410 Page 3 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOSS OF RIVER WATER SUPPLY FOLLOW-UP ACTIONS (continued)

12. Send an operator to the Intake Structure to verify alarms and check RWS pump breaker condition. NOTE The RHRSW Pumps should be kept in operation to support use of the RHR System as directed by EOPs (Torus Cooling/Shutdown Cooling).

The RHRSW Pumps may be secured when it is determined that ESW/RHRSW pit level cannot be maintained above 4 feet. Securing RHRSW Pumps prior to reaching 4 feet in the pits will preserve the ESW Supply to the SBDGs. 13. Minimize use of water from the RHRSW/ESW pit as follows: a. Secure RHRSW pumps unless required to support operation of the RHR System b. Shutdown any SBDG not required to ensure one Essential Bus is energized and/or required to ensure adequate core cooling. (1) Verify SBDG Cooling Valves CV-2080 and CV-2081 close when the respective SBDG is secured. c. Verify Well Water is available for cooling the operating Control Building Chiller and then secure ESW to the Control Building Chillers by unlocking and closing V-13-122 and V-13-125 on the Reactor Building 812' level. 14. Minimize heat addition to the Torus (Reliefs, HPCI, RCIC). 15. Use the Turbine Bypass valves and/or Bypass Jack for Reactor pressure control and Reactor cooldown and continue to bleed steam to the Main Condenser for as long as possible.

16. At 1C15 and 1C17, place the HI COND BACKPRESS BYPASS switches in BYPASS. 17. Notify Security at 7254 prior to opening Pumphouse doors to arrange for required Security compensatory measures.

I AOP 410 Page 4 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOSS OF RIVER WATER SUPPLY FOLLOW-UP ACTIONS (continued)

18. Establish makeup to the ESW/RHRSW Wet Pits from the Fire System using 2-1/2" and/or 5" hoses as follows: {C001} a. Using 2-1/2" hoses (N/A if not used): (1) Notify Mechanical Maintenance to remove the cover from outside hose head (octopus head). (2) Obtain 2-1/2" hoses from the warehouse and fire brigade trailer, and rig as many as possible (8 preferred) from the octopus head to the stilling basin. (3) Lash the hoses together with rope to prevent hose whip when the lines are charged. b. Using 5" hoses (N/A if not used): (1) Obtain 5" hoses from the B5b hose trailer. (2) Connect a 5" hose to any of the following fire hydrants as required:

• FH-1 located east of the Turbine Building
• FH-2 located southeast of the Turbine Building
• FH-7 located northeast of the Turbine Building (3) Rig 5" hoses as needed to the Stilling Basin. (4) Lash the hoses together with rope to prevent hose whip when the lines are charged. c. When directed by the CRS , start 1 P-48 or 1 P-49 and valve in hoses as necessary to maintain RHRSW/ESW pit level. d. Monitor RHRSW/ESW pit level at 1 C29 , computer points B279 and B280 , or on group display AOP 410. I AOP 410 Page 5 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOSS OF RIVER WATER SUPPLY FOLLOW-UP ACTIONS (continued)

19. Establish makeup to the RHRSW/ESW pits from GSW as follows: a. Verify at least one GSW pump is running. b. At 1 C452 (located in the B RHRSW/ESW pump room) open the following valves with the appropriate handswitch:

CV-8035A A RHRSW/ESW WET PIT CHLORINE HS-8035A INJECTION ISOLATION CV-8035B B RHRSW/ESW WET PIT CHLORINE HS-8035B INJECTION ISOLATION CV-8034 RHRSW/ESW DILUTION WATER HS-8034 SUPPLY VALVE c. In the CHLORINE BOOSTER PUMP ROOM , note and record the position of V-80-154, then fully open V-80-154 GSW Dilution Water Supply Balancing valve using a wrench from the tool board. NOTE NPSH requirement is 8 feet for the Circ Water Pumps. NPSH reguirement is 4 feet for the ESW, RHRSW, and GSW pumps. 20. Monitor the Circ Water Pit level at Computer Point F092 and secure Circ Water Pumps if level cannot be maintained or restored greater than 8 feet and GSW pumps as necessary to prevent cavitation. 21. Monitor RHRSW/ESW Pit level and secure RHRSW and ESW pumps as necessary to prevent cavitation. 22. When the status of each River Water Supply pump is known and there is at least one operable pump in each RWS loop, select the operable pumps on HSS-2911A and B to be the pumps to auto restart on the diesel. 23. Comply with Technical Specifications for River Water Supply. 24. When River Water Supply pump operation is restored, return system to normal operation per 01410. 25. Update the Online Risk Monitor for the status of River Water Supply Pumps. I AOP 410 Page 6 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOSS OF RIVER WATER SUPPLY FOLLOW-UP ACTIONS (continued)

26. Open and Lock open the following valves:

• V-13-122 ESW Loop A Return Header Isolation
• V-13-125 ESW Loop B Return Header Isolation

27. Independently verify the following valves are locked open:

• V-13-122 ESW Loop A Return Header Isolation IV
• V-13-125 ESW Loop B Return Header Isolation IV I AOP 410 Page 7 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOSS OF RIVER WATER SUPPLY PROBABLE ANNUNCIATORS 1C06A , A-1 " A" RWS PIT LO LEVEL A-2 "B" RWS PIT LO LEVEL A-3 " B" RWS PUMP 1P-117B TRIP A-4 " D" RWS PUMP 1P-117D TRIP B-1 " A" RWS PUMP 1P-117A TRIP B-2 " C" RWS PUMP 1P-117C TRIP B-5 " A" COOLING TOWER BASIN HI/LO LEVEL B-6 " B" COOLING TOWER BASIN HI/LO LEVEL D-1 " A" RHRSW/ESW PIT LO LEVEL D-2 " B RHRSW/ESW PIT LO LEVEL D-11 CIRC WATER PIT LO LEVEL PROBABLE INDICATIONS 1C06 -River Water makeup flow stopped at FR-4916 and FR-4917 or Fl-4916 and Fl-4917 I AOP 410 Page 8 of 24 Rev. 281 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 cooldown?

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. 2nd Stage Reheat can be left in N/A N/A N/A N/A Section 5, approximately 200 MWe, remove the 1 E-service and the turbine can be step (10) 18A[B] 2nd Stage Reheat System from tripped if necessary. service 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. will automatically go into the "hold" step (10) mode as power and flow are lowered. IPOl4, Commence primary containment purge No. This is only necessary if a N/A N/A N/A N/A Section 3 per 01573. Drywell entry is anticipated. step (11) IP014, At the refueling bridge, verify that the Main No. Control rod insertion will not be N/A N/A N/A N/A Section 3 Disconnect is closed and that the inhibited. step (13) SYSTEM START pushbutton has been depressed. IPOl4, Prior to disconnecting the generator from No. Aux Boiler is not required to N/A N/A N/A N/A Section 3 the grid, perform the following: (a) If accomplish shutdown. step (14) needed, start up the Auxiliary Boiler per 01 727. IPOl 4, Following Turbine Trip: (a) Verify that No. These systems can be left in N/A N/A N/A N/A Section 3 Reactor Coolant Chloride and service if necessary. step (22) Conductivity analyses have been performed. (b) Operate the Turbine Lube Oil 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 coo Id own? IPOl4, Shut down the following generator support No. These systems can be left in N/A NIA N/A N/A Section 3 systems, as desired: Isolated Phase Bus service if necessary. step (24) Cooling -01 698, Stator Water Cooling -01 697, H2 Seal Oil -01 695.1, H2 and CO2 Gas -01 695.2 IP014, Secure hydrogen, oxygen and/or air No. The Hydrogen Water N/A N/A N/A N/A Section 3 injection per 01 563, Hydrogen Water Chemistry System will secure itself step (26) Chemistrv. if left in service. IP014, 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 necessary to limit necessary to limit plant cooldown step (27) reactor vessel depressurization following rate. the reactor scram: (b) Start 1 P32 Mechanical Vacuum Pump per 01 691. (c) Secure the SJAEs and Offgas per 01 691 and 01672. IPOl4, For the remainder 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 depressurize the reactor plant cooldown rate. vessel to maintain a controlled cooldown (b) No -operated from the rate less than the TS Limit of 100°F in any Control Room 1 hour period. (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 with RCIC per 01 150 closed if necessary to limit (preferred

_method if MSIVs are closed), plant cooldown rate. (c) If desired cooldown with HPCI per 01 (e) No. The MSIVs car:i be 152 (RCIC may become inadequate as closed if necessary to limit pressure lowers) (d) Control steam flow plant cooldown rate. from the reactor vessel to the main condenser through steam seals and steam drains, (e) Secure _steam seals per 01 692 as required to limit 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 coo Id own? IPOl4, As plant cooldown continues perform the No. The MSIVs can be closed if N/A N/A N/A N/A Section 4 following: (NA if MSIVs are closed) (a) necessary to limit plant cooldown step (7) Control steam seal pressure 3 to 4 psig rate. using M0-1169, MAIN STEAM SUPPLY, M0-1170, REGULATOR BYPASS and/or M0-1171, MANUAL UNLOADER on 1C07, (b) Start 1P-32 MECHANICAL VACUUM PUMP per 01 691, (c) When reactor pressure approaches 500 psig or cooldown rate cannot be controlled within the limit, then secure SJAEs and Offgas -System per 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 psig above reactor pressure as reactor pressure decreases. Otherwise, N/A. IP014, At approximately 400 psig, secure the No. The Feed Pumps and N/A N/A N/A N/A Section 4 operating feed pump per 01 644. Condensate Pumps can be tripped step (8) from the Control Room if necessary. IPOl4, When RHR Shutdown Cooling Isolation No, this system can be placed in N/A N/A N/A N/A Section 4 Interlocks can be reset service from the Control Room if step (9) (approximately 100 psig), reset the necessary. isolation, then initiate Shutdown Cooling per 01149. IPOl4, Perform the following after the turbine trip, No. These systems can be left in N/A N/A N/A N/A Section 4 if needed: (a) Verify that Reactor Coolant service if necessary. step (10) Chloride and Conductivity analysis has been performed, (b) Operate the Turbine Lube Oil and Turning Gear System per 01 693.3, (c) Shutdown the Main Generator per 01 698, (d) Shutdown the Main 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 NIA 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) H2Seal Oil per 01695.1, (d) H2and CO2 Gas per 01 695.2, (e) Secure SJAEs per 01 691 and Offgas per 01 672 if not oreviouslv oerformed. 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. IPOl4, 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 01693.3, open Condenser Vacuum rate. Breaker valves V-03-67 and V-03-73. IP014, 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 oer 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 HAS 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. AOP 915 SHUTDOWN OUTSIDE CONTROL ROOM SECTION 1 I TRANSFER OF CONTROL TO THE REMOTE SHUTDOWN PANEL CONDITIONAL STATEMENTS 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 ATWS 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 (RCS) 3.4.6 RCS Specific Activity RCS Specffic Activity 3.4.6 LCO 3.4.6 The specific activity of the reactor coolant shall be limited to DOSE EQUIVALENT 1-131 specific activity s 0.2 µCi/gm. APPLICABILITY: MODE 1. HODES 2 and 3 with any main steam line not isolated. ACTIONS CONDITION REQUIRED ACTION COMPLErION TIME A. Reactor coolant -------------NOTE----------- specific activity LCO 3.0.4.c is applicable. > 0.2 µCi/gm and -------*-------------------- s 2.0 µCi/gm DOSE EQUIVALENT I-131. A.1 Determine DOSE Once per 4 hours EQUIVALENT I-131. Mm A.2 Restore DOSE 48 hours EQUIVALENT l-131 to within limits. B. Required Action and 8.1 Determine DOSE Once per 4 hours associated Completion EQUIVALENT I-131. Time of Condition A not met. AHO .QB B.2.1 Isolate all maih 12 hours steam 1 i nes. Reactor Coolant specific activity> 2.0 DB µCi/gm DOSE EQUIVALENT I-131. (continued) 2.0 uci/gm chosen as EAL threashold since levels above that activitx U directly influence continued plant operation. OAEC 3.4-13 Af!D 255"' I \_) u 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 hour period* and c. s 2 gpm increase in unidentified LEAKAGE within the previous 24 hour period in MODE 1. APPLICABILITY:

MODES 1. 2. and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Unidentified LEAKAGE A.I Reduce LEAKAGE to 4 hours not within limit. within limits. OR Total LEAKAGE not within limit. . . B. Unidentified LEAKAGE B.1 Reduce unidentified 4 hours increase not within LEAKAGE increase to 1 imit. within limits. OR (continued) Developer Notes: For EAL #1 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 ~Fo~j\.L #2 ent e r the higher of25 gpm or th e value specified in DAEC's Technical Specifications for this type ofleakage.

)-7 DAEC uses a total leakage (identified+ unidentifi e d) spec of25 gpm averaged over 24 hour 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 to the critical condition. (10) When a control rod reaches position 48 , perform 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 1C05A, 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 is 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 power. j 1POl 2 Page 5 of 46 Rev. 148 DAEC EMERGENCY PLAN EMERGENCY COMMUNICATIONS (1)-EOF

• FPLE Duane Arnold Corpont1e Offices
• County Sheriff's Offices
• Palo Fre Departmen1

' Merty Hosphl

• Slate Highway Petrol
• Slate Emergency Man,gemen t Division
• UnHersly of loWo 'N RC 'DOE 'FEMA
• Li'ln County Emergency Management
• Benton County Emergency M1n1gement FIGURE F-5 DAEC TELEPHONE SYSTEMS -------------1 1,----, I I I I
• emergency unlisted Ines (Blue Phooes) in Con~ol Room. TSC , CAS , SAS Local Telephone Company Central Office Normal Telephone Services Qwest (Cedar Rapids) To other Be l Central Offices DAEC PBX Room Sa t ellite C o mmunicati o ns DAEC Microwave to Alliant Tower j Alliant Tower Microwave toDAEC Offsite Laboratory and Decontamination Center EOF Emergency Operations Fac i lity (1) Denotes a Dedicated Line Joint Public Information Center SECTION 'F' Rev. 29 Page 15 of 17 DAECEMERGENCYPLAN SECTION 'F' EMERGENCY COMMUNICATIONS Rev.29 Page 16 of 17 FIGURE F-6 FEDERAL TELEPHONE SYSTEM (FTS-2001)

NRC EROS EOF TSC Q) EN S © HPN @ RSC L © P M C L @ MCL DAEC EMERGENCY PLAN SECTION 'F' Rev. 29 EMERGENCY COMMUNICATIONS Page 17 of 17 FIGURE F-7 ALL-CALL TELEPHONE SYSTEM Interne t LINN COUNTY Sheriff's Office Backup F acility Malnfad1ity t-------..'\ V', -~-------~ I AOP 902 FLOOD FOLLOW-UP ACTIONS (continued) NOTE River Water Level is required to be recorded hourly and verified to be < 757 feet whenever river level is >753 feet , IAW TLCO 3.7.1. 9. IF river water level THEN Commence recording river water level hourly. reaches 753' 10. Pre-stage equipment in the location per Attachment 3 for the Pump House, Turbine Building , Control Building , Reactor and Recombiner Building , and Radwaste/LLRPSF CAUTION Deenergizing MCC 189106 [182106] will cause RWS Screen Wash Pump 1 P-112A[B] to be inoperable. Refer to applicable Tech. Spec. Sections. NOTE Once level is greater than 754', the only method to get into the Intake Structure without walking through water is to remove the louver on the Intake Air Supply and use a boat to get to the building.

11. Prior to river level reaching 754', open the following breakers to prevent energizing electrical equipment in the Intake Structure that may become submerged. Other loads at the Intake are on the 2nd floor and will be deenergized later at higher river levels. I AOP 902 Breaker Load 189106 Screen Wash Pump 1 P-112A 189107 Traveling Screen Drive 1 F-36A 189108 Screen Wash Control Panel 1C-154A 189109 189111 189112 182106 182107 182108 182109 182110 182111 182112 182114 Radial Sand and Side Gate Hoist 1 H-26A Intake Structure 480VAC Power Receptacles Screen Wash Nozzle Shutoff M0-2902 Screen Wash Pump 1 P-1128 Traveling Screen Drive 1 F-368 Screen Wash Control Panel 1 C-1548 Intake Structure Trash Rake 1 S-83 Intake Structure 480VAC Power Receptacles Radial Sand and Side Gate Hoist 1 H-268 Screen Wash Nozzle Shutoff M0-2903 Instrument Enclosure 1 C412 Page 10 of 49 Rev. 54 j I AOP 902 FLOOD FOLLOW-UP ACTIONS (continued)

12. When river level reaches 756', at the Pump House, Turbine Building , Control Building , Reactor and Recombiner Building , and Radwaste/LLRPSF implement action per Attachment 3 to pump water/monitor levels that may leak into these buildings in the future if river level continues to rise. 13. IF it is determined that the THEN a. flood level will reach Shut down the plant to Cold Shutdown 757' OR does reach 757' b. Deenergize equipment not required to shutdown the plant or for personal safety or for security or fuel pool cooling , at the following:

(1) Cooling towers (2) Administration Bldg. (3) Pump House (4) Low Level Radwaste Process Storage Facility (LLRPSF) Building (5) Machine Shop (6) Offgas Retention Facility (7) Security Bldg (8) TSC Bldg (9) Data Acquisition Center (10) Badging Center (11) Electric Shop (12) Construction support Center (13) West/East Warehouse (14) Fabrication Shop (15) Sewage Treatment Plant (16) Shooting Range (17) Barn c. Refer to EPIP 1.1 for EAL assessment. CAUTION Major 4160V loads should be started at approximately 10 second intervals to avoid overloading the diesel generator. Diesel generator load should be monitored at 1 C08 as bus loads are added. I AOP 902 d. Start both SBDGs and transfer the 4160VAC Essential Susses 1 A3 and 1A4 to the SBDGs per 01 304.2 , Section 7.6. Page 11 of 49 Rev. 541 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH ABNORMAL OPERATING PROCEDURE AOP 410 LOSS O F RI VER WATE R SUPPLY/HIGH RIVER BED ELEV A T ION/LOW RIVER WATER DEPTH Usage Level REFERENCE NOTE This AOP is normally coordinated by the Reactor Operator. Record the following: Date/Time: ______ / ____ Initials: __ _ NOTE: User shall perform and document a Temp Issue/Rev. Check to ensure rev i sion is current , in accordance with procedure use and adherence requirements. Enter the following as appl i cable: LOSS OF RIVER WATER SUPPLY HIGH RIVER BED ELEVATION LOW RIVER WATER DEPTH NOTE Refer to EPIP 1.1 for EAL ASSESSMENT. I AOP 410 Page 1 of 24 PAGE PAGE PAGE 2 9 16 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOSS OF RIVER WATER SUPPLY NOTE (The portions of this procedure that minimize river make-up flow and maximizes (make-up flow from other sources may be used as necessary in the event of an (intrusion of excess foreign material from the Cedar River and/or Intake Structure (w ithout a Loss of River Water Supply. IMMEDIATE ACTIONS 1. None AUTOMATIC ACTIONS -CV-491 4 and CV-491 5 open and CV-4910A and CV-49108 close on low level in the ESW/RHRSW wet pi t s I AOP 410 Page 2 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOSS OF RIVER WATER SUPPLY FOLLOW-UP ACTIONS 1. Establish critical parameter monitoring of Circ Pit Level and ESW/RHRSW Pit Level , as priorities allow. 2. IF power is available AND annunciator 1 C06A (A-1 [2]) " A"[" B") RWS PIT LO LEVEL is not on THEN attempt to start standby pumps in both RWS Subsystems as necessary to restore needed makeup flow. 3. IF standby pumps d i d not start THEN attempt to restart the tripped River Water Pumps. 4. IF power is not available THEN attempt to restore power from 1 COS 5. IF no RWS pumps can be started M ANU.-\L SC R AM 6. IF offsite power was lost AND River Water Supply pumps fail to start from the diesel AND attempt to restart pumps. THEN Reduce recirc flow to 39 Mlbm/hr in accordance with IPOI 4 AND manually scram the reactor. THEN Attempt to start pumps manually , that have their start permissive light on. AND restore RWS makeup to the stilling basin. 7. Update the Online Risk Monitor for the status of River Water Supply Pumps. 8. Maximize Well Water flow for makeup to the Circ Pit , while maintaining

170 psig well water system pressure at the main plant BEECO Backflow Preventer.

9. Close the Circ Water Inlet to Slowdown Line valve M0-4253 to maintain Circ Water Pit inventory.

10. Secure Cooling Tower Fans as allowable

11. Secure one circ water pump and one cooling tower as soon as possible. I AOP 410 Page 3 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOSS OF RIVER WATER SUPPLY FOLLOW-UP ACTIONS (continued)

12. Send an operator to the Intake Structure to verify alarms and check RWS pump breaker condition.

NOTE The RHRSW Pumps should be kept in operation to support use of the RHR System as directed by EOPs (Torus Cooling/Shutdown Cooling). The RHRSW Pumps may be secured when it is determined that ESW/RHRSW pit level cannot be maintained above 4 feet. Securing RHRSW Pumps prior to reaching 4 feet in the pits will preserve the ESW Supply to the SBDGs. 13. Minimize use of water from the RHRSW/ESW p i t as follows: a. Secure RHRSW pumps unless required to support operation of the RHR System b. Shutdown any SBDG not required to ensure one Essential Bus is energized and/or required to ensure adequate core cooling. (1) Verify SBDG Cooling Valves CV-2080 and CV-2081 close when the respective SBDG is secured. c. Verify Well Water is available for cooling the operating Control Building Chiller and then secure ESW to the Control Building Chillers by unlocking and closing V-13-122 and V-13-125 on the Reactor Building 812' level. 14. Minimize heat addition to the Torus (Reliefs , HPCI, RCIC). 15. Use the Turbine Bypass valves and/or Bypass Jack for Reactor pressure control and Reactor cooldown and continue to bleed steam to the Main Condenser for as long as possible. 16. At 1C15 and 1C17 , place the HI COND BACKPRESS BYPASS switches in BYPASS. 17. Notify Security at 7254 prior to opening Pumphouse doors to arrange for required Security compensatory measures. I AOP 410 Page 4 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOSS OF RIVER WATER SUPPLY FOLLOW-UP ACTIONS (continued)

18. Establish makeup to the ESW/RHRSW Wet Pits from the Fire System using 2-1/2" and/or 5" hoses as follows: {C001} a. Using 2-1/2" hoses (N/A if not used): (1) Notify Mechanical Maintenance to remove the cover from outside hose head (octopus head). (2) Obtain 2-1/2" hoses from the warehouse and fire brigade trailer, and rig as many as possible (8 preferred) from the octopus head to the stilling basin. (3) Lash the hoses together with rope to prevent hose whip when the lines are charged. b. Using 5" hoses (N/A if not used): (1) Obtain 5" hoses from the 85b hose trailer. (2) Connect a 5" hose to any of the following fire hydrants as required:

• FH-1 located east of the Turbine Building
• FH-2 located southeast of the Turbine Building
• FH-7 located northeast of the Turbine Building (3) Rig 5" hoses as needed to the Stilling Basin. (4) Lash the hoses together with rope to prevent hose whip when the lines are charged. c. When directed by the CRS , start 1 P-48 o r 1 P-49 and valve in hoses as necessary to maintain RHRSW/ESW pit level. d. Monitor RHRSW/ESW pit level at 1C29 , computer points 8279 and 8280 , or on group display AOP 410. I AOP 410 Page 5 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOSS OF RIVER WATER SUPPLY FOLLOW-UP ACTIONS (continued) 1 9. Establish makeup to the RHRSW/ESW pits from GSW as follows: a. Verify at least one GSW pump is running. b. At 1 C452 (located in the B RHRSW/ESW pump room) open the following valves with the appropriate handswitch

CV-8035A A RHRSW/ESW WET PIT CHLORINE HS-8035A INJECTION ISOLATION CV-8035B B RHRSW/ESW WET PIT CHLORINE HS-8035B INJECTION ISOLATION CV-8034 RHRSW/ESW DILUTION WATER HS-8034 SUPPLY VALVE c. In the CHLORINE BOOSTER PUMP ROOM , note and record the position of V-80-154 , then fully open V-80-154 GSW Dilution Water Supply Balancing valve using a wrench from the tool board. NOTE NPSH requirement is 8 feet for the Circ Water Pumps. NPSH requirement is 4 feet for the ESW , RHRSW, and GSW pumps. 20. Monitor the Circ Water Pit level at Computer Point F092 and secure Circ Water Pumps if level cannot be maintained or restored greater than 8 feet and GSW pumps as necessary to prevent cavitation. 21. Monitor RHRSW/ESW Pit level and secure RHRSW and ESW pumps as necessary to prevent cavitation. 22. When the status of each River Water Supply pump is known and there is at least one operable pump in each RWS loop , select the operable pumps on HSS-2911A and B to be the pumps to auto restart on the diesel. 23. Comply with Technical Specifications fo r River Water Supply. 24. When River Water Supply pump operation is restored , return system to normal operation per 01 410. 25. Update the Online Risk Monitor for the status of River Water Supply Pumps. I AOP 410 Page 6 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOSS OF RIVER WATER SUPPLY FOLLOW-UP ACTIONS (continued)

26. Open and Lock open the following valves:

• V-13-122 ESW Loop A Return Header Isolation
• V-13-125 ESW Loop B Return Header Isolation

27. Independently verify the following valves are locked open:

• V-13-122 ESW Loop A Return Header Isolation IV
• V-13-125 ESW Loop B Return Header Isolation IV I AOP 410 Page 7 of 24 Rev. 281 AOP 410 LOSS OF RIVER WATER SUPPLY/HIGH RIVER BED ELEVATION/LOW RIVER WATER DEPTH LOSS OF RIVER WATER SUPPLY PROBABLE ANNUNCIATORS 1C06A , A-1 "A" RWS PIT LO LEVEL A-2 "B" RWS PIT LO LEVEL A-3 " B" RWS PUMP 1 P-117B TRIP A-4 " D" RWS PUMP 1P-117D TRIP B-1 " A" RWS PUMP 1P-117A TRIP B-2 " C" RWS PUMP 1P-117C TRIP B-5 " A" COOLING TOWER BASIN HI/LO LEVEL B-6 " B" COOLING TOWER BASIN HI/LO LEVEL D-1 " A" RHRSW/ESW PIT LO LEVEL D-2 " B" RHRSW/ESW PIT LO LEVEL D-11 CIRC WATER PIT LO LEVEL PROBABLE INDICATIONS 1C06 -River Water makeup flow stopped at FR-4916 and FR-4917 or Fl-4916 and Fl-4917 I AOP 410 Page 8 of 24 Rev. 281 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 NRG 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 NRG 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. Tt')ere 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) is 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 equipment, 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 reduction 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 within 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 response 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 ICs, 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 reliability of the second train is a concern. In addition, escalation to the Alert classification should not occur if the damage from the hazardous event is limited to a SAFETY SYSTEM that was inoperable, or out of service, prior to the event occurring. As such, the proposed 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 Fission Product Barrier and/or Abnormal Radiation Levels/Radiological Effluent recognition categories would provide an escalation path to a Site Area Emergency or a General Emergency. The 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 NRG-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 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.

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 NRG-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): D INFORMATION ONLY, MAINTAIN EPFAQ 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 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 4

! L_ 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/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. Operators will make a determination of VISIBLE DAMAGE based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. Escalation of the emergency classification level would be via IC 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 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) a. The occurrence of ANY of the following hazardous events: Basis:

• 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 L___ __ -, 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/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. 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 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 maintain level above -25 inches and spray cooling cannot be established
• 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) maintain level above -39 inches while spray cooling DWfT-3 D Max i mize drywell cool i ng. IF .-If necessary , bypass drywell cooling i solation and fan speed interlock (Defeat 4). THEN Drywell pressure drops below 2.0 psig Verify containmen t sprays isolate. DWfT-4 drywell tempe r ature reaches 280°F D DWfT-5 D IF torus water level i s below 13.5 ft , AND drywell temperature is below the Drywell Spray In i t i at i on Limit (Graph 7), THEN 1. Shut down recirc pumps. 2. I n i tiate d ry well sprays using only pumps not required for adequate core cooling (Table 9). DWfT-6 drywell temperature cannot be restored and maintained below 340°F 5 LJ IF Torus pressure drops below 2.0 psi! Drywell pressure drops below 2.0 p!: Pr i mary containment pressure reduc required to:
• Restore and ma i ntain adequate c OR
• Reduce the total offsite radiation PC/P-3 D PC/P-4 D Initiate to1 required f 11r OK con leve belc PC/P-5 D *~*

L AOP 301.1 STATION BLACKOUT STATION BLACKOUT FOLLOW-UP ACTIONS (continued)

18. Direct an operator and electricians (if available) to perform Attachment 10 Alternate AC power to 125VDC and 250VDC chargers. NOTE The inverters will automatically trip at 105 VDC decreasing.

19. After one hour has elapsed , dispatch an operator to implement Attachment 13, Load Shedding to Preserve Station Batteries (from AOP 301.1 hanging file). Attachment 13 is to be completed within two hours of the SBO event. 20. IF a SBDG is available for operation , THEN restore power to its essential bus per Restoration of Standby Diesel Generator Power section. 21. WHEN the DAEC Switchyard inspection has been completed, THEN restore power to the switchyard per Restoration of Offsite Power section. 22. WHEN sufficient offsite power becomes available , THEN restore power to the non-essential buses per AOP 304.1. I AOP 301.1 Page 7 of 45 Rev. 61 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 1A4 480V Load Centers Load Centers 183,189 184, 1820 480V Motor Control Motor Control Centers Centers 1832, 1834 1842, 1844 125 V Distribution Panels Distribution 1D10, 1D11, Panels 1D13 1D20 , 1D21, RCIC Motor Control 1D23 Center 1D14 250V NIA Distribution Panel 1D40 Motor Control Centers 1 D41 and 1 D42 (a) Each division of the AC and DC electrical power distribution systems is a subsystem.

DAEC B 3.8-73 Amendment 223 ATTACHMENT 5 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER LICENSE AMENDMENT REQUEST TSCR-166 DAEC EAL SCHEME WALLBOARDS [FOR INFORMATION ONLY] 2 pages follow}}