NG-18-0090, Response to Request for Additional Information Regarding License Amendment Request (TSCR-166), Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors. (: Difference between revisions

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{{Adams
{{Adams
| number = ML18212A232
| number = ML18212A231
| issue date = 07/26/2018
| issue date = 07/26/2018
| title = Response to Request for Additional Information Regarding License Amendment Request (TSCR-166), Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors. (Pa
| title = Response to Request for Additional Information Regarding License Amendment Request (TSCR-166), Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors. (Pa
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| case reference number = NG-18-0090
| case reference number = NG-18-0090
| document type = Response to Request for Additional Information (RAI)
| document type = Response to Request for Additional Information (RAI)
| page count = 296
| page count = 125
}}
}}


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{{#Wiki_filter:NEI 99 QI (RevisioA
* I ATTACHMENT 2 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELATING TO LICENSE AMENDMENT REQUEST TSCR-166 UPDATED CLEAN COPY OF THE PROPOSED DAEC EAL SCHEME 125 pages follow Duane Arnold Energy Center (DAEC) Emergency Action Levels Technical Bases Document TBD,2018 TABLE OF CONTENTS 1 BASIS FOR EMERGENCY ACTION LEVELS .................................................................
: 6) }>Joye me er 2Q 12 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS 17 9 
1 1.1 OPERATING REACTORS ..................................................................................................
}>ffil 99 QI (RevisioR e) November 2012 Tal>le H 1: Reeognition CategoFY "H" Initiating Condition MatFix UNUSUAL EVENT HUl Confirmed SECURITY CONDITION or Op. A1edes: All HU2 Seismic e 1 ,rent greater than OBE Op. A1edes: All HUJ Hazardous e¥eflt Op. A1edes: All HU4 FIRE potentially degrading the level of safety of the plant. Op. },1edes: All ALERT HAl HOSTILE ACTION within the OWNER CO~ffROLLED AREA or airborne attack threat *within 30 minutes. Op. },fades: All HA5 Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdo 1 Nn. Op. ,\1edes: All Hf...(i Control Room evacuation resulting in transfer of plant control to alternate locations.
1 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSl) .....................................
Op. ,\1edes: All SITE AREA EMERGENCY HSl HOSTILE ACTION within the PROTECTED AREA. Op. A1edes: All HS(i Inability to control a key safety function from outside the Control Room. Op. },1edes: All 180 I GENERAL EMERGENCY HGl HOSTILE ACTION resulting in loss of physical control of the facility.
2 1.3 NRC ORDEREA-12-051
Op. ,\1edes: All Table intended for use by 1 EAL deYelopers.  
................................................................................................
: lnelusion in lieensee : doeurnents is not required.
3 2 KEY TERMINOLOGY USED IN DAEC EAL SCHEME .....................................................
L------------------J UNUSUAL EVENT HU7 Other conditions exist which in the judgment of the Emergency Director 1tvarrant deelaration of a (NO)UE. Op. J,1odes: All ALERT HA7 Other conditions e>,ist which in the judgment of the Emergency Director warrant deelaration ofan Alert. Op. A/odes: All ECL: Notification of Unusual Event SITE AREA EMERGENCY HS7 Other conditions e>,ist which in the judgment of the Emergency Director warrant deelaration of a Site Area Emergency.
4 2.1 EMERGENCY CLASSIFICATION LEVEL (ECL) ..............
Op. A1ode s: All Initiating Condition:
.' ................................................
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) ...............................
7 3.2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS ....................
10 3.3 DAEC-SPECIFIC ORGANIZATION AND PRESENTATION OF GENERIC INFORMATIONll 3.4 IC AND EAL MODE APPLICABILITY
............................................................................
12 4 DAEC SCHEME DEVELOPMENT
......*.***.****......*****.....*.*........*.*....*......**.*......***..*.*.....
13 4.1 GENERAL DEVELOPMENT PROCESS ............................................................................
13 4.2 CRITICAL CHARACTERISTICS
......................................................................................
13 4.3 INSTRUMENTATION USED FOR EALS ..........................................................................
14 4.4 EAL/THRESHOLD REFERENCES TO AOP AND EOP SETPOINTS/CRITERIA
..............
14 5 GUIDANCE ON USING THE DAEC EALS ....******.......******....
ll ****************************************
15 5.1 GENERAL CONSIDERATIONS
........................................................................................
15 5.2 CLASSIFICATION METHODOLOGY
...............................................................................
16 5.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS
........................................
16 5 .4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION
..............................
17 5.5 CLASSIFICATION OF IMMINENT CONDITIONS
.............................................................
17 5.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING
.................
17 5.7 CLASSIFICATION OF SHORT-LIVED EVENTS ...............................................................
18 5.8 CLASSIFICATION OF TRANSIENT CONDITIONS
............................................................
18 5 .9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION
..............
19 5.10 RETRACTION OF AN EMERGENCY DECLARATION
.......................................................
19 11 6 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS ........................
20 7 COLD SHUTDOWN/
REFUELING SYSTEM MALFUNCTION ICS/EALS ....................
36 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS ..............
58 9 FISSION PRODUCT BARRIER ICS/EALS ******************************************************************
60 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS .........
75 11 SYSTEM MALFUNCTION ICS/EALS ***************************************************************************
96 APPENDIX A -ACRONYMS AND ABBREVIATIONS
........................................................
A-1 APPENDIX B -DEFINITIONS
*******************************************************************************************
1-1 iii L___ -
DUANE ARNOLD EMERGENCY ACTION LEVELS TECHNICAL BASIS DOCUMENT 1 BASIS FOR EMERGENCY ACTION LEVELS 1.1 OPERATING REACTORS Title 10, Code of Federal Regulations (CFR), Energy, contains the U.S. Nuclear Regulatory Commission (NRC) regulations that apply to nuclear power facilities.
Several of these regulations govern various aspects of an emergency classification scheme. A review of the relevant sections listed below will aid the reader in understanding the key terminology provided in Section 3.0 of this document.
* 10 CFR § 50.47(a)(l)(i)
* 10 CFR § 50.47(b)(4)
* 10 CFR § 50.54(q)
* 10 CFR § 50.72(a)
* 10 CFR § 50, Appendix E, IV.B, Assessment Actions
* 10 CFR § 50, Appendix E, IV.C, Activation of Emergency Organization The above regulations are supplemented by various regulatory guidance documents.
Three documents of particular relevance to NEI 99-01 are: NUREG-0654/FEMA-REP-1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, October 1980. [Refer to Appendix 1, Emergency Action Level Guidelines for Nuclear Power Plants] NUREG-1022, Event Reporting Guidelines 10 CFR § 50. 72 and§ 50. 73 Regulatory Guide 1.101, Emergency Response Planning and Preparedness for Nuclear Power Reactors 1 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) Selected guidance in NEI 99-01 is applicable to licensees electing to use their 10 CFR 50 emergency plan to fulfill the requirements of 10 CFR 72.32 for a stand-alone ISFSI. The emergency classification levels applicable to an ISFSI are consistent with the requirements of 10 CFR 50 and the guidance in NUREG 0654/FEMA-REP-1.
The initiating conditions germane to a 10 CFR 72.32 emergency plan (as described in NUREG-1567) are subsumed within the classification scheme for a 10 CFR 50.47 emergency plan. The generic ICs and EALs for an ISFSI are presented in Section 8, ISFSI ICs/EALs.
IC E-HUl covers the spectrum of credible natural and man-made events included within the scope of an ISFSI design. This IC is not applicable to installations or facilities that may process and/or repackage spent fuel (e.g., a Monitored Retrievable Storage Facility or an ISFSI at a spent fuel processing facility).
In addition, appropriate aspects of IC HUI and IC HAI should also be included to address a HOSTILE ACTION directed against an ISFSI. The analysis of potential onsite and off site consequences of accidental releases associated with the operation of an ISFSI is contained in NUREG-1140, A Regulatory Analysis on Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees.
NUREG-1140 concluded that the postulated worst-case accident involving an ISFSI has insignificant consequences to public health and safety. This evaluation shows that the maximum offsite dose to a member of the public due to an accidental release of radioactive materials would not exceed 1 rem Effective Dose Equivalent.
Regarding the above information, the expectations for an offsite response to an Alert classified under a 10 CFR 72.32 emergency plan are generally consistent with those for a Notification of Unusual Event in a 10 CFR 50.47 emergency plan (e.g., to provide assistance ifrequested).
Also, the licensee's Emergency Response Organization (ERO) required for 10 CFR 72.32 emergency plan is different than that prescribed for a 10 CFR 50.47 emergency plan (e.g., no emergency technical support function).
2 1.3 NRC ORDER EA-12-051 The Fukushima Daiichi accident of March 11, 2012, was the result of a tsunami that exceeded the plant's design basis and flooded the site's emergency electrical power supplies and distribution systems. This caused an extended loss of power that severely compromised the key safety functions of core cooling and containment integrity, and ultimately led to core damage in three reactors.
While the loss of power also impaired the spent fuel pool cooling function, sufficient water inventory was maintained in the pools to preclude fuel damage from the loss of cooling. Following a review of the Fukushima Daiichi accident, the NRC concluded that several measures were necessary to ensure adequate protection of public health and safety under the provisions of the backfit rule, 10 CFR 50.109(a)(4)(ii).
Among them was to provide each spent fuel pool with reliable level instrumentation to significantly enhance the ability of key decision-makers to allocate resources effectively following a beyond design basis event. To this end, the NRC issued Order EA-12-051, Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, on March 12, 2012, to all US nuclear plants with an operating license, construction permit, or combined construction and operating license. NRC Order EA-12-051 states, in part, "All licensees
... shall have a reliable indication of the water level in associated spent fuel storage pools capable of supporting identification of the following pool water level conditions by trained personnel:
(1) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck, and (3) level where fuel remains covered and actions to implement make-up water addition should no longer be deferred." To this end, all licensees must provide:
* A primary and back-up level instrument that will monitor water level from the normal level to the top of the used fuel rack in the pool;
* A display in an area accessible following a severe event; and
* Independent electrical power to each instrument channel and provide an alternate remote power connection capability.
NEI 12-02, Industry Guidance for Compliance with NRC Order EA-12-051, "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, "provides guidance for complying with NRC Order EA-12-051.
NEI 99-01, Revision 6, includes three EALs that reflect the availability of the enhanced spent fuel pool level instrumentation associated with NRC Order EA-12-051.
These EALs are included within ICs RA2, RS2, and RG2. 3 2 KEY TERMINOLOGY USED IN DAEC EAL SCHEME There are several key terms that appear throughout the EAL methodology.
These terms are introduced in this section to support understanding of subsequent material.
As an aid to the reader, the following table is provided as an overview to illustrate the relationship of the terms to each other. Emergency Classification Level Unusual Event I Alert I SAE I GE Initiating Condition Initiating Condition Initiating Condition Initiating Condition Emergency Action Emergency Action Emergency Action Emergency Action Level (1) Level (1) Level (1) Level (1)
* Operating Mode
* Operating Mode
* Operating Mode
* Operating Mode Applicability Applicability Applicability Applicability
* Notes
* Notes
* Notes
* Notes
* Basis
* Basis
* Basis
* Basis (1) -When making an emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of an Initiating Condition.
This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes and the informing Basis information.
In the Recognition Category F matrices, EALs are referred to as Fission Product Barrier Thresholds; the thresholds serve the same function as an EAL. 2.1 EMERGENCY CLASSIFICATION LEVEL (ECL) One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are: Notification of Unusual Event (NOUE) Alert Site Area Emergency (SAE) General Emergency (GE) 2.1.1 Notification of Unusual Event (NOUE) Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.
No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs. Purpose: The purpose of this classification is to assure that the first step in future response has been carried out, to bring the operations staff to a state of readiness, and to provide systematic handling of unusual event information and decision-making.
4 
--------------
-----------2.1.2 Alert Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA PAG exposure levels. Purpose: The purpose of this classification is to assure that emergency personnel are readily available to respond if the situation becomes more serious or to perform confirmatory radiation monitoring if required, and provide off site authorities current information on plant status and parameters.
2.1.3 Site Area Emergency Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to, equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the site boundary.
Purpose: The purpose of the Site Area Emergency declaration is to assure that emergency response centers are staffed, to assure that monitoring teams are dispatched, to assure that personnel required for evacuation of near-site areas are at duty stations if the situation becomes more serious, to provide consultation with offsite authorities, and to provide updates to the public through government authorities.
2.1.4 General Emergency (GE) Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.
Releases can be reasonably expected to exceed EPA P AG exposure levels off site for more than the immediate site area. Purpose: The purpose of the General Emergency declaration is to initiate predetennined protective actions for the public, to provide continuous assessment of information from the licensee and offsite organizational measurements, to initiate additional measures as indicated by actual or potential releases, to provide consultation with offsite authorities, and to provide updates for the public through government authorities.
5 2.2 INITIATING CONDITION (IC) An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences.
Discussion:
An IC describes an event or condition, the severity or consequences of which meets the definition of an emergency classification level. An IC can be expressed as a continuous, measurable parameter (e.g., RCS leakage), an event (e.g., an earthquake) or the status of one or more fission product barriers (e.g., loss of the RCS barrier).
Appendix 1 of NUREG-0654 does not contain example Emergency Action Levels (EALs) for each ECL, but rather Initiating Conditions (i.e., plant conditions that indicate that a radiological emergency, or events that could lead to a radiological emergency, has occurred).
NUREG-0654 states that the Initiating Conditions form the basis for establishment by a licensee of the specific plant instrumentation readings (as applicable) which, if exceeded, would initiate the emergency classification.
Thus, it is the specific instrument readings that would be the EALs. 2.3 EMERGENCY ACTION LEVEL (EAL) A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given emergency classification level. Discussion:
EAL statements may utilize a variety of criteria including instrument readings and status indications; observable events; results of calculations and analyses; entry into particular procedures; and the occurrence of natural phenomena.
2.4 FISSION PRODUCT BARRIER THRESHOLD A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier. Discussion:
Fission product barrier thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment.
This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment.
The primary fission product barriers are: Fuel Clad Reactor Coolant System (RCS) Containment Upon determination that one or more fission product barrier thresholds have been exceeded, the combination of barrier loss and/or potential loss thresholds is compared to the fission product barrier IC/EAL criteria to determine the appropriate ECL. In some accident sequences, the ICs and EALs presented in the Abnormal Radiation Levels/ Radiological Effluent (R) Recognition Category will be exceeded at the same time, or shortly after, the loss of one or more fission product barriers.
This redundancy is intentional as the former ICs address radioactivity releases that result in certain offsite doses from whatever cause, including events that might not be fully encompassed by fission product barriers (e.g., spent fuel pool accidents, design containment leakage following a LOCA, etc.). 6 L 3 DESIGN OF THE DAEC EMERGENCY CLASSIFICATION SCHEME 3.1 ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS (ECLS) An effective emergency classification scheme must incorporate a realistic and accurate assessment of risk, both to plant workers and the public. There are obvious health and safety risks in underestimating the potential or actual threat from an event or condition; however, there are also risks in overestimating the threat as well (e.g., harm that may occur during an evacuation).
The DAEC emergency classification scheme attempts to strike an appropriate balance between reasonably anticipated event or condition consequences, potential accident trajectories, and risk avoidance or minimization.
There are a range of "non-emergency events" reported to the US Nuclear Regulatory Commission (NRC) staff in accordance with the requirements of 10 CFR 50.72. Guidance concerning these rep01iing requirements, and example events, are provided in NUREG-1022.
Certain events reportable under the provisions of 10 CFR 50.72 may also require the declaration of an emergency.
In order to align each Initiating Conditions (IC) with the appropriate ECL, it was necessary to determine the attributes of each ECL. The goal of this process is to answer the question, "What events or conditions should be placed under each ECL ?" The following sources provided information and context for the development ofECL attributes.
Assessments of the effects and consequences of different types of events and conditions DAEC abnormal and emergency operating procedure setpoints and transition criteria DAEC Technical Specification limits and controls Offsite Dose Assessment Manual (ODAM) radiological release limits Review of selected Updated Final Safety Analysis Report (UFSAR) accident analyses Environmental Protection Agency (EPA) Protective Action Guidelines (PAGs) NUREG 0654, Appendix 1, Emergency Action Level Guidelines for Nuclear Power Plants Industry Operating Experience Input from DAEC subject matter experts The following ECL attributes are used to aid in the development of ICs and Emergency Action Levels (EALs ). The attributes may be useful in briefing and training settings (e.g., helping an Emergency Director understand why a particular condition is classified as an Alert). 7 3.1.1 Notification of Unusual Event (NODE) A Notification of Unusual Event, as defined in section 2.1.1, includes but is not limited to an event or condition that involves: (A) A precursor to a more significant event or condition. (B) A minor loss of control of radioactive materials or the ability to control radiation levels within the plant. (C) A consequence otherwise significant enough to warrant notification to local, State and Federal authorities.
3.1.2 Alert An Alert, as defined in section 2.1.2, includes but is not limited to an event or condition that involves: (A)A loss or potential loss of either the fuel clad or Reactor Coolant System (RCS) fission product barrier. (B) An event or condition that significantly reduces the margin to a loss or potential loss of the fuel clad or RCS fission product barrier. (C) A significant loss of control of radioactive materials resulting in an inability to control radiation levels within the plant, or a release of radioactive materials to the environment that could result in doses greater than 1 % of an EPA PAG at or beyond the site boundary. (D)A HOSTILE ACTION occurring within the OWNER CONTROLLED AREA, including those directed at an Independent Spent Fuel Storage Installation (ISFSI). 3.1.3 Site Area Emergency (SAE) A Site Area Emergency, as defined in section 2.1.3, includes but is not limited to an event or condition that involves: (A)A loss or potential loss of any two fission product barriers -fuel clad, RCS and/or containment. (B) A precursor event or condition that may lead to the loss or potential loss of multiple fission product barriers within a relatively short period of time. Precursor events and conditions of this type include those that challenge the monitoring and/or control of multiple SAFETY SYSTEMS. (C) A release of radioactive materials to the environment that could result in doses greater than 10% of an EPA PA G at or beyond the site boundary. (D)A HOSTILE ACTION occurring within the plant PROTECTED AREA. 8 3.1.4 General Emergency (GE) A General Emergency, as defined in section 2.1.4, includes but is not limited to an event or condition that involves: (A) Loss of any two fission product barriers AND loss or potential loss of the third barrier -fuel clad, RCS and/or containment. (B) A precursor event or condition that, unmitigated, may lead to a loss of all three fission product barriers.
Precursor events and conditions of this type include those that lead directly to core damage and loss of containment integrity. ( C) A release of radioactive materials to the environment that could result in doses greater than an EPA PA G at or beyond the site boundary. (D)A HOSTILE ACTION resulting in the loss of key safety functions (reactivity control, core cooling/RPV water level or RCS heat removal) or damage to spent fuel. 3.1.5 Risk-Informed Insights Emergency preparedness is a defense-in-depth measure that is independent of the assessed risk from any particular accident sequence; however, the development of an effective emergency classification scheme can benefit from a review ofrisk-based assessment results. To that end, the development and assignment of certain ICs and EALs also considered insights from several site-specific probabilistic safety assessments.
Some generic insights from this review included:
: 1. Accident sequences involving a prolonged loss of all AC power are significant contributors to core damage frequency at many Boiling Water Reactors (BWRs). For this reason, a loss of all AC power for greater than 15 minutes, with the plant at or above Hot Shutdown, was assigned an ECL of Site Area Emergency.
Precursor events to a loss of all AC power were also included as an Unusual Event and an Alert. A station blackout coping analyses performed in response to 10 CFR 50.63 and Regulatory Guide 1.155, Station Blackout, may be used to determine a time-based criterion to demarcate between a Site Area Emergency and a General Emergency.
The time dimension is critical to a properly anticipatory emergency declaration since the goal is to maximize the time available for State and local officials to develop and implement offsite protective actions. 2. For severe core damage events, uncertainties exist in phenomena important to accident progressions leading to containment failure. Because of these uncertainties, predicting the status of containment integrity may be difficult under severe accident conditions.
This is why maintaining containment integrity alone following sequences leading to severe core damage is an insufficient basis for not escalating to a General Emergency.
: 3. PSAs indicated that leading contributors to latent fatalities were sequences involving a containment bypass, a large Loss of Coolant Accident (LOCA) with early containment failure, a Station Blackout lasting longer than the DAEC coping period, and a reactor coolant pump seal failure. The generic EAL methodology needs to be sufficiently rigorous to address these sequences in a timely fashion. 9 3.2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS The NEI 99-01 methodology makes use of symptom-based, barrier-based and based ICs and EALs. Each type is discussed below. Symptom-based ICs and EALs are parameters or conditions that are measurable over some range using plant instrumentation (e.g., core temperature, reactor coolant level, radiological effluent, etc.). When one or more of these parameters or conditions are normal, reactor operators will implement procedures to identify the probable cause(s) and take corrective action. Fission product barrier-based ICs and EALs are the subset of symptom-based EALs that refer specifically to the level of challenge to the principal barriers against the release of radioactive material from the reactor core to the environment.
These barriers are the fuel cladding, the reactor coolant system pressure boundary, and the containment.
The barrier-based ICs and EALs consider the level of challenge to each individual barrier -potentially lost and lost -and the total number of barriers under challenge.
Event-based I Cs and EALs define a variety of specific occurrences that have potential or actual safety significance.
These include the failure of an automatic reactor scram to shut down the reactor, natural phenomena (e.g., an earthquake), or man-made hazards such as a toxic gas release. 10 3.3 DAEC-SPECIFIC ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION The scheme's generic information is organized by Recognition Category in the following order. R -Abnormal Radiation Levels / Radiological Effluent -Section 6 C -Cold Shutdown I Refueling System Malfunction
-Section 7 E -Independent Spent Fuel Storage Installation (ISFSI) -Section 8 F -Fission Product Barrier -Section 9 H -Hazards and Other Conditions Affecting Plant Safety -Section 10 S -System Malfunction
-Section 11 Each Recognition Category section contains a matrix showing the ICs and their associated emergency classification levels. The following information and guidance is provided for each IC: ECL -the assigned emergency classification level for the IC. Initiating Condition
-provides a summary description of the emergency event or condition.
Operating Mode Applicability
-Lists the modes during which the IC and associated EAL(s) are applicable (i.e., are to be used to classify events or conditions).
Emergency Action Level(s)-Provides examples of reports and indications that are considered to meet the intent of the IC. For Recognition Category F, the fission product barrier thresholds are presented in tables and arranged by fission product barrier and the degree of barrier challenge (i.e., potential loss or loss). This presentation method shows the synergism among the thresholds, and supports accurate assessments.
Basis -Provides background information that explains the intent and application of the IC and EALs. In some cases, the basis also includes relevant source information and references.
11 3 .4 IC AND EAL MODE APPLICABILITY The DAEC emergency classification scheme was developed recognizing that the applicability of ICs and EALs will vary with plant mode. For example, some based ICs and EALs can be assessed only during the power operations, startup, or hot standby/shutdown modes of operation when all fission product barriers are in place, and plant instrumentation and SAFETY SYSTEMS are fully operational.
In the cold shutdown and refueling modes, different symptom-based ICs and EALs will come into play to reflect the opening of systems for routine maintenance, the unavailability of some SAFETY SYSTEM components and the use of alternate instrumentation.
The following table shows which Recognition Categories are applicable in each plant mode. The ICs and EALs for a given Recognition Category are applicable in the indicated modes. MODE APPLICABILITY MATRIX Recognition Category Mode R C E F H s Power Operations X X X X X Startup X X X X X Hot Shutdown X X X X X Cold Shutdown X X X X Refueling X X X X Defueled X X X X DAEC Operating Modes Power Operations (1): Startup (2): Hot Shutdown (3): Cold Shutdown ( 4): Refueling (5): Mode Switch in Run Mode Switch in Startup/Hot Standby or Refuel (with all vessel head closure bolts fully tensioned)
Mode Switch in Shutdown, Average Reactor Coolant Temperature
>212 °F (with all vessel head closure bolts fully tensioned)
Mode Switch in Shutdown, Average Reactor Coolant Temperature~
212 °F (with all vessel head closure bolts fully tensioned)
Mode Switch in Shutdown or Refuel (with one or more vessel head closure bolts less than fully tensioned) 12 4 DEVELOPMENT OF THE DAEC EMERGENCY CLASSIFICATION SCHEME 4.1 GENERAL DEVELOPMENT PROCESS The DAEC ICs and EALs were developed to be unambiguous and readily assessable.
The IC is the fundamental event or condition requiring a declaration.
The EAL(s) is the pre-determined threshold that defines when the IC is met. Useful acronyms and abbreviations associated with the DAEC emergency classification scheme are presented in Appendix A, Acronyms and Abbreviations.
Many words or terms used in the DAEC emergency classification scheme have specific definitions.
These words and terms are identified by being set in all capital letters (i.e., ALL CAPS). The definitions are presented in Appendix B, Definitions.
4.2 CRITICAL CHARACTERISTICS When crafting the scheme, DAEC ensured that certain critical characteristics have been met. These critical characteristics are listed below.
* The ICs, EALs, Operating Mode Applicability criteria, Notes and Basis information are consistent with industry guidance; while the actual wording may be different, the classification intent is maintained.
With respect to Recognition Category F, DAEC includes a user-aid to facilitllte timely and accurate classification of fission product baiTier losses and/or potential losses. The user-aid logic is consistent with the classification logic presented in Section 9.
* The I Cs, EALs, Operating Mode Applicability criteria, Notes and Basis information are technically complete and accurate (i.e., they contain the information necessary to make a correct classification).
* EAL statements use objective criteria and observable values.
* I Cs, EALs, Operating Mode Applicability and Note statements and formatting consider human factors and are user-friendly.
* The scheme facilitates upgrading and downgrading of the emergency classification where necessary.
* The scheme facilitates classification of multiple concurrent events or conditions.
13 4.3 INSTRUMENTATIONUSEDFOREALS DAEC incorporated instrumentation that is reliable and routinely maintained in accordance with site programs and procedures.
Alarms referenced in EAL statements are those that are the most operationally significant for the described event or condition.
EAL setpoints are within the calibrated range of the referenced instrumentation, and consider any automatic instrumentation functions that may impact accurate EAL assessment.
In addition, EAL setpoint values do not use terms such as "off-scale low" or "off-scale high" since that type of reading may not be readily differentiated from an instrument failure. 4.4 EAL/THRESHOLD REFERENCES TO AOP AND EOP SETPOINTS/CRITERIA Some of the criteria/values used in several EALs and fission product barrier thresholds are drawn from DAEC AOPs and EOPs. This approach is intended to maintain good alignment between operational diagnoses and emergency classification assessments.
Appropriate administrative controls are in place to ensure that a subsequent change to an AOP or EOP is screened to determine if an evaluation pursuant to 10 CFR 50.54( q) is required.
14 5 GUIDANCE ON USING THE DAEC EALS 5.1 GENERAL CONSIDERATIONS When making an emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes and the informing Basis information.
In the Recognition Category F matrices, EALs are referred to as Fission Product Barrier Thresholds; the thresholds serve the same function as an EAL. NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, Interim Staff Guidance, Emergency Planning/or Nuclear Power Plants. All emergency classification assessments should be based upon valid indications, reports or conditions.
A valid indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy.
For example, validation could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel.
The validation of indications should be completed in a manner that supports timely emergency declaration.
For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.
A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that 1) the activity proceeds as planned and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component.
In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected.
Events or conditions of this type may be subject to the reporting requirements of 10 CFR 50.72. The assessment of some EALs is based on the results of analyses that are necessary to asce11ain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.); the EAL and/or the associated basis discussion will identify the necessary analysis.
In these cases, the 15-minute declaration 15 period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available).
The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift).
While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary.
This scheme provides the Emergency Director with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to Category H). The Emergency Director will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition.
A similar provision is incorporated into the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier. 5.2 CLASSIFICATION METHODOLOGY To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded.
The evaluation of an EAL(s) must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, then the IC is considered met and the associated ECL is declared in accordance with plant procedures.
When assessing an EAL that specifies a time duration for the off-normal condition, the "clock" for the EAL time duration runs concurrently with the emergency classification process "clock." For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01.
5.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared.
Additionally, there is no "additive" effect from multiple EALs meeting the same ECL. For example: If two Alert EALs are met, an Alert should be declared.
Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events. 16 5.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable.
If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared).
Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition.
For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response.
In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher. 5.5 CLASSIFICATION OF IMMINENT CONDITIONS Although EALs provide specific thresholds, the Emergency Director must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT).
If, in the judgment of the Emergency Director, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met. While applicable to all emergency classification levels, this approach is particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures.
5.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING An ECL may be downgraded when the event or condition that meets the highest IC and EAL no longer exists, and other site-specific downgrading requirements are met. If downgrading the ECL is deemed appropriate, the new ECL would then be based on a lower applicable IC(s) and EAL(s). The ECL may also simply be terminated.
17 The following approach to downgrading or tenninating an ECL is recommended.
ECL Unusual Event Alert Site Area Emergency with no long-term plant damage Site Area Emergency with long-term plant damage General Emergency Action When Condition No Longer Exists Terminate the emergency in accordance with plant procedures.
Downgrade or terminate the emergency in accordance with plant procedures.
Downgrade or terminate the emergency in accordance with plant procedures.
Terminate the emergency and enter recovery in accordance with plant procedures.
Terminate the emergency and enter recovery in accordance with plant procedures.
As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02. 5.7 CLASSIFICATION OF SHORT-LIVED EVENTS As discussed in Section 3 .2, event-based I Cs and EALs define a variety of specific occurrences that have potential or actual safety significance.
By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed.
If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration.
Examples of such events include a failure of the reactor protection system to automatically scram the reactor followed by a successful manual scram or an earthquake.
5.8 CLASSIFICATION OF TRANSIENT CONDITIONS Many of the I Cs and/or EALs contained in this document employ time-based criteria.
These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted.
In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes).
The following guidance should be applied to the classification of these conditions.
EAL momentarily met during expected plant response -In instances where an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are perfonned in accordance with procedures.
18 EAL momentarily met but the condition is corrected prior to an emergency declaration
-If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required.
For illustrative purposes, consider the following example. An ATWS occurs and the auxiliary feedwater system fails to automatically start. Steam generator levels rapidly decrease and the plant enters an inadequate RCS heat removal condition (a potential loss of both the fuel clad and RCS barriers).
If an operator manually starts the auxiliary feedwater system in accordance with an EOP step and clears the inadequate RCS heat removal condition prior to an emergency declaration, then the classification should be based on the ATWS only. It is important to stress that the 15-minute emergency classification assessment period is not a "grace period" during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event; emergency classification assessments must be deliberate and timely, with no undue delays. The provision discussed above addresses only those rapidly evolving situations where an operator is able to take a successful corrective action prior to the Emergency Director completing the review and steps necessary to make the emergency declaration.
This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.
5.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition.
This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery.
This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process. In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 is applicable.
Specifically, the event should be reported to the NRC in accordance with 10 CFR 50.72 within one hour of the discovery of the undeclared event or condition.
The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements.
5 .10 RETRACTION OF AN EMERGENCY DECLARATION Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022.
19 6 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS 20 RU1 ECL: Notification of Unusual Event Initiating Condition:
Release of gaseous or liquid rad i oactivity greater than 2 times the ODAM limits for 60 minutes or longer. Operating Mode Applicability:
All Emergency Action Levels: Notes:
* The Emergency Director shou ld declare the event promptly upon determining that the applicable time ha s been exceeded, or w ill lik e l y be exceeded.
* If an ongo in g r e l ease is detected an d the r e l ease start time i s unknown , assume that the release duration has exceeded the specified time limit.
* If the effluent flow past an effl u ent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no l onger va lid for classification purposes.
RUI .1 Reading on ANY Tab l e R-1 effluent radiation monitor greater than column " NOUE" for 60 minutes or lon ger: RUl.2 RUl.3 Table R-1 -Effluent Monitor Classification Thresholds
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V) ::l 0 (l) V) ro l!J Monitor Rea cto r Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8) Turbine Building ventilation rad monit or (Kaman 1/2) Offga s Stack rad monitor (Kaman 9/10) LLRPSF rad monitor (Kam an 12) GSW r a d monitor (RIS-4767) RHRSW & ESW rad monitor (RM-199 7) RHRSW & ESW Rupture Di sc rad monitor (RM-4268) NOUE 8.0E-04 uci/cc 8.0E-04 uci/cc 2.0E-01 uci/cc l.2E-03 uci/cc 1.SE+03 cps 8.4E+02 cps l.OE+03 cps Reading on ANY effluent.radiatio n monitor g reater than 2 times the alarm setpoint established by a current radioactivity dischar ge permit for 60 minutes or l onger. Sample ana l ysis for a gaseous or liquid release indi cates a conce ntra tion or release rate greater than 2 times the ODAM limits for 60 minute s or lon ger. 21 Definitions:
None Basis: -----------
-------------------------Th i s IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radio lo gical release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release).
It includes any gaseous or liquid radiological release , monitored or monitored, including those for which a radioactivity discharge permit is normally prepared.
DAEC incorporates design features intended to control the release of radioactive effluents to the environment.
Further, there are administrat i ve controls established to prevent unintentional releases , and to control and monitor intentional releases. The occurrence of an extended , uncontrolled radioactive re l ease to the env ironm ent i s indicative of degradation in these features and/or controls.
Radiological effluent EALs are also included to provide a basi s for classifyin g events and cond iti ons that cannot be readi l y or appropr i ate l y class ifi ed on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
Classification based on effluent monitor readin gs assumes that a release path to the environment is estab li shed. If the effluent flow past an effluent monitor i s known to have stopped due to actions to iso l ate the re l ease path , then the effluent monitor reading is no longer valid for classification purposes.
Releases should n ot be prorated or averaged.
For example, a release exceeding 4 t im es release limits for 30 minutes does not meet the EAL. EAL RUl .1 -This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or l iquid effluent pathways. EAL RUl .2 -Th i s EAL addresses radioactiv it y releases that cause effl u ent radiation monitor readings to exceed 2 times the limit estab li s h ed b y a radioactivity discharge permit. This EAL will typically be associated wit h planned batch re l eases fro m non-continuou s release pathwa ys (e.g., radwaste, waste gas). EAL RU 1.3 -This EAL addresses uncontro ll ed gaseous or liquid relea ses that are detected by sa mple analysis or environmental surveys , particul arly on unmonitored pathways (e.g., spi ll s of radioactive liquids into storm drains , heat exchanger l eakage in river water systems, etc.). Esca lati on of the emergency classification l eve l wou ld be via IC RA 1. 22 ECL: Notificat ion of Un u s ual Eve nt Initiating Condition:
UNPLANNED l oss of water l eve l a bo ve irradi ated fue l. Operating Mode Applicability:
A ll Emergency Action Levels: RU2 RU2.l a. UNPLANNED water level d rop in the REFUELING PATHWAY as indicated b y ANY of the following:
* Report to control room (v i sua l observat i on)
* F uel pool l evel indication (LI-3413) l ess than 36 feet an d lo we rin g
* WR GEMAC F l ood up indication (LI-4541) com in g on sca l e AND b. UNPLANNED rise in area radiation l eve l s as indicated by ANY of the following radiation monitors.
* Spen t F uel Pool Area, R I-9178
* North Refuel Floor, RI-9163
* New F u e l Vault Area , RI-9153
* Sout h Refuel Floor, RI-9164
* NW Drywell Area Hi Range Rad Monitor , RIM-9184A
* South Dr ywe ll Area H i Range Rad Monitor , RIM-9184B Definitions:
UNPLANNED:
A p arameter change or an eve nt that is not 1) the result of an intended evolut i on or 2) an expected plant re spo nse to a transient.
The cause of the parameter change or event may be known or unknown. REFUELING PATHWAY: T h e r eactor r efuel in g cavity , spent fue l pool and fuel transfer canal. 23 Basis: This IC addresses a decrease in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant. A water level decrease will be primarily determined by indications from available l evel instrumentation.
Other sources of level indications may include reports from plant personnel ( e.g., from a refueling crew) or video camera observations.
A significant drop in the water level may also cause an increase in the radiation levels of adjacent areas that can be detected by monitors in those locations. The effects of planned evolutions should be considered.
For example , a refueling br i dge area radiation monitor reading may increase due to planned evolutions such as lifting of the reactor vessel head or movement of a fuel assembly.
Note that this EAL is applicable only in cases where the elevated reading is due to an UNPLANNED loss of water level. During preparation for reactor cavity flood up prior to entry into refuel mode, reactor vessel leve l instrument LI-4541 (WR GEMAC , FLOODUP) on control room panel 1C04 is placed in service by I&C personnel connecting a compensating air signal after the reference leg is disconnected from the reactor head. Normal refuel water level is above the top of the span of this flood up level indicator.
A va l id indication (e.g., not due to loss of compensating air signal or other instrument channel failure) of reactor cavity level coming on span for this instrument is used at DAEC as an indicator of uncontrolled reactor cavity level decrease.
DAEC Technica l Specifications require a minimum of 36 feet of water in the spent fuel pool when moving irradiated fuel into the secondary containment.
During refueling, the gates between the reactor cavity and the refueling cavity are removed and the spent fuel pool level indicator LI-3413 is used to monitor refueling water level. Procedures require that a normal refueling water level be maintained at 37 feet 5 inches. A low level alarm actuates when spent fuel pool level drops below 37 feet 1 inch. Symptoms of inventory loss at DAEC include visual observation of decreasing water levels in reactor cavity or spent fuel storage pool , Reactor Building (RB) fuel storage pool radiation monitor or refueling area radiation monitor alarms, observation of a decreasing trend on the spent fuel pool water level indicator , and actuation of the spent fuel pool low water level alarm. To eliminate minor level perturbations from concern , DAEC uses LI-3413 indicated water level below 36 feet and lowering. [ncreased radiation levels can be detected by the local area radiation monitors surrounding the spent fuel pool and refueling cavity areas. Applicable area radiation monitors are those listed in AOP 981. A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes. Escalation of the emergency classification level would be via IC RA2. 24 
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RA1 ECL: Alert Initiating Condition:
Release of gaseous or liquid radioactivity resulting in offsite dose greate r than 10 mrem TEDE or 50 mrem thyroid CDE. Operating Mode Applicability:
All Emergency Action Levels: Notes:
* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time i s unknown , assume that the release duration has exceeded the specified time limit.
* If the effluent flow past an effluent monitor is known to have stopped due to actions to i so late the release path, then the effluent monitor reading i s no l onger valid for classification purposes.
* The pre-calculated effluent monitor values presented in EAL 1.1 should only be used for emergency classification assessments until the results from a dose assessment u sing actual meteorology are available.
RAl.1 RAl.2 RAl.3 RAl.4 Reading on ANY Table R-1 effluent radiation monitor greater than column "A lert" for 15 minutes or longer: Table R-1-Effluent Monitor Classification Thresholds
----------------
------V, ::::, 0 QJ V, n, (.!J Monitor Alert Reactor Building ventilation rad monitor (Kaman 3/4 , 5/6 , 7 /8) Turbine Building ventilation rad monitor (Kaman 1/2) Offgas Sta c k rad monitor (K aman 9/10) LLR PS F rad monitor (Kaman 12) GSW rad monitor (RIS-4767) RHRSW & ESW rad monitor (RM-1997) RHRSW & ESW Rupture Disc rad monitor (RM-4268) 1.lE-02 uci/cc 1.4E-02 uci/cc 4.SE+Ol uci/cc 1.4E-02 uci/cc 1.7E+04 cps 1.2E+04 cps 1.8E+04 cps Do se assessment using actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond SITE BOUNDARY.
[Preferr ed] Analysis of a liquid effluent sample indicates a concentration or release rate that would result in do ses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for one hour of expos ure. Field survey results indicate EITHER of the fo llowin g at or beyond the SITE BOUNDARY:
* Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer.
* Analyses of field survey samples indicat e thyroid CDE g reater than 50 mrem for one hour of inhalation.
25 Definitions:
SITE BOUNDARY: That line beyond which the land is neither owned , nor leased , nor otherwise controlled by the licensee.
Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1 % of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.
Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).
This IC is modified by a note that EAL RAl .1 is only assessed for emergency classification until a qualified dose assessor is performing assessments using dose projection software incorporating actual meteorological data and current radiological conditions.
Radiological effluent EAL s are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 1% of the EPA PAG of 1 , 000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established.
If the effluent flow past an effluent monitor is known to have stopped due to action s to isolate the release path , then the effluent monitor reading is no longer valid for classification purposes.
Escalation of the emergency classification level would be via IC RS 1. 26 RA2 ECL: Alert Initiating Condition:
Significant lowering of water level above , or damage to, irradiated fuel. Operating Mode Applicability:
All Emergency Action Levels: RA2.l RA2.2 RA2.3 Uncovery of irradiated fuel in the REFUELING PATHWAY. Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by Hi Rad alarm for ANY of the following ARMs:
* Spent Fuel Pool Area, Rl-9178
* North Refuel Floor , Rl-9163
* New Fuel Vault Area , Rl-9153
* South Refuel Floor , Rl-9164 OR Reading greater than 5 R/hr on ANY of the following radiation monitors (in Mode 5 only):
* NW Drywell Area Hi Range Rad Monitor , RIM-9184A
* South Drywell Area Hi Range Rad Monitor , RIM-9184B Lowering of spent fuel pool level to 25.17 feet. Definitions:
REFUELING PATHWAY -The reactor refueling cavity , spent fuel pool and fuel transfer canal. Basis: This IC addresses events that have caused IMMIN E NT or actual damage to an irradiated fuel assembly , or a significant lowering of water level within the spent fuel pool. These events present radiological safety challen g es to plant personnel and are precursors to a release of radioactivity to the environment.
As such , they represent an actual or potential substantial degradation of the level of safety of the plant. Expected radiation monitor alarm(s) during preplanned transfer of highly radioactive material through the affected areas are not considered valid alarms for the purpose of comparison to these EALs. 27 This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed , damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E-HUl. Escalation of the emergency would be based on either Recognition Category R or C !Cs. EALRA2.l This EAL escalates from RU2 in that the loss of level , in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels , or other plant parameters.
Computational aids may also be used. Classification of an event using this EAL should be based on the totality of available indications , reports, and observations.
While an area radiation monitor could detect an increase in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered.
To the degree possible , readings should be considered in combination with other available indications of inventory loss. A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes. EALRA2.2 This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping , bumping or binding of an assembly, or dropping a heavy load onto an assembly.
An alarm on these radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event ( e.g., a fuel handling accident).
Threshold values for the Drywell monitors are only applicable in Mode 5 since the calculated radiation levels from damage to irradiated fuel would be masked by the typical background levels on these monitors during plant operation, and mechanical damage to a fuel assembly in the vessel can only happen with the reactor head removed. EALRA2.3 Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool. Escalation of the emergency classification level would be via ICs RSI or RS2. 28 RA3 ECL: Alert Initiating Condition:
Radiation levels that impede access to areas necessary for normal plant operation.
Operating Mode Applicability:
All Emergency Action Levels: RA3.1 Dose rate greater than 15 mR/hr in ANY of the following areas:
* Control Room (RM-9162)
* Central Alarm Station (by survey) Definitions:
None Basis: This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown.
As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the increased radiation levels and determine if another IC may be applicable.
Escalation of the emergency classification level would be via Recognition Category R , C or FI Cs. 29 RS1 ECL: Site Area Emergency Initiating Condition:
Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE. Operating Mode Applicability:
All Emergency Action Levels: Notes:
* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown , assume that the release duration has exceeded the specified time limit.
* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the relea se path , then the effluent monitor reading is no longer valid for classification purposes.
* The pre-calculated effluent monitor values presented in EAL 1.1 should only be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
RS 1.1 Reading on ANY Table R-1 effluent radiation monitor greater than column "SAE" for 15 minutes or longer: RSl.2 RSl.3 Ill ::::, 0 (lJ Ill n, l9 Monitor Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7 /8) Turbine Building ventilation rad monitor (Kaman 1/2) Offgas Stack rad monitor (Kaman 9/10) Dos e assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid C DE at or beyond the SITE BOUNDARY. [Preferred]
Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
* Closed window dose rates grea ter than 100 mR/hr expected to continue for 60 minutes or longer.
* Analyses of field survey samples indicate thyroid CDE g reater than 500 mrem for one hour of inhalation.
30 Definitions:
SITE BOUNDARY: That line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PA Gs). It includes both monitored and un-monitored releases.
Releases of.this magnitude are associated with the failure of plant systems needed for the protection of the public. This IC is modified by a note that EAL RS 1.1 is only assessed for emergency classification until a qualified dose assessor is performing assessments using dose projection software incorporating actual meteorological data and current radiological conditions.
However , if Kaman monitor readings are sustained for 15 minutes or longer and the required MIDAS dose assessments cannot be completed within this period, then the declaration can be made using Kaman readings PROVIDED the readings are not from an isolated flow path. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 10% of the EPA PAG of 1 , 000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established.
If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. If Kaman readings are not valid , field survey results may be utilized to assess this IC using EAL RSI.3. Escalation of the emergency classification level would be via IC RG 1. 31 ECL: Site Area Emergency Initiating Condition:
Spent fue l pool level at 16.36 feet. Operating Mode Applicability:
All Emergency Action Levels: RS2.1 Lowering of spent fuel pool level to 16.36 feet. Definitions:
None Basis: RS2 This IC addresses a significant l oss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failure s of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.
It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however , it is included to provide c l assification diversity. Escalation of the emergency class ifi cation level would be via IC RGl or RG2. 32 RG1 ECL: General Emergency Initiating Condition:
Release of gaseo u s radioactivity resulting in offsite dose greater than 1 , 000 mrem TEDE or 5,000 mrem thyroid CDE. Operating Mode Applicability:
All Emergency Action Level s: Notes:
* The Emerge ncy Director shou ld declare the event promptly upon determining that the applicable time ha s been exceeded, or will lik ely be exceeded.
* If an ongo in g release is detected and the release sta rt tim e is unknown , assume that the release duration has exceeded the spec ifi ed time limit.
* If the efflue nt flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then t he effl uent monitor reading is no l onger valid for c la ssification purposes.
* The pre-calculated efflue nt monitor values presented in EAL 1.1 s hould on l y be u sed for emergency classification assessments until the results from a dose assessment u s in g actual met eorology are available.
RG 1.1 Reading on ANY Tab l e R-1 effl u e nt radiation monitor greater than column " GE" for 15 minutes or lon ger: RGl.2 RGl.3 Mon i tor Reactor Building ventilation r ad monitor (K ama n 3/4, 5/6, 7 /8) :l Turbine Building vent il ation r ad mon it o r (Kam a n 1/2) V) "' l9 GE 1.1E+OO uci/cc 1.4E+OO uci/cc Dose assessment using actual meteorology indicates doses greater than 1 , 000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the SITE BOUNDARY.
[Preferred]
Field survey results indicate EITHER of the fo llowin g at or beyond the SITE BOUNDARY:
* Closed window dose rates greater than 1 , 000 mR/hr expected to cont inu e for 60 minutes or longer.
* Analyses of field s urve y samp le s indic ate thyroid CDE greater than 5 , 000 mrem for one hour of inh a l ation. 33 Definitions:
SITE BOUNDARY:
That line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.
Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.
Releases of this magnitude will require implementation of protective actions for the public. This IC is modified by a note that EAL RG 1.1 is only assessed for emergency classification until a qualified dose assessor is performing assessments using dose projection software incorporating actual meteorological data and current radiological conditions.
However, if Kaman monitor readings are sustained for 15 minutes or longer and the required MIDAS dose assessments cannot be completed within this period, then the declaration can be made using Kaman readings PROVIDED the readings are not from an isolated flow path. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established.
If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
If Kaman readings are not valid, field survey results may be utilized to assess this IC using EAL RG 1.3. 34 *--_j l______ __ RG2 ECL: General Emergency Initiating Condition:
Spent fuel pool level cannot be restored to at least 16.36 feet for 60 minutes or longer. Operating Mode Applicability:
All Emergency Action Levels: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
RG2.1 Spent fuel pool level cannot be restored to at least 16.36 feet for 60 minutes or longer. Definitions:
None Basis: This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment.
It is recognized that this IC would likely not be met until well after another General Emergency IC was met; however, it is included to provide classification diversity.
35 7 COLD SHUTDOWN/
REFUELING SYSTEM MALFUNCTION ICS/EALS 36 ECL: Notification of Unusual Event Initiating Condition:
UNPLANNED loss ofRPV inventory for 15 minutes or longer. Operating Mode Applicability:
4, 5 Emergency Action Levels: CU1 Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
CUI.I CUl.2 UNPLANNED loss of reactor coolant results in RPV level less than a required lower limit for 15 minutes or longer. a. RPV level cannot be monitored.
AND b. UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool. Definitions:
UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. Basis: This IC addresses the inability to restore and maintain water level to a required minimum level ( or the lower limit of a level band), or a loss of the ability to monitor RPV level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant. Refueling evolutions that decrease RCS water inventory are carefully planned and controlled.
An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered. EAL CUI.I recognizes that the minimum required RPV level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented.
This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document.
The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level. 37 EAL CUI .2 addresses a condition where all means to determine RPV level have been lost. If all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RSC inventory loss was occurring by observing sump and Suppression Pool level changes. The drywell floor and equipment drain sumps, reactor building equipment and floor drain sumps receive all liquid waste from floor and equipment drains inside the primary containment and reactor building.
A rise in Suppression Pool water level may be indicative of valve misalignment or leakage in systems that discharge to the Torus. Sump and Suppression Pool level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CAI or CA3. 38 CU2 ECL: Notification of Unusual Event Initiating Condition:
Loss of all but one AC power source to essential buses for 15 minutes or longer. Operating Mode Applicability:
4, 5, Defueled Emergency Action Levels: Note: The Emergency Director should declare the event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
CU2.l a. AC power capability to 1A3 and 1A4 buses is reduced to a single power source for 15 minutes or longer. AND b. Any additional single power source failure will result in loss of ALL AC power to SAFETY SYSTEMS. Definitions:
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
Basis: This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.
When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant. An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.
* A loss of all offsite power with a concurrent failure of aH but one emergency power source (e.g., an onsite diesel generator).
* A loss of emergency power sources (e.g., onsite diesel generators) with a single train of essential buses being fed from an offsite power source. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2. 39 ECL: Notification of Unusual Event Initiating Condition:
UNPLANNED increase in RCS temperature.
Operating Mode Applicability:
4, 5 Emergency Action Levels: CU3 Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
CU3.l CU3.2 UNPLANNED increase in RCS temperature to greater than 2I2°F. Loss of ALL RCS temperature and RPV level indication for 15 minutes or longer. Definitions:
UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.
For DAEC, this is considered to be Secondary Containment as required by Technical Specifications.
Basis: This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to determine RCS temperature and level, represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to IC CA3. A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.
EAL CU3.1 involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications.
During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.
During an outage, the level in the reactor vessel will normally be maintained above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled.
A loss of forced decay heat removal at reduced inventory may result in a rapid increase in reactor coolant temperature depending on the time after shutdown.
40 EAL CU3.2 reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.
Frfteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
* Escalation to Alert would be via IC CAl based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.
41 ECL: Notification of Unusual Event Initiating Condition:
Loss of Vital DC power for 15 minutes or longer. Operating Mode Applicability:
4, 5 Emergency Action Levels: CU4 Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
CU4.l Indicated voltage is less than 105 VDC on BOTH Div 1 and Div 2 125 VDC buses for 15 minutes or longer. Definitions:
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
Basis: This IC addresses a loss of Vital DC power which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode. In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions increase the time available to restore a vital DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant. As used in this EAL, "required" means the Vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment.
For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train Bis in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification.
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Depending upon the event, escalation of the emergency classification level would be via IC CAI or CA3, or an IC in Recognition Category R. 42 ECL: Notification of Unusual Event Initiating Condition:
Loss of all onsite or offsite communications capabilities.
Operating Mode Applicability:
4, 5, Defueled Emergency Action Levels: CU5.l Loss of ALL of the following onsite communication methods:
* Plant Operations Radio System
* In-Plant Phone System
* Plant Paging System (Gaitronics)
CU5 CU5.2 Loss of ALL of the following offsite response organization communications methods:
* DAEC All-Call phone
* All telephone lines (PBX and commercial)
* Cell Phones (including fixed cell phone system)
* Control Room fixed satellite phone system
* FTS Phone system CU5.3 Loss of ALL of the followingNRC communications methods:
* FTS Phone system
* All telephone lines (PBX and commercial)
* Cell Phones (including fixed cell phone system)
* Control Room fixed satellite phone system Basis: This IC addresses a significant loss of on-site or offsite communications capabilities.
While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to offsite response organizations and the NRC. This IC should be assessed only when extraordinary means are being utilized to make communications possible ( e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.). EAL CU5.l addresses a total loss of the communications methods used in supp011 of routine plant operations.
43 EAL CU5.2 addresses a total loss of the communications methods used to notify all offsite response organizations of an emergency declaration.
The offsite response organizations referred to here are the State of Iowa, Linn County, and Benton County. EAL CU 5 .3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
44 ECL: Alert Initiating Condition:
Loss ofRPV inventory.
Operating Mode Applicability:
4, 5 Emergency Action Levels: CA1 Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
CAl.1 CAl.2 Loss ofRPV inventory as indicated by level less than 119.5 inches. a. RPV level cannot be monitored for 15 minutes or longer AND b. UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool due to a loss ofRPV inventory.
Definitions:
UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. Basis: This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier).
This condition represents a potential substantial reduction in the level of plant safety. For EAL CAl.1, a lowering of water level below 119.5 inches indicates that operator actions have not been successful in restoring and maintaining RPV water level. The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery.
Although related, EAL CA 1.1 is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal ( e.g., loss of a Residual Heat Removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3. For EAL CAl.2, the inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation.
If water level cannot be monitored, the operators would need to determine that RCS inventory loss was occurring by observing sump and Suppression Pool level changes. The drywell floor and equipment drain sumps, reactor building equipment and floor drain sumps receive all liquid waste from floor and equipment drains inside the primary containment and reactor building.
A rise in Suppression Pool water level may be indicative of valve misalignment or leakage in systems that discharge to the Torus. Sump and Suppression Pool level increases must be 45 evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS 1 If the RPV inventory level continues to lower, then escalation to Site Area Emergency would be via IC CSL 46 ECL: Alert Initiating Condition:
Loss of all offsite and all onsite AC power to essential buses for 15 minutes or longer. Operating Mode Applicability:
4, 5, Defueled Emergency Action Levels: CA2 Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
CA2.l Loss of ALL offsite and ALL onsite AC Power to 1A3 and 1A4 buses for 15 minutes or longer. Definitions:
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
Basis: This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore an emergency bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via IC CS 1 or RS 1. 47 ECL: Alert Initiating Condition:
Inability to maintain the plant in cold shutdown.
Operating Mode Applicability:
4, 5 Emergency Action Levels: CA3 Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
CA3.l CA3.2 UNPLANNED increase in RCS temperature to greater than 212°F for greater than the duration specified in Table C-2. Table C-2 RCS Heat-up Duration Thresholds RCS Integrity CONTAINMENT CLOSURE Heat-up Duration Status Intact Not applicable 60 minutes* Not intact Established 20 minutes* Not Established 0 minutes
* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.
UNPLANNED RCS pressure increase greater than 10 psig due to a loss of RCS cooling. Definitions:
UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.
For DAEC, this is considered to be Secondary Containment as required by Technical Specifications.
Basis: This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant. A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.
RCS integrity is intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams). 48 The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact. The 20-minute criterion was included to allow time for operator action to address the temperature increase.
The RCS Heat-up Duration Thresholds table also addresses an increase in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature increase without a substantial degradation in plant safety. Finally, in the case where there is an increase in RCS temperature, the RCS is not intact, and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes).
This is because 1) the evaporated reactor coolant may be released directly into the Containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel. EAL CA3.2 provides a pressure-based indication of RCS heat-up. Escalation of the emergency classification level would be via IC CSl or RSl. 49 CA6 ECL: Alert Initiating Condition:
Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. Operating Mode Applicability:
4, 5 Emergency Action Levels: Notes: CA6.1
* If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then this emergency classification is not warranted.
* If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.
: a. AND b. The occurrence of ANY of the Table C-3 hazardous events: 1.
* Seismic event (earthquake)
* Internal or external flooding event
* High winds or tornado strike
* FIRE
* EXPLOSION
* Other events with similar hazard characteristics as determined by the Shift Manager or Emergency Director Event damage has caused indications of degraded performance in one train of a SAFETY SYSTEM needed for the current operating mode. AND 2. EITHER of the following:
* Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode, OR
* The event has resulted in VISIBLE DAMAGE to the second train of a SAFETY SYSTEM needed for the current operating mode. 50 Definitions:
FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
EXPLOSION:
A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction, or overpressurization.
A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion.
Such events may require a post-event inspection to determine if the attributes of an explosion are present. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis.
The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.
Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not VISIBLE DAMAGE. Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of perfonnance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance for criteria CA6.l.b.1 of this EAL; commercial nuclear power plants are designed to be able to support single systerri issues without compromising public health and safety from radiological events. An event affecting a single-train SAFETY SYSTEM (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the one train) would not be classified under CA6 because the two-train impact criteria that underlie the EALs and Bases would not be met. If an event affects a single-train SAFETY SYSTEM, then the emergency classification should be made based on plant parameters/symptoms meeting the EALs for another IC. Depending upon the circumstances, classification may also occur based on Shift Manager/Emergency Director judgement.
Indications of degraded perfonnance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available.
The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. 51 VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make this determination based on the totality of available event and damage report information.
This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. Escalation of the emergency classification level would be via IC RS I. 52 CS1 ECL: Site Area Emergency Initiating Condition:
Loss of RPV inventory affecting core decay heat removal capability.
Operating Mode Applicability:
4, 5 Emergency Action Levels: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
CSl.1 CSl.2 CSl.3 a. CONTAINMENT CLOSURE not established.
AND b. RPV level less than +64 inches a. CONTAINMENT CLOSURE established.
AND b. RPV level less than + 15 inches a. RPV level cannot be monitored for 30 minutes or longer. AND b. Core uncovery is indicated by EITHER of the following:
* Drywell Monitor (9184A/B) reading greater than 5.0 R/hr
* UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool of sufficient magnitude to indicate core uncovery Definitions:
CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.
For DAEC, this is considered to be Secondary Containment as required by Technical Specifications.
UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. 53 Basis: This IC addresses a significant and prolonged loss of RPV inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.
Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If reactor vessel level cannot be restored, fuel damage is probable.
Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions.
The difference in the specified reactor vessel levels of EALs CS 1.1.b and CS 1.2.b reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.
In the Cold Shutdown and Refueling Modes, LT/LI-4559, 4560, and 4561 (RX VESSEL NARROW RANGE LEVEL) instruments read up to 22" high due to hot calibrations.
LI-4541 (WR GEMAC, FLOODUP) should be used in these Modes for comparison to EAL thresholds since it is calibrated cold and reads accurately.
If normal means ofRPV level indication are not available due to plant evolutions, redundant means of RPV level indication will be normally installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted.
In EAL CS 1.3 .a, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties).
It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.
The inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation.
If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV. These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.
Escalation of the emergency classification level would be via IC CGl or RGI. 54 
----------------------------*---
CG1 ECL: General Emergency Initiating Condition:
Loss ofRPV inventory affecting fuel clad integrity with containment challenged.
Operating Mode Applicability:
4, 5 Emergency Action Levels: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
CGl.1 CGl.2 a. RPV level less than + 15 inches for 3 0 minutes or longer. AND b. ANY indication from the Secondary Containment Challenge Table C-1. a. RPV level cannot be monitored for 30 minutes or longer. AND b. Core uncovery is indicated by EIHER of the following:
* Drywell Monitor (9184A/B) reading greater than 5.0 R/hr.
* UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool of sufficient magnitude to indicate core uncovery.
AND c. ANY indication from the Secondary Containment Ch_allenge Table C-1. Table C-1 Secondary Containment Challenge
* CONTAINMENT CLOSURE not established*
* Drywell Hydrogen or Torus Hydrogen greater than 6% AND Drywell Oxygen or Torus Oxygen greater than 5%
* UNPLANNED increase in containment pressure
* Secondary containment radiation monitors above max safe operating limits (MSOL) of EOP 3, Table 6
* If CONTAINMENT CLOSURE is re-established prior to exceeding the 30 minute time limit, then declaration of a General Emergency is not required.
55 Definitions:
CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.
For DAEC, this is considered to be Secondary Containment as required by Technical Specifications.
UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. Basis: This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged.
This condition represents actual or IMMINENT substantial core degradation or melting with pote11tial for loss of containment integrity.
Releases can be reasonably expected to exceed EPA P AG exposure levels off site for more than the immediate site area. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If reactor vessel level cannot be restored, fuel damage is probable.
With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment.
If CONTAINMENT CLOSURE is established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.
The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen bum (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity.
It therefore represents a challenge to Containment integrity.
In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment.
If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged.
In EAL CG 1.2.a, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties).
It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.
For EAL CGl.2.b, the calculated radiation level on the Drywell Monitors (9184A/B) is without the reactor head in place. Calculated in radiation levels with the reactor head in place are below the normal variation in background readings of these monitors.
56 The inability to monitor RPV level may be caused by instrumentation and/or power failures or water level dropping below the range of available instrumentation.
If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV. For the Containment Challenge Table, Secondary Containment max safe operating (MSOL) limits from EOP 3 are defined as the highest parameter value at which neither: (1) equipment necessary for the safe shutdown of the plant will fail nor (2) personnel access necessary for the safe shutdown of the plant will be precluded.
These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.
57 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI} ICS/EALS 58 ECL: Notification of Unusual Event Initiating Condition:
Damage to a loaded cask CONFINEMENT BOUNDARY.
Operating Mode Applicability:
All Emergency Action Levels: E-HU1 E-HUl .1 Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by a radiation reading greater than the values shown on Table E-1 on the spent fuel cask. Table E-1 Cask Dose Rates 61BTDSC 3 feet from HSM Surface 800 mrem/hr Outside HSM Door-Centerline of DSC 200 mrem/hr End Shield Wall Exterior 40 mrem/hr Definition:
CONFINEMENT BOUNDARY:
The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. Basis: This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage. The existence of "damage" is determined by radiological survey. The technical specification multiple of "2 times", which is also used in Recognition Category RIC RUl, is used here to distinguish between non-emergency and emergency conditions.
The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask. Security-related events for ISFSis are covered under I Cs HUl and HAl. 59 9 FISSION PRODUCT BARRIER ICS/EALS FA1 / L p L p 'L p RCS CTMT / 2/3-FS1 L p L p L p RCS CTMT FG1 60 FAlALERT ANY Loss OR ANY Potential Loss of EITHER the Fuel Clad OR RCS barrier. ... *:::: .. ,, ,. Fuel ([;:lad. Barrier ... ;:,.* Table F-1: DAEC EAL Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Barriers *-,i, FSl SITE AREA EMERGENCY Loss OR Potential Loss of ANY two barriers.
0 erating Mode A licability:
1, 2, 3 : *RCS Barrier :1 -: , .. _, ...*. FGlGENERALEMERGENCY
: Contaitimenf Bartier. '" ! " LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS 1. RCS Activity 1. Primary Containment Conditions
: 1. Primary Containment Conditions A. Coolant activity Not Applicable A. Primary Not Applicable A. UNPLANNED A. Torus pressure greater than 3 00 containment rapid drop in greater than 5 3 µCi/gm dose pressure greater Drywell pressure ps1g equivalent I-131. than 2 psig due to following Drywell OR RCS leakage. pressure rise B. Drywell or Torus OR H2 cannot be B. Drywell pressure determined to be response not less than 6% and consistent with Drywell OR Torus LOCA conditions.
02 cannot be OR determined to be C. UNISOLABLE less than 5% direct downstream OR pathway to the C. HCL (Graph 4 of environment exists EOP 2) exceeded.
after primary containment isolation signal OR D. Intentional primary containment venting per EOPs 61 Fuel Clad Barrier RCS Barrier Containment Barrier ,, LOSS I POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS 2. RPV Water Level 2. RPV Water Level 2. RPV Water Level A. SAG entry is A. RPV water level A. RPV water level Not Applicable Not Applicable A. SAG entry is required cannot be restored cannot be restored required and maintained and maintained above + 15 inches above + 15 inches OR cannot be OR cannot be determined.
determined.
: 3. Not Applicable
: 3. RCS Leak Rate 3. Primary Containment Isolation Failure Not Applicable Not Applicable A. UNISOLABLE A. UNISOLABLE A. UNISOLABLE Not Applicable break in Main primary system primary system Steam, HPCI, leakage that leakage that Feedwater, results in results in RWCU, or RCIC exceeding the exceeding the as indicated by the Max Normal Max Safe failure of both Operating Limit Operating Limit isolation valves in (MNOL) of EOP (MSOL) of EOP ANY one line to 3, Table 6 for 3, Table 6 for close AND EITHER of the EITHER of the EITHER: following:
following:
* HighMSL
* Temperature
* Temperature flow or steam OR OR tunnel
* Radiation
* Radiation Level temperature Level annunciators OR
* Direct report of steam release OR B. Emergency RPV Depressurization required.
62 Fuel Clad Barrier RCS Barrier Containment Barrier LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS 4. Primary Containment Radiation
: 4. Primary Containment Radiation
: 4. Primary Containment Radiation A. Drywell Monitor Not Applicable A. Drywell Monitor Not Applicable Not Applicable A. Drywell Monitor (9184A/B)
(9184A/B)
(9184A/B) reading greater reading greater reading greater than 2000 R/hr. than 5 R/hr after than 5000 R/hr. OR reactor shutdown OR B. Torus Monitor B. Torus Monitor (9185A/B)
(9185A/B) reading greater reading greater than 200 R/hr than 500 R/hr 5. Other Indications
: 5. Other Indications
: 5. Other Indications A. Fuel damage Not Applicable Not Applicable Not Applicable Not Applicable A. Fuel damage assessment assessment indicates at least indicates at least 5% fuel clad 20% fuel clad damage. damage. 6. Emergency Director Judgment 6. Emergency Director Judgment 6. Emergency Director Judgment A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in the opinion of the the opinion of the the opinion of the the opinion of the the opinion of the the opinion of the Emergency Emergency Emergency Emergency Emergency Emergency Director that Director that Director that Director that Director that Director that indicates Loss of indicates Potential indicates Loss of indicates Potential indicates Loss of indicates Potential the Fuel Clad Loss of the Fuel the RCS Barrier. Loss of the RCS the Containment Loss of the Barrier. Clad Barrier. Barrier. Barrier. Containment Barrier. 63 Basis Information For DAEC EAL Fission Product Barrier Table F-1 DAEC FUEL CLAD BARRIER THRESHOLDS:
The Fuel Clad barrier consists of the zircalloy or stainless steel fuel bundle tubes that contain the fuel pellets. 1. RCS Activity Loss I.A This threshold indicates that RCS radioactivity concentration is greater than 300 µCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier. It is recognized that sample collection and analysis ofreactor coolant with highly elevated activity levels could require several hours to complete.
Nonetheless, a sample-related threshold is included as a backup to other indications.
There is no Potential Loss threshold associated with RCS Activity.
: 2. RPV Water Level Loss 2.A The Loss threshold represents any EOP requirement for entry into the Severe Accident Guidelines.
This is identified in the BWROG EPGs/SAGs when adequate core cooling cannot be assured. Potential Loss 2.A This water level co1Tesponds to the top of the active fuel and is used in the EOPs to indicate a challenge to core cooling. The RPV water level threshold is the same as RCS barrier Loss threshold 2.A. Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of the RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency.
64 DAEC FUEL CLAD BARRIER THRESHOLDS (cont.): This threshold is considered to be exceeded when, as specified in the EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization.
EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration ofRPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.
The term "cannot be restored and maintained above" means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation below the *limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.
In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL).
Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority.
For such events, ICs SA6 or SS6 will dictate the need for emergency classification.
Since the loss of ability to determine if adequate core cooling is being provided presents a significant challenge to the fuel clad barrier, a potential loss of the fuel clad barrier is specified.
: 3. Not Applicable (included for numbering consistency between barrier tables) 65 DAEC FUEL CLAD BARRIER THRESHOLDS (cont.): 4. Primary Containment Radiation Loss 4.A and Loss 4.B The Drywell and Torus radiation monitor readings correspond to an instantaneous release of all reactor coolant mass into the Drywell or Torus, assuming that reactor coolant activity equals 300 µCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier. The radiation monitor readings in this threshold are higher than that specified for RCS Barrier Loss threshold 4.A since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the emergency classification level to a Site Area Emergency.
There is no Potential Loss threshold associated with Primary Containment Radiation.
: 5. Other Indications Loss 5.A Results obtained from procedure PASAP 7.2, Fuel Damage Assessment, indicate at least 5% fuel clad damage. There is no Potential Loss threshold associated with Other Indications.
: 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors that are to be used by the Emergency Director in detennining whether the Fuel Clad Barrier is lost. Potential Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Fuel Clad Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
66 DAEC RCS BARRIER THRESHOLDS:
The RCS Barrier is the reactor coolant system pressure boundary and includes the RPV and all reactor coolant system piping up to and including the isolation valves. 1. Primary Containment Conditions Loss I.A 2 psig is the drywell high pressure scram setpoint which indicates a LOCA by automatically initiating ECCS. There is no Potential Loss threshold associated with Primary Containment Pressure.
: 2. RPV Water Level Loss 2.A + 15 inches corresponds to the top of active fuel (T AF) and is used in the EOPs to indicate challenge to core cooling. The RPV water level threshold is the same as Fuel Clad barrier Potential Loss threshold 2.A. Thus, this threshold indicates a Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier and that appropriately escalates the emergency classification level to a Site Area Emergency.
This threshold is considered to be exceeded when, as specified in the EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack oflow pressure RPV injection sources preclude Emergency RPV depressurization EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this RCS barrier Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.
The term, "cannot be restored and maintained above," means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.
67 DAEC RCS BARRIER THRESHOLDS (cont.): In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL).
Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority.
For such events, ICs SAS or SSS will dictate the need for emergency classification.
There is no RCS Potential Loss threshold associated with RPV Water Level. 3. RCS Leak Rate Loss Threshold 3.A Large high-energy lines that rupture outside primary containment can discharge significant amounts of inventory and jeopardize the retaining capability of the RCS until they are isolated.
If it is determined that the ruptured line cannot be promptly isolated from the Control Room, the RCS barrier Loss threshold is met. Loss Threshold 3.B Emergency RPV Depressurization in accordance with the EOPs is indicative of a loss of the RCS barrier. If Emergency RPV Depressurization is performed, the plant operators are directed to open safety relief valves (SRVs) and keep them open. Even though the RCS is being vented into the suppression pool, a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary.
Potential Loss Threshold 3.A Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Nonna! Operating values in areas such as main steam line tunnel, RCIC, HPCI, etc., which indicate a direct path from the RCS to areas outside primary containment.
A Max Normal Operating Limit (MNOL) value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly.
The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification.
A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system. An UNISOLABLE leak which is indicated by MNOL values escalates to a Site Area Emergency when combined with Containment Barrier Loss threshold 3.A (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.
68 DAEC RCS BARRIER THRESHOLDS (cont.): 4. Primary Containment Radiation Loss 4.A The Drywell monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold 4.A since it indicates a loss of the RCS Barrier only. There is no Potential Loss threshold associated with Primary Containment Radiation.
: 5. Other Indications There are no Loss or Potential Loss thresholds associated with Other Indications.
: 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the RCS barrier is lost. Potential Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
69 DAEC CONTAINMENT BARRIER THRESHOLDS:
The Primary Containment Barrier includes the drywell, the wetwell, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.
: 1. Primary Containment Conditions Loss l .A and l .B Rapid UNPLANNED loss of drywell pressure (i.e., not attributable to drywell spray or condensation effects) following an initial pressure increase indicates a loss of drywell integrity.
Drywell pressure should increase as a result of mass and energy release into the primary containment from a LOCA. Thus, drywell pressure not increasing under these conditions indicates a loss of primary containment integrity.
These thresholds rely on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned.
The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition.
Loss l.C The use of the modifier "direct" in defining the release path discriminates against release paths through interfacing liquid systems or minor release pathways, such as instrument lines, not protected by the Primary Containment Isolation System (PCIS). The existence of a filter is not considered in the threshold assessment.
Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream. Following the leakage of RCS mass into primary containment and a rise in primary containment pressure, there may be minor radiological releases associated with allowable primary containment leakage through various penetrations or system components.
Minor releases may also occur if a primary containment isolation valve(s) fails to close but the primary containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of primary containment but should be evaluated using the Recognition Category R I Cs. Loss l.D EOPs may direct primary containment isolation valve logic(s) to be intentionally bypassed, even if offsite radioactivity release rate limits will be exceeded.
Under these conditions with a valid primary containment isolation signal, the containment should also be considered lost if primary containment venting is actually performed.
Intentional venting of primary containment for primary containment pressure or combustible gas control to the secondary containment and/or the environment is a Loss of the Containment.
Venting for primary containment pressure control when not in an accident situation (e.g., to control pressure below the drywell high pressure scram setpoint) does not meet the threshold condition.
70 DAEC CONTAINMENT BARRIER THRESHOLDS (cont.): L ~----Potential Loss l .A The threshold pressure is the Torus internal design pressure.
Structural acceptance testing demonstrates the capability of the primary containment to resist pressures greater than the internal design pressure.
A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier. Potential Loss l .B If hydrogen concentration reaches or exceeds the lower flammability limit, as defined in plant EOPs, in an oxygen rich environment, a potentially explosive mixture exists. If the combustible mixture ignites inside the primary containment, loss of the Containment barrier could occur. Potential Loss 1.C The Heat Capacity Limit (HCL) is the highest suppression pool temperature from which Emergency RPV Depressurization will not raise:
* Suppression chamber temperature above the maximum temperature capability of the suppression chamber and equipment within the suppression chamber which may be required to operate when the RPV is pressurized, OR
* Suppression chamber pressure above Primary Containment Pressure Limit A, while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent. The HCL is a function of RPV pressure, suppression pool temperature and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant and therefore, the inability to maintain plant parameters below the limit constitutes a potential loss of containment.
71 DAEC CONTAINMENT BARRIER THRESHOLDS (cont.): 2. RPV Water Level There is no Loss threshold associated with RPV Water Level. Potential Loss 2.A The Potential Loss threshold is identical to the Fuel Clad Loss RPV Water Level threshold 2.A. The Potential Loss requirement for Primary Containment Flooding indicates adequate core cooling cannot be restored and maintained and that core damage is possible.
BWR EPGs/SAGs specify the conditions that require primary containment flooding.
When primary containment flooding is required, the EPGs are exited and SA Gs are entered. Entry into SA Gs is a logical escalation in response to the inability to restore and maintain adequate core cooling. PRA studies indicate that the condition of this Potential Loss threshold could be a core melt sequence which, if not corrected, could lead to RPV failure and increased potential for primary containment failure. In conjunction with the RPV water level Loss thresholds in the Fuel Clad and RCS barrier columns, this threshold results in the declaration of a General Emergency.
: 3. Primary Containment Isolation Failure These thresholds address incomplete containment isolation that allows an UNISOLABLE direct release to the environment.
Loss 3.A The Max Safe Operating Limit (MSOL) for Temperature and Radiation Level are each the highest value of these parameters at which neither: (1) equipment necessary for the safe shutdown of the plant will fail, nor (2) personnel access necessary for the safe shutdown of the plant will be precluded.
EOPs utilize these temperatures and radiation levels to establish conditions under which RPV depressurization is required.
The temperatures and radiation levels should be confirmed to be caused by RCS leakage from a primary system. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system. In combination with RCS potential loss 3.A this threshold would result in a Site Area Emergency.
There is no Potential Loss threshold associated with RCS Leak Rate. 72 DAEC CONTAINMENT BARRIER THRESHOLDS (cont.): 4. Primary Containment Radiation There is no Loss threshold associated with Primary Containment Radiation.
Potential Loss 4.A The drywell radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the drywell, assuming that 20% of the fuel cladding has failed. The radiation monitor reading for the torus corresponds to an instantaneous release of all reactor coolant mass directly into the torus, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds.
NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency classification level to a General Emergency.
: 5. Other Indications There is no Loss threshold associated with Other Indications Potential Loss 5 .A Results obtained from procedure PASAP 7.2, Fuel Damage Assessment, indicate at least 25% fuel clad damage. PASAP 7.2 only shows whether fuel damage is greater than or less than 25%, thus this indication is not likely to be declared before containment barrier potential loss 4.A which indicates 20% fuel damage. However, this potential loss threshold adds an additional layer of diversity to the scheme. 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Containment barrier is lost. Potential Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
73 
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74 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS 75 ECL: Notification of Unusual Event Initiating Condition:
Confirmed SECURITY CONDITION or threat. Operating Mode Applicability:
Confirmed SECURITY CONDITION or threat. Operating Mode Applicability:
All Emergency Action Levels: HU1 HUl.l A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by DAEC Security Shift Supervision.
All Emergency Action Levels: NE! 99 01 (RevisioR
HUl.2 HUl.3 Notification of a credible security threat directed at DAEC. A validated notification from the NRC providing information of an aircraft threat. Definitions:
: 6) November 2012 GENERAL EMERGENCY HG7 Other conditions e>,ist which in the judgment of the Emergency Director warrant declaration of a General Emergency.
Op. ,\1odes: All HU1 E xample Emergency Action Levels: (1 or 2 or 3) A SECURJTY CONDITION that does not involve a HOSTILE ACTION as reported by the (site specific security shift supervision).DAEC Security Shift Supervision. Notification of a credible security threat directed at the siteDAEC. A validated notification from the NRC providing information of an aircraft threat. Definitions:
SECURITY CONDITION:
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.
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 S E CURITY 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.
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. -terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). 181 
Basis: This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR 73.71 or 10 CFR 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HAI, HSI and HGl. Timely and accurate communications between DAEC Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and offsite response organizations.
}ffil 99 0 I (Re,*isioA 6) }fo 1 remeer 2012 SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including theincluding the ECCS. These systems are classified as safety-related.
76 Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].
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 f-73.71 or 10_-CFR--§ 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HAI, HSI and HGl. Timely and accurate communications between DAEC Security Shift Superv i sion 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 personne l and GRGoffsite response organization
EAL HUI .I references DAEC Security Shift Supervision because these are the individuals trained to confirm that a security event is occurring or has occurred.
: s. 182 NEI 99 QI (RevisieA a) !>foveme er 2012 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}.
Training on security event confirmation and classification is controlled due to the nature of Safeguards and IO CFR § 2.390 information.
183 NEI 99 Q 1 (ReYisioA
EAL HUI .2 addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with Abnormal Operating Procedure (AOP) 914, Security Events. EAL HUI .3 addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft.
*i) l'loYeFAeeF 2Q12 EAL HUI.I references (site specific security shift supervision)DAEC Security Shift Supervision because these are the indi vidua l s trained to co nfirm that a security event i s occurring or has occ urr ed. Training on security event confirmation and classification i s controlled due to the nature of Safeguards and 10 CFR § 2.39 Q inform atio n. EAL HUI.2 addresses the receipt of a credib l e security threat. The credibi lity of the threat is assessed in accordance with Abnormal Operating Procedure (AOP) 914, Security Events. specific procedure). EAL HU 1.3 addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) wi ll comm unicat e to the li censee i f the threat inv olves an aircraft.
The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with Abnormal Operating Procedure (AOP) 914, Security Events. Emergency plans and implementing procedures are public documents; therefore, EALs do not incorporate Security-sensitive information.
T h e status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat i s performed in accordance w ith (site specific procedure)
This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location.
Abnormal Operating Procedure (AOP) 914, Security Events . .,. Emergency plans and implementing procedures are public documents; therefore , EALs should do not in corporate Security-sensitive in formation. This includ es information that may be advantageous to a potential adversary , such as the particulars concerning a specific threat or threat l ocat ion. Security-sensitive information should beis contained in non public documents such as the Security Plan. Esca l ation of the emerge n cy class ific ation l eve l wou ld be via IC HA 1. Developer Notes: The (site specific security shift supervision) is the title of the on shift individual responsible for supervision of the on shift security force. The (site specific procedure) is the procedure(s) used by Control Room and/or Security personnel to determine if a security threat is credible, and to validate receipt of aircraft threat information.
Security-sensitive information is contained in the Security Plan. Escalation of the emergency classification level would be via IC HAI. 77 ECL: Notification of Unusual Event Initiating Condition:
Emergency plans and implementing procedures are public documents; therefore, E/\Ls should not incorporate Security sensitive information. This includes information that may be advantageous to a potential adversary , such as the particulars concerning a specific threat or threat location.
Security sensitive information should be contained in non public documents such as the Security Plan. With due consideration given to the above developer note, EALs may contain alpha or numbered references to selected events described in the Security Plan and associated implementing procedures.
Such references should not contain a recognizable description of the event. For example , an EAL may be worded as "Security event #2, #5 or #9 is reported by the (site specific security shift supervision)
." EGL Assignment Attributes:
3 .1.1.A 184 NEI 99 0 I (Re 1 1isioR 6) }io¥ember 2012 ECL: Notification of Unusual Event Initiating Condition:
Seismic event greater than OBE levels. Operating Mode Applicability:
Seismic event greater than OBE levels. Operating Mode Applicability:
All Emergency Action Levels: HU2 HU2.1 Seismic event greater than Operating Basis Earthquake (OBE) as indicated by receipt of the Amber Operating Basis Earthquake Light and the wailing seismic alarm on 1C35. Definitions:
All Exa1Rple Emergency Action Levels: H 2.1 Seismic event greater than Operating Basis Earthquake (OBE) as indicated by+ HU2 --+----receipt of the Amber Operating Basis Earthquake Light and the wailing seismic alarm on IC35. 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.
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.
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).
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 t. An earthquake greater than an OBE but less than a Safe ShutdownDesign Basis Ear thquake (SSeDBE)i 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.
Given the time necessary to perform walk-downs and inspections, and fully under sta nd any impacts , this event represents a potential degradation of the level of safety of the plant. Eve nt verification with external sources s hould not be necessary durin g or following an OBE. Ea rthquakes of this magnitude should be readily felt by on-site personnel and recognized as a se ismic event (e.g., typical lateral accelerations are in e>wess of 0.08g). The Shift Manager or E mergenc y Director ma y seek external verification if deemed appropriate ( e.g., a call to the + AR OBE is ¥ibratory grouRd motioR fur whieh those features ofa Ruelear power plaRt Reeessary for eoRtiRued operatioR without uRdue risk to the health aRd safety of the publie will remaiR fuAetioAal.
OBE events are detected in accordance with AOP 901. The OBE is associated with a peak horizontal acceleration of+/- 0.06g. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SAS. 78 HU3 ECL: Notification of Unusual Event Initiating Condition:
;! AR SSE is Yibratory grouRd motioR fur whieh eertaiR (geRerally, safety related) struetures , systems , aAd eompoReRts must be desigRed to remaiR fuAetioRal.
Hazardous events Operating Mode Applicability:
185 NEI 99 0 I (Re\*ision
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.
: 6) NoYember 2012 USGS , check internet news so urces , etc.); however , the verification action must not preclude a timely emergency declaration.
HU3.l HU3.2 HU3.3 HU3.4 A tornado strike within the PROTECTED AREA. Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode. Movement of personnel within the PROTECTED AREA is impeded due to an off site event involving hazardous materials ( e.g., an off site chemical spill or toxic gas release).
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 .£.A.9SA8. De"-1 el0per Netes: This "site specific indication that a seismic event met or exceeded QBE limits" should be based on the indications, alarms and displays of site specific seismic monitoring equipment.
A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles.
Indications described in the EAL should be limited to those that are immediately available to Control Room personnel and which can be readily assessed. Indications available outside the Control Room and/or 1.vhich require lengthy times to assess (e.g., processing of scratch plates or recorded data) should not be used. The goal is to specify indications that can be assessed within 15 minutes of the actual or suspected seismic event. For sites that do not have readily assessable QBE indications within the Control Room, developers should use the following alternate EAL (or similar 1.vording). (1) a. b. Control Room personnel feel an actual or potential seismic event. The occurrence of a seismic event is confirmed in manner deemed appropriate by the Shift Manager or Emergency Director.
Definitions:
The EAL l .b statement is included to ensure that a declaration does not result from felt vibrations caused by a non seismic source (e.g., a dropped heavy load). The Shift Manager or Emergency Director may seek e><ternal verification if deemed appropriate (e.g., a call to the USGS, check internet nev,rs sources, etc.); howe*,rer, the verification action must not preclude a timely emergency declaration.
PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
It is recognized that this alternate EAL *.vording may cause a site to declare an Unusual Event *while another site , similarly affected but with readily assessable QBE indications in the Control Room, may not. The above alternate
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.
*n<ording may also be used to develop a compensatory EAL for use during periods vrhen a seismic monitoring system capable of detecting an QBE is out of service for maintenance or repair. EGL Assignment Attributes:
Classification is also required if the water level or related wetting* causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode. EAL HU3.3 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel.
3.1.1.A 186 NEI 99 Q 1 (ReYision
within the PROTECTED AREA. 79 EAL HU3.4 addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles.
: 6) No 1 1ember 2Ql2 HU3 ECL: Notification of Unu s ual Event Initiating Condition:
Examples of such an event include site flooding caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road. This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the HmTicane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011. Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C. 80 HU4 ECL: Notification of Unusual Event Initiating Condition:
Hazardous event.&sect; Operating Mode Applicability:
FIRE potentially degrading the level of safety of the plant. Operating Mode Applicability:
All Emergency Action Levels: Example EmeFgeney Aetion Levels: (1 or 2 or 3 or 4 or 5 or 6) Note: E AL HU3.4 does not apply to routine traffic impediments such as fog , snow , ice , or vehicle breakdown s o r accident s. ~1 !k 2 H~J3.3 H LJ 3.4 A tornado strike within the PROTECTED AREA. Internal room or area floodin g 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 PROTECT E D AREA is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).
All Emergency Action Levels: Notes: HU4.1 HU4.2 HU4.3 HU4.4
A hazardous event that results in on-site condition s sufficient to prohibit the plant staff from accessing the site via personal vehicles. Definitions:
* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.  
PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related
: a. A FIRE is NOT extinguished within 15-minutes of ANY of the following FIRE detection indications:
.A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the EGGS. Systems classified as safety re l ated. Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. EAL HU3.1 addresses a tornado striking (touching down) within the Protected Area. E AL HU3.2 addresses floodin g of a building room or area that results in operators isolating power to a SAFETY SYST E M 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 SAF E TY SYSTEM component from its power source ( e.g., a breaker or relay trip). To 187 Jloffil 99 01 (Revision a) Jl,fo,,cember 2012 warrant classification , operability of the affected component must be required by Technical Specifications for the current operating mode. EAL HU3.3 addresses a hazardous materials event orig in ating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA. 188 NEI 99 01 (RevisioR e) NoYeR'leer 2012 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 fai lur e , 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 ve hicl e breakdowns or accidents , but rather to more s i gnificant conditions such as the Hurricane Andrew strike on Turkey Point in 1992 , the flooding around the Cooper Stat i on during the Midwest floods of 1993, or the flooding around Ft. Ca lh oun Station in 2011. EAL .H..!LL5 addresses (site specific description). . Escalation of the emergency classification l eve l wou ld be based on I Cs in Recognition Categories AR , F, Sor C. De'i'elepeF Netes: The "Site specific list of natural or technological hazard events" should include other events that may be a precursor to a more significant event or condition , and that are appropriate to the site location and characteristics. Notwithstanding the events specifically included as EALs above , a " Site specific list of natural or technological hazard events" need not include short lived events for which the extent of the damage and the resulting consequences can be determined 1.vithin a relatively short time frame. In these cases , a damage assessment can be performed soon after the event, and the plant staff 1.vill be able to identify potential or actual impacts to plant systems and structures.
This 1.vill enable prompt definition and implementation of compensatory or corrective measures with no appreciable increase in risk to the public. To the e>, tent that a short lived event does cause immediate and significant damage to plant systems and structures, it will be classifiable under the Recognition Category f, S and C ICs and EALs. Events of lesser impact would be e>,pected to cause only small and localized damage. The consequences from these types of events are adequately assessed and addressed in accordance with Technical Specifications.
In addition , the occurrence or effects of the event may be reportable under the requirements of 10 CFR 50.72. EGL Assignment Attributes:
3.1.1.A and 3.1.1.C 189 NEI 99 Q l (RtwisieR e) Ne&#xa5;ember 2012 HU4 ECL: Not ificati on of U nu s u a l Eve nt Initiating Condition:
F I RE potentially degrading the lev el of safety of the plant. Operating Mode Applicability:
All Emergency Action Levels: Example Emergeeey Aetiee Levels: (1 or 2 or 3 or 4) Note~:
* T h e E m ergency Director s hould declar e the Unusual Eventevent promptl y upon determining that the app li ca ble time ha s b ee n exceede d , or will lik e l y b e excee ded. H LJ 4.l a. A FIRE i s NOT ext in guis h ed w ithin 15-minutes of ANY of the follow in g FIRE d e t ectio n indi catio ns: H 4.2 H 4.4
* Report from the field (i.e., visual observation)
* Report from the field (i.e., visual observation)
* Receipt of multiple (more than 1) fire alarms or indications
* Receipt of multiple (more than 1) fire alarms or indi cat i o n s
* Field verification of a single fire alann AND b. The FIRE is located within ANY Table H-1 plant rooms or areas a. Receipt of a single fire alarm with no other indications of a FIRE. AND b. The FIRE is located within ANY Table H-1 plant rooms or areas AND c. The existence of a FIRE is not verified within 30-minutes of alarm receipt. A FIRE within the plant or ISFSI PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication.
* Fie ld verificat ion of a s in g l e fire a l ar m AND b. The FIRE i s l oca ted w ithin ANY of the followingTable H-1 plant rooms or areas+ a. (site specific list of plant rooms or areas) Receipt of a sing l e fire alarm ~with n o other indications of a FIRE j. ---AND b. The FIRE i s l ocate d w ithin ANY of the followingTable H-1 plant roo m s or areas (site specific I ist of plant rooms or areas) ----AND c. The ex i ste nc e of a FIRE is not verified within 3 0-minut es of a larm receipt. A FIRE w ithin the plant or ISFSI [forplemts wit.Li an ISFSI eutside t,l1ep/a,9t Preteeted Afe.at-PROTECTED AREA n ot ext in g ui s h ed within 60-minutes of the initi a l report , alar m or indication.
A FIRE within the plant or ISFSI PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.
A FIRE w ithin the pl ant or ISF SI [for plants wit.Li an ISFSI eutside the plant Pfflteeted Afe.at-PROTECTED AREA that requires firefighting s upp ort by an offs it e fire response agency to extinguish.
Table H-1 Fire Areas
Table H-1 Fire Areas 190
* 1 G31 DG and Day Tank Rooms
* I G31 DG and Day Tank Rooms,
* 1G21 DG and Day Tank Rooms
* I G21 DG and Day Tank Rooms,
* Battery Rooms
* Battery Rooms,
* Essential Switchgear Rooms
* Essential Switchgear Rooms,
* Cable Spreading Room
* Cable Spreading Room
* TorusRoom
* Torus Room * ~Intake Structure;
* Intake Structnre
* Pumphouse
* Pumphouse " Drywell
* Drywell,
* Torus
* Torus
* NE, NW, SE Comer Rooms
* NE. NW. SE Corner Rooms,
* HPCIRoom
* HPCI Room,
* RCICRoom
* RCIC Room,
* RHR Valve Room
* RHR Va l ve Room,
* North CRD Area
* North CRD Area,
* South CRD Area
* South CRD Area,
* CSTs
* CSTs
* Control Building
* Control Building,
* Remote Shutdown Panel 1 C388 Area
* Remote Shutdown Panel I C388 Area,
* Panel 1C55/56 Area
* Panel ICSS/56 Area;
* SBGTRoom 81 Definitions:
* SBGTRoom 191 Jl,ffil 99 g I (ReyisioR e) jl,fo,,cemeer 2('.)12 Definitions:
FIRE: Combustion characterize.cl by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
NEI 99 0 I (ReYisioR
PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. EALHU4.1 The intent of the IS-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc. Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was perfonned.
: 6) }fo 1 i8FR88F 20] 2 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.
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.
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. E AL HU4.1 The intent of the 15-minute duration i s to size the FIRE and to discriminate against small FIRES that are readi l y extinguished (e.g., smoldering waste paper basket). In addit ion to alarms , other indications of a FIRE cou ld 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.
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.
Similarly , the fire duration clock also starts at the time ofreceipt of the initial alarm , indication or report. EAL HU4.2 This EAL addresses receipt of a sing l e fire a l arm , 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 a l arm. 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.
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.
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 , additiona l time is allowed to verify the valid it y 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.
EALHU4.3 In addition to a FIRE addressed by EAL HU4.1 or EAL HU4.2, a FIRE within the plant or ISFSI PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. This basis extends to a FIRE occurring within the PROTECTED AREA of an ISFSI located outside the plant PROTECTED AREA. 82 EALHU4.4 If a FIRE within the plant or ISFSI PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded.
192 
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.
~IBI 99 01 (ReYisioA
Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions. Basis-Related Requirements from Appendix Rand NFPA-805 Criterion 3 of Appendix A to 10 CFR 50 states in part that "structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions." The Nuclear Safety Goal ("NSG") in NFPA 805, Section 1.3.1 states, "The nuclear safety goal is to provide reasonable assurance that a fire during any operational mode and plant configuration will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition." When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance because a safe shutdown success path, free of fire damage, must be available to meet the nuclear safety goals, objectives and performance criteria for a fire under any plant operational mode or configuration.
: 6) ~Jovemeer 20 I 2 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.
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.
EAL HU4.3 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. [Sentence for plants with an ISFSI outside the plant Protested i\rea] EAL HU4.4 If a FIRE within the plant or ISFSI [for plants with tm ISFSI eutside the plant PretectedArea]
In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of I-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). Even though DAEC has adopted the alternate approach provided by NFPA-805 in lieu of the deterministic requirements of Appendix R, the 30-minutes to verify a single alarm as used in EAL HU4.2 is considered a reasonable amount of time to determine if an actual FIRE exists without presenting a challenge to the nuclear safety performance criteria.
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.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SAS. 83 HU6 ECL: Notification of Unusual Event Initiating Condition:
Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions. Basis-Related Requirements from Appendix Rand NFPA-805 Criterion 3 of Appendix A to 10 CFR 50 states in part that "structures, systems, and components important to safety shall be designed and located to minimize.
Other conditions exist which in the judgment of the Emergency Director warrant declaration of a NOUE. Operating Mode Applicability:
consistent with other safety requirements, the probability and effect of fires and explosions
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.
." The Nuclear Safety Goal (" NSG") in NFPA 805, Section 1.3.1 states, "The nuclear safety goal is to provide reasonable assurance that a fire during any operational mode and plant configuration will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition." When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance because a safe shutdown success path, free of fire damage, must be available to meet the nuclear safety goals, objectives and performance criteria for a fire under any plant operational mode or configuration.
No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs. Definitions:
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.
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.
193 J>ffil 99 O 1 (Re,*isioR e) J>fo*yemaer 2012 In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of I-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). Even though DAEC has adopted the alternate approach provided by NFPA-805 in lieu of the deterministic requirements of Appendix R, the 30-minutes to verify a single alarm as used in EAL HU4.2 is considered a reasonable amount of time to determine if an actual FIRE exists without presenting a challenge to the nuclear safety performance criteria.Basis Related Requirements from Appendix R Appendi>l R to 10 CFR 50 , states in part: Criterion 3 of Appendi>, A to this part specifies that "Structures , systems, and components important to safety shall be designed and located to minimize , consistent 1.vith other safety requirements , the probability and effect of fires and e>,plosions." '.&#xa5;hen considering the effects of fire, those systems associated 1.vith achieving and maintaining safe shutdovm conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil off. Because fire may affect safe shutdovm 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.
Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a NOUE. 84 HA1 ECL: Alert Initiating Condition:
In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1 hour fire barriers for the enclosure of cable and equipment and associated non safety circuits of one redundant train (G.2.c). As used in EAL #2 , the 30 minutes to verify a single alarm is ,veil within this worst case 1 hour time period. Depending upon the plant mode at the time of the event, escalation of the emergency classification level wou ld be via IC CA6 or SA-9SA8. De...eloper Notes: The "site specific list of plant rooms or areas" should specify those rooms or areas that contain SAFETY SYSTEM equipment.
As noted in the EALs and Basis section , include the term ISFSI if the site has an ISFSI outside the plant Protected Area. EGL Assignment Attributes:
3 .1.1.A 194 ECL: Notification of Unusual Event NEJ 99 01 (RevisioA
: 6) ~fo*,<emeer 2012 HU-76 Initiating Condition:
Other conditions exist which in the judgment of the Emergency Director warrant declaration of a fN 0 1 UE. Operating Mode Applicability:
All l Exemple Emergency Action Levels: 1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.
No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS safety systems occurs. Definitions:
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or p l acing it in the cold shutdown condition, i nc l uding the ECCS. These systems are classified as safety-related
.A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdovm condition, including the EGGS. Systems classified as safety re l ated. Basis: This IC addresses unanticipated conditions not addressed explicit l y elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fa ll under the emergency classificat i on level description for a NOUE. 195 NEJ 99 QI (RevisioR
: 6) No 1 t'0FR00F 2()12 HA1 ECL: Alert Initiating Condition:
HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes. Operating Mode Applicability:
HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes. Operating Mode Applicability:
All Emergency Action Levels: HAI.I HAl.2 A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the DAEC Security Shift Supervision.
All Example Emergency Action Levels.:.: (1 or 2) A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the (site specific security shift supervision)DAEC Security Shift Supervision. A validated notification from NRC of an aircraft attack threat within 30 minutes of the site. Definitions:
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.
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.
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 CONTROLL E D AREA: The site property owned by or otherwise under the control of the licensee. PROJECTILE
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. Th i s event will require rapid response and assistance due to the possibi l ity 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 Sh i ft Supervision and the Control Room is essentia l for proper c l assification of a security-re l ated 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}.
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].
196 NEI 99 0 I (RevisioA
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).
*i) NoYeFAeer 2012 As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures ( e.g., evacuation , dispersal or sheltering).
85 The Alert declaration will also heighten the awareness of Offsite Response Organizations, allowing them to be better prepared should it be necessary to consider further actions. This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR 73.71 or 10 CFR 50.72. EAL HAI.I is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes any action directed against the ISFSI which is located outside the plant PROTECTED AREA. EAL HAI .2 addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and offsite response organizations are in a heightened state of readiness.
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-&sect;-73.71 or 10 CFR-&sect;-50.72. EAL HA I.I is applicable for any HOSTILE ACTION occurring , or that has occurred , in the OWNER CONTROLLED AREA. This includes any action directed against att-the ISFSI that 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 GRGoffsite response organization s are in a heightened state of readiness.
This EAL is met when the threat-related information has been validated in accordance with Abnormal Operating Procedure (AOP) 914, Security Events. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft.
This EAL is met when the threat-related information has been validated in accordance with fAbnormal Operating Procedure (AOP) 914, Security Events site specific procedure).&sect;.,_
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.
The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft.
This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location.
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 should do not incorporate Security-sensitive information.
Security-sensitive information is contained in the Security Plan. Escalation of the emergency classification level would be via IC HS 1. 86 HAS ECL: Alert Initiating Condition:
This includes information that may be advantageous to a potential adversary , such as the particulars concerning a specific threat or threat location.
Security-sensitive information should be ~contained in non public documents such as the Security Plan. Escalation of the emergency classification level would be via IC HS 1. DevelopeF Notes: The (site specific security shift supervision) is the title of the on shift individual responsible for supervision of the on shift security force. Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location.
Security sensitive information should be contained in non public documents such as the Security Plan. 197 
~ffil 99 0 I (Re*,isioA e) urth 8 . November 201? .r1ue cons1aeration given to the abo*c e ., -**~hoR>0 :'*'&deg;'""***
te soloote0 ovoAls Bosoribo~
iElet~el';J'er
*.***. EALs "'"'.I ***teio alpha e, imp ementmg proceaures.
Such references h la n e ec~nty Plan ans associates event. For e:irnm I e s ou not contam a re
* bl . ( * . ----p*e , ao -AL fllay be wer0e8 as " S _eeg,,,za
* Elesenptieo efthe site specific security shift supervision)." ecunty event #2 , #5 or #9 is reportea by the See the relates Developer Note . A
* aevelopment of a scheme aefinition for1:h;*~~~~~~BCgefinitions , for guiaance on the ** }>HROLLBD A~ A eCL Assignment Attributes:
3.1.2.D '..:. n. 198 NEI 99 01 (RevisioR e) ?>lovember 2012 HA5 ECL: Alert Initiating Condition:
Gaseous release impeding access to equipment necessary for normal plant operations , cooldo'+'m , or shutdown. Operating Mode Applieal>ility:
All Example Emergeney Aetion Levels: Note: If the equipment in the listed room or area \Vas already inoperable or out of service before the event occurred , then no emergency classification is warranted.
follo'+'i'ing plant rooms or areas: (site specific I ist of plant rooms or areas 1 Nith entry related mode applicability identified) ,t\..l\JD
: b. Entry into the room or area is prohibited or impeded. This IC addresses an event involving a release of a hazardous gas that precludes or impedes access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An Alert declaration is warranted if entry into the affected room/area is, or may be , procedurally required during the plant operating mode in effect at the time of the gaseous release. The emergency classification is not contingent upon whether entry is actually necessary at the time of the release. Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the Emergency Director's judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly impede procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis , report of ill effects on personnel, advice from a subject matter ex.pert or operating e>tperience vrith the same or similar hazards. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment , such as SCBAs , that is not routinely employed).
An emergency declaration is not 1 Narranted if any of the following conditions apply.
* The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous release).
For e>rnmple, the plant is in Mode 1 1.vhen the gaseous release occurs , and the procedures used for normal operation , cooldown and shutdown do not require entry into the affected room unti I Mode 4 . 199 
: h. h address the r , measures 'tV ie . ludes compensate. Y , tern testing). . :~:::;:::*;;;n;e~d;a~c~ftp~,'i~t)~
'~ th~a~t;m:;c;;Efir:es:u:p:p:re:s:s:10:n::S)~:s*:::re:c:o:r:dk:e:e:pm:g
* TAe gas ,el ease IS a.:'~~ efa roam a, area (e.g;j is ef!lfl admiaistrot,ve e, erary iaaeeess, , , rea eat,y is req**ff> d "*euld Rel temp . fer wAieA ,eem,a fRe iRSj)eet10as).
f """'Y """'"'* aA "
* The action I rounds or mu I 'ati, '0 or precau I E norma f consen
* naturee.g., I easures are o a . The aeeess eeRtro* '." ede a required aeheA. d . te daagerous levels. aeUJally preveat er ,mp . tAe level ef eKygea iR !lie ~=s:.i eaviroameat. -!~" al,le ef redue1ag . I eiag aiflA aa eae * ., 4.ieA eaa lea e ' R aspAy,<iaat is a ga s *IIJ' ats werk l,y merely d1Sp ** al !~tel ef al'0uad 19 ,S , " n moni)', asphyxia , n below the nor Mest eem eRtmtiea ef e*) ge """" deatla. reduces the cone I
* unconsciousness , or 200 201 NEI 99 O 1 (RevisieR e) },fovemaer 2012 ECL: Alert NEI 99 01 (RevisioR e) ~fo*refReer 2012 HA6HA5 Initiating Condition:
Control Room evacuation resulting in transfer of plant control to alternate locations.
Control Room evacuation resulting in transfer of plant control to alternate locations.
Operating Mode Applicability:
Operating Mode Applicability:
All Emergency Action Level: HAS.I An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel (1 C388). Definitions:
All Example Emergency Action Level s: H 65.1 An event has resulted in plant control being transferred from the Control Room to specific remote shutdo 1 Nn panels and local control stations)the Remote Shutdown Panel (1 C388). Definitions:
None Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety. Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations.
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.
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.
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 HS 62_. Developer Notes: The "site specific remote shutdovrn panels and local control stations" are the panels and control stations referenced in plant procedures used to cooldown and shutdown the plant from a location(s) outside the Control Room. EGL Assignment Attributes:
Escalation of the emergency classification level would be via IC HS5. 87 HAG ECL: Alert Initiating Condition:
3.1.2.B 202 ECL: Alert NEI 99 01 (Re11ision t'i) ~lo*reff!eer 2012 HA7HA6 Initiating Condition:
Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert. Operating Mode Applicability:
Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert. Operating Mode Applicability:
All Emergency Action Level: HA6.1 Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. Definitions:
All l E,ample Eme.-geeeyEmergency Action Leve ls: 1 Other conditions exist which , in the judgment of the Emergency Director , indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. Definitions:
HOSTILE ACTION: An act toward DAEC or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION: 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  
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:
{i.e .. this may include violent acts between individuals in the owner controlled area). HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. PROJECTILE:
An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an Alert. 88 ECL: Site Area Emergency Initiating Condition:
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. 203 ECL: Site Area Emergency Initiating Condition:
HOSTILE ACTION within the PROTECTED AREA. Operating Mode Applicability:
HOSTILE ACTION within the PROTECTED AREA. Operating Mode Applicability:
All Emergency Action Levels: HS1 HSl.1 A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the DAEC Security Shift Supervision.
All Example EmergeneyEmergency Action Levels: NEI 99 01 (RevisioR e) ~lovemeer 2012 HS1 S 1.1 A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the (site specific security shift supervision)DAEC Security Shift Supervision. Definitions:
Definitions:
HOSTIL E 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 device s used to deliver destructive force. Other acts that satisfy the overall intent may be included.
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-ba s ed EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). HOST AGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTIL E 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.
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:
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.
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 acce s s 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].
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 , Safeguard s Contingency Plan [and Independent Spent Fuel Storage Installation Security Program}.
89 As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures ( e.g., evacuation, dispersal or sheltering).
204 NEI 99 01 (RevisioA
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.
: 6) NoYemeer 2012 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).
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.
The Site Area Emergency declaration will mobilize GRGoffsite 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 atrthe 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 f-73.71 or 10 CFR f-50.72. Emergency plans and implementing procedures are public documents; therefore, EALs should 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.
This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location.
Security-sensitive information is contained in the Security Plan. Escalation of the emergency classification level would be via IC HG 1. 90 HS5 ECL: Site Area Emergency Initiating Condition:
Security-sensitive information should beis contained in non public documents such as the Security Plan. Escalation of the emergency classification level would be via IC HGl. Develeper Netes: The (site specific security shift supervision) is the title of the on shift individual responsible for supervision of the on shift security force. Emergency plans and implementing procedures are public documents; therefore , EALs should not incorporate Security sensitive information.
This inc l udes information that may be advantageous to a potential adversary , such as the particulars concerning a specific threat or threat location.
Security sensitive information should be contained in non public documents such as the Security Plan. 'Nith due consideration given to the above developer note , EALs may contain alpha or numbered references to selected ev e nts described in the Security Plan and associated implementing procedures.
Such references should not contain a recognizable description of the event. For e).ample, an EAL may be 1.vorded as " Security event #2 , #5 or #9 is reported by the (site specific security shift supervision)." See the related Developer Note in Appendix B , Definitions , for guidance on the development of a scheme definition for the PROTECTED AREA. EGL i\ssignment Attributes:
3.1.3.D 205 l'JEI 99 QI (Re*,isioR e) }'fo*,emaer 2Q 12 HS6HS5 ECL: Site Area Emergency Initiating Condition:
Inability to control a key safety function from outside the Control Room. Operating Mode Applicability:
Inability to control a key safety function from outside the Control Room. Operating Mode Applicability:
All Emergency Action Levels: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
All Example Emergency Action Levels: Note: The Emergency Director should declare the Site Area Emergencyevent promptly upon determining that (site specific number the applicable timeof'.20 minutes) has been exceeded , or will likely be exceeded.  
HS5.l a. An event has resulted in plant control being transfen-ed from the Control Room to the Remote Shutdown Panel (1C388). AND b. Control of ANY of the following key safety functions is not reestablished within 20 minutes.
: a. b. Definitions:
Basis: An event has resulted in plant control being transferred from the Control Room to (site specific remote shutdown panels and_control stations) the Remote Shutdown Panel (1 C388). AND Control of ANY of the following key safety functions is not reestablished within (site specific number of20 minutes).
* Reactivity control
* Reactivity control
* RPV water level
* Core cooling [PWR] I RPV water level [BWR]
* RCS heat removal Definitions:
* RCS heat removal 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 g ain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time. The determination of whether or not "control" is established at the remote safe shutdown location(s)Remote Shutdown Panel (1 C388-islli based on Emergency Director judgment.
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 (the site specific time for transfer)
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).
~20 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s).
AOP 915, "Shutdown Outside Control Room" provides the following CAUTION -"For Control Room evacuation as the result of a fire, transfer of control at panels 1 C388, 1 C389, 1 C390, JC391, and JC392 is required to be completed within 20 minutes." Escalation of the emergency classification level would be via IC FGI or CGI. 91 HS6 ECL: Site Area Emergency Initiating Condition:
AOP 915, " Shutdown Outside Control Room" provides the following CAUTION -" For Control Room evacuation as th e r e sult ofa fir e , transfer of control at panels 1 C388, 1 C389, 1 C390. JC391, JC392and JC392 i s required to be completed within 20 minutes." E scalation of the emergency classification level would be via IC FG I or CG 1. 206 Developer Notes: NEJ 99 01 (ReYisioR
Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency.
: 6) No1rember 2012 The "site specific remote shutdown panels and local control stations" are the panels and control stations referenced in plant procedures used to eooldown and shutdown the plant from a location(s) outside the Control Room. The "site specific number of minutes" is the time in which plant control must be (or is expected to be) reestablished at an alternate location as described in the site specific fire response analyses.
Operating Mode Applicability:
Absent a basis in the site specific analyses, 15 minutes should be used. Another time period may be used with appropriate basis/justification.
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.
EGL Assignment Attributes:
Definitions:
3.1.3.B 207 ECL: Site Area E m ergency }ffil 99 QI (Re\1 isioR *i) }loyemeer 2012 HS7HS6 Initiating Condition:
Other conditions ex i s t w hi c h in th e judgment of the E mer ge nc y Dir ec tor warrant declaration of a S it e Area Emerge nc y. Operating Mode Applicability:
Al l E .. mple Emergency Action Level s: 1 Other conditions ex i st w hich in the jud gment of the Emerge nc y Dir ector indicat e t ha t events are in progress or hav e occ urr e d which invol ve act ual or lik e l y major failures of plant f un ct ion s n eede d fo r protection of the public or HOSTILE ACTI O N that r es ult s in int ent i o nal d amage or m a lici o u s acts , (1) toward si t e p erso nnel or eq uipm e nt that could l ead to th e lik e l y fai lur e of or , (2) th at prevent effect i ve access to equipme nt needed fo r the protection of the p ubli c. Any releases are not expected to result in expos ur e l eve l s w hich exceed EPA Protective Action Guideline expos ur e l evels b eyo nd the s it e boundar y. 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: 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:
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. terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individual s 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.
: An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. Basis: T hi s IC addresses unanti ci pat e d conditions not ad dr esse d explicitly e l sew h ere but that warrant d e claration of an emergency because conditions exist which are believed b y the E mergenc y Director to fall under the e mer ge nc y classification l eve l d esc ription for a Site Area E mer ge ncy. 208 NE! 99 01 (Re\'isioR 6) ~foYemaer 2012 HG1 ECL: General Emergency Initiating Condition:
92 HG1 ECL: General Emergency Initiating Condition:
HOSTILE ACTION resulting in loss of physical control of the facility.
HOSTILE ACTION resulting in loss of physical control of the facility.
Operating Mode Applicability:
Operating Mode Applicability:
All Emergency Action Level: HGl.1 a. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the DAEC Security Shift Supervision.
All Example Emergency Action Level s: H 1.1 a. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the (site specific security shift supervision)DAEC Security Shift Supervision. AND b. EITHER of the following has occurred:  
AND b. EITHER of the following has occurred:  
: 1. ANY of the following safety functions cannot be controlled or maintained.
: 1. ANY of the following safety functions cannot be controlled or maintained.
* Reactivity control
* Reactivity control
* RPV water level
* Core cooling [PWR] I RPV water level [BWR]
* RCS heat removal OR 2. Damage to spent fuel has occurred or is IMMINENT.
* RCS heat removal OR 2. Damage to spent fuel has occurred or is IMMINENT.
Definitions:
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: An act toward DAEC or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not pati of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE FORCE: One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.
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. terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). HOST AGE: 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:
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:
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. 209 NEI 99 QI (Re;cisioA a) November 2012 PROJECTILE:
An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. 93 Basis: This IC addresses an event in which a HOSTILE FORCE has taken physical control of the facility to the extent that the plant staff can no longer operate equipment necessary to maintain key safety functions.
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. 210 Basis: 211 tffil 99 01 (Re1, 1 isioR fi) tfo1, 1 emeer 2012 NEI 99 C:l 1 (Re,*isioA 6) Novemeer 2C:ll2 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.
It also addresses a HOSTIL E 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].
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].
E mergency plans and implementing procedures are public documents; therefore, EALs should do not incorporate Security-sensitive information.
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.
This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location.
Security-sensitive information should beis contained in non public documents such as the Security Plan. 212 Developer Notes: l>ffil 99 Q l (ReYisieA a) l>foi,*ember 2Ql 2 The (site specific security shift supervision) is the title of the on shift individual responsible for supervision of the on shift security force. Emergency plans and implementing procedures are public documents; therefore , EALs should not incorporate Security sensitive information.
Security-sensitive information is contained in the Security Plan. 94 HG6 ECL: General Emergency Initiating Condition:
This includes information that may be advantageous to a potential adversary , such as the particulars concerning a specific threat or threat location. Security sensitive information should be contained in non public documents such as the Security Plan. With due consideration given to the above developer note , EALs may contain alpha or numbered references to selected events described in the Security Plan and associated implementing procedures. Such references should not contain a recognizable description of the event. For e1rnmple , an EAL may be worded as " Security event #2, #5 or #9 is reported by the (site specific security shift supervision)." 8ee the related Developer Note in Appendix.
B , Definitions , for guidance on the development of a scheme definition for the PROTECTED AREA. EGL Assignment Attributes:
3 .1.4 .D 213 ECL: General Emergency l'>ffil 99 01 (ReYisioR
: 6) NoYember 2012 HG7HG6 Initiating Condition:
Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency.
Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency.
Operating Mode Applicability:
Operating Mode Applicability:
All Emergency Action Level: HG6.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occmTed which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.
All 6 E,ae,ple Emergency Action Leve ls: 1 Other conditions exist which in the judgment of the Emerge ncy Director indicate that events are in progress or have occurred which involve actua l or IMMINENT substantial core degradation or melting with potential for l oss of containment integrity or HOSTILE ACTION that results in an actua l loss of physica l control of the facility.
Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels off site for more than the immediate site area. Definitions:
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: An act toward DAEC or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-te1Torism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. IMMINENT:
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. 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:
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.
An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a General Emergency.
95 11 SYSTEM MALFUNCTION ICS/EALS 96 ECL: Notification of Unusual Event Initiating Condition:
214 11 SYSTEM MALFUNCTION ICS/EALS NEI 99 01 (ReYisieR e) }foyemaer 2012 Table S 1: Reeognition Categorv "S" Initiating Condition Matrix UNUSUAL EVENT SUl Loss of all offsite AC power capability to emergency buses for 15 minutes or longer. Op. },fades: 1. 2. 3. 4Pewer Operetien, Stertup , Het St611'lde:y, Hat 8/mtdewl'l SU2 illlPL",1'll>lED loss of Control Room indications for 15 minutes or longer. Op. ,\fades: Pewer Operetien , Stertup , Hat Stemie;*, Hat 8!1utdewnL...l,_
Loss of ALL offsite AC power capability to essential buses for 15 minutes or longer. Operating Mode Applicability:
: 3. 4 SUJ Reactor coolant actiYity greater tkan Technical Specification allowable limits. Op. },fades: 1. 2. 3. 4Pewer Operntien, Sf6lrtup, Het Stendh;*, Het Shutdewn SU4 RCS leakage for 15 minutes or longer. Op. }.fades: 1. 2. 3, '/Pewer Operetien , Stertup, Het Stendby , Het Shutdem'l SUS Automatic or manual (trip [PWRJ / scram [BWR]) fails to shutdown the reactor. Op. }.lodes: Pewer OpaGltien}
1, 2, 3 Emergency Action Level: SU1 Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
ALERT SAl Loss of all but one AC power source to emergency buses for 15 minutes or longer. Op. },fades: L...1....J...
SUI.I Loss of ALL offsite AC power capability to 1A3 AND 1A4 buses for 15 minutes or longer. Definitions:
1:.Pewer Opeffllien , Stertup , Het Sf6lndby, Het 8!1utdewn SA2 ill~PLA1'Jl>ffiD loss of Control Room indications for 15 minutes or longer witk a significant transient in progress.
None Basis: This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC*essential buses. This condition represents a potential reduction in the level of safety of the plant. The intent of this EAL is to declare a Notification of Unusual Event when offsite power has been lost and both of the emergency diesel generators have successfully started and energized their respective 4kv essential bus. For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the essential buses, whether or not the buses are powered from it. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power. Escalation of the emergency classification level would be via IC SAl. 97 SU3 ECL: Notification of Unusual Event Initiating Condition:
Op. Mades: 1. 2. 3. 4 Pewer Operetien , Stertup , Het Stendby , Het Shu1dew1q SAS Automatic or manual (trip [PWR] / scram [BWR]) fails to shutdown the reactor, and subsequent manua l actions taken at the reactor control consoles are not successful in shutting down the reactor. Op. }.{edes: Pewer Operetienl_
UNPLANNED loss of Control Room indications for 15 minutes or longer. Operating Mode Applicability:
2 15 SITE A.REA EMERGENCY SS1 Loss of all offsite and all onsite AC pov,rer to emergency buses for 15 minutes or longer. Op. },fades: 1, 2, 3, 4Pewer Operetien, Stertup, Het Stendb;*, Het Shutdewn SSS Inability to shutdovm the reactor causing a challenge to (core cooling [PWR] I R.0 V water level [BWR]) or RC8 heat removal. Op. }.fades: Pewer Operetienl 1 GENERAL EMERGENCY SGl Prolonged loss of all offsite and all onsite AC power to emergency buses. Op. }.fades: 1, 2. 3, 4Pewer Operetien, Stertup, Het Stendh;*, Het 8!1utdewn
1, 2, 3 Emergency Action Level: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
-------------------
SU3.1 An UNPLANNED event results in the inability to monitor one or more of the Table S-1 parameters from within the Control Room for 15 minutes or longer. -Table S-1>Safety System Parameters  
: Table intended for use by 1 EAL developers.
'
: Inclusion in licensee I 0 * . d , ocuments 1s not require . L------------------1 UNUSUAL EVENT SU6 Loss of all onsite or offsite eommunieations eapabilities.
Op. },fades: 1, 2, 3 , 4.Pewer Opereti e n , Stcwtup , Hat St a l'ldhy , Het Slw tde wn SU7 Failure to i s olate eontainment or loss of eontainment pressure eontrol. [PWR] Op. Afade s: 1 , 2 , 3, 1. Pewe,* Operatie,"i , Starh , tp, Het Stendhy , Hat Shi1tdewl'I A,LERT SITE AREA EMERGENCY NEI 99 01 (Re,,*isioR e) },foyemeer 2012 GENER .... L EMERGENCY SS8 Loss of all Vital DC SG8 Loss of all AC and SA9 Ha:mrdous event affeeting a SAFETY SYSTEM needed for the eurrent operating mode. Op. Med es: 1, 2 , 3, 4.Pewer Operatiel'I , Startitp , Hat S:emihy , Hat Shutdewl'1 po 1 Ner for 15 minutes or Op. },fades: 1 , 2, 3, 4Pewer Operetien , Startitp , H e t Standby*, Hat Shutdewl'I 216 Vital DC power souree s for 15 minutes or longer. Op. },1edes: 1, 2, 3, 4Pewer Operetien , St*lrt1,!fJ , Het Stendey , Hat Shutdewn ,-------------------, : Table intended for use b)' I EAL de>,<elopers.
: lnelusion in lieensee I d * . d , oeuments ts not require . 1 L------------------J ECL: Notification of Unusual Event NE! 99 0 I (ReYisioR
: 6) NoYember 2012 SU1 Initiating Condition:
Loss of alt-ALL offsite AC power capability to emergency essential buses for 15_-minutes or longer. Operating Mode Applicability:
Power Operation , Startup , Hot Standby , Hot Shutdown.1..1.,_]_
Example Emergency Action Level s: Note: The E mergency Director should declare the Unusual E~vent promptly upon determining that the applicable time 15 minutes has been exceeded , or will likely be exceeded.
S 1.1 Loss of ALL offsite AC power capability to (site specific emergency buses)1A3 AND 1 A4 buses for 15 minutes or longer. Definitions:
Basis: This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC emergency essential buses-. .,_ This condition represents a potential reduction in the level of safety of the plant. The intent of this EAL is to declare an 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 emergency 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. E scalation of the emergency classification level would be via IC SAL De*,zeloper Notes: The " site specific emergency buses" are the buses fed by offsite or emergency AC power sources that supply power to the electrical distribution system that powers SAFETY SYSTEMS. There is typically 1 emergency bus per train of SAFETY SYSTEMS. At multi unit stations , the EALs may credit compensatory measures that are proceduralized and can be implemented 1 within 15 minutes. Consider capabilities such as power source cross ties, " s 1.ving" generators , other power sources described in abnormal or emergenC)' operating procedures , etc. Plants that have a proceduralized capability to supply offsite AC power to an 217 NEl 99 QI (ReyisioR
: 6) tlo*,<emeer 2Q 12 affeeted unit via a eross tie to a eompanion unit may eredit this power souree in the EAL provided that the planned eross tie strategy meets the requirements of 10 CFR 50.63. EGL Assignment Attributes:
3.1.1.A 218 ECL: Notification of Unusua l Event NEI 99 QI (Re'iisioA
: 6) No&#xa5;ember 2Q 12 SU2SU3 Initiating Condition:
UNPLANNED l oss of Control Room indications for 15 minutes or longer. Operating Mode Applicability:
Power Operation , Startup , Hot Standby , Hot Shutdown.L..LJ Examf)le Emergency Action Level s: Note: The Emergency Director shou ld declare the Unusual E~ve n t promptly up o n determining that the applicable time 15 minutes h as been exceeded , or w ill l ikely be exceeded.
S 3.1 ++-a.--An UNPLANNED event results in the inability to monitor one or more of the Definitions:
Reactor Power R.0 V '.Vater Level RPV Pressure Primary Containment Pressure Suppression Pool Le 1 rel
* Suppression Pool Temperature Suppression Pool Temperature Table S-1 Safety System Parameters
* Reactor power
* Reactor power
* RPV Water Level
* RPV Water Level
Line 811: Line 268:
* Primary Containment Pressure
* Primary Containment Pressure
* Suppression Pool Level
* Suppression Pool Level
* Suppression Pool Temperature Definitions:
* 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. These systems are 219 
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.
}ffil 99 0 I (ReYisioA
UNPLANNED:
: 6) N0Ye1'l'!eer 2012 classified as safety-related.A system required for safe plant operation, cooling dovm the plant and/or placing it in the cold shutdovm condition, including the EGGS. Systems classified as safety related. 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.
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 p l ant 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).
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 , dig i tal and recorder source within the Control Room. 220 NE! 99 0 I (Re,*isioH
). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room. 98 An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required.
: 6) 1-foyemeer 2012 An event involving a los s of plant indications , annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required.
The event would be reported if it significantly impaired the capability to perform emergency assessments.
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.
In particular , emergency assessments necessary to implement abnormal operating procedures , emergency operating procedures , and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.
This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV level and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition.
This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control , core cooling [PWR] I RPV level [BWR] and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition.
In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for 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.
In addition, if all indication so urces for one or more of the listed parameters are lost , then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level [PWR] / RPV water level [BWR] cannot be determined from the indications and recorders on a main control board , the SPDS or the plant computer, the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Esca lation of the emergency classification level would be via IC SA+/-}. De*,releper Netes: In the PWR parameter list column , the "site specific number" should reflect the minimum number of steam generators necessary for plant cooldown and shutdovm. This criterion may also specify whether the level value should be *wide range , narrow range or both , depending upon the monitoring requirements in emergency operating procedures.
Escalation of the emergency classification level would be via IC SA3. 99 SU4 ECL: Notification of Unusual Event Initiating Condition:
Developers may specify either pressurizer or reactor 1 1essel level in the PWR parameter column entry for RCS Le 1 1el. The number , type , location and layout of Control Room indications , and the range of possible failure modes, can challenge the ability of an operator to accurate!)'
determine , within the time period available for emergenC)' classification assessments , if a specific percentage of indications have been lost. The approach used in this EAL facilitates prompt and accurate emergency classification assessment s by focusing on the indications for a selected subset of parameters. By focusing on the availability of the specified parameter values , instead of the sources of those values , the EAL recognizes and accommodates the wide variety of indications in nuclear power plant Control Rooms. Indication types and sources may be analog or digital , safety related or not , primary or alternate , individual meter value or computer group display , etc. /1 , loss of plant annunciators will be evaluated for reportability in accordance
*with 10 CFR 50.72 (and the associated guidance in 1'ruREG 1022), and reported if it significantly impairs the capability to perform emergency assessments.
Compensatory measures for a loss of annunciation can be readily implemented and may include increased monitoring of main control boards and more frequent plant rounds by non licensed operators.
Their alerting function notwithstanding , annunciators do not provide the parameter values or specific component status information used to operate the plant , or process through AOPs or EOPs. Based on these considerations , a loss of annunciation is considered to be adequately addressed by reportability criteria, and therefore not included in this JC and EAL. 221 
~ffil 99 QI (RevisieA
: 6) ~fo>,*emeer 2012 With respect to establishing event severity , the response to a loss of radiation monitoring data (e.g., process or effluent monitor values) is considered to be adequately bounded by the requirements of 10 CPR 50.72 (and associated guidance in NUREG 1022). The reporting of this event 1,vill ensure adequate plant staff and NRG awareness , and drive the establishment of appropriate compensatory measures and corrective actions. In addition , a loss of radiation monitoring data , by itself , is not a precursor to a more significant event. Personnel at sites that have a Failure Modes and Effects Analysis (H,4EA) included within the design basis of a digital I&C system should consider the FMEA information when developing their site specific EALs. Due to changes in the configurations of SAFETY SYSTEMS , including associated instrumentation and indications , during the cold shutdown , refueling , and defueled modes , no analogous IC is included for these modes of operation. EGL Assignment Attributes:
3.1.1.A 222 2 NEI 99 QI (ReYisioR e) Noyemeer 2Q 12 SU3SU4 ECL: Notification of Unusual Event Initiating Condition:
Reactor coolant activity greater than Technical Specification allowable limits. Operating Mode Applicability:
Reactor coolant activity greater than Technical Specification allowable limits. Operating Mode Applicability:
1, 2, 3 Emergency Action Levels: SU4.1 SU4.2 Pretreatment Offgas System (RM-4104)
Power Operation , Startup, Hot Standby , Hot Shutdov,mL 2, 3 Example Emergency Action Levels: (1 or 2) (Site specific radiation monitor) reading greater than (site specific value). Pretreatment Off gas System (RM-4104)
Hi-Hi Radiation Alarm. Sample analysis indicates that reactor coolant specific activity is greater than 2.0 &#xb5;Ci/gm dose equivalent I-131 for 12 hours or longer. Definitions:
Hi-Hi Radiation Alarm. Sample analysis indicates that reactor coolant specific activity is greater than 2.0 &#xb5;Ci/gm dose equivalent 1-131 for 12 hours or longerSample analysis indicates that a reactor coolant activity value is greater than an allowable limit specified in Technical Specifications.,,.,.
None Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications.
Definitions:
This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant. For EAL SU4.l, RM-4104 Hi-Hi Radiation Alarm has been chosen because it is operationally significant, is readily recognizable by the Control Room Operations Staff, and is set at a level corresponding to noble gas release rate, after 30-minute delay and decay of 1 Ci/sec. For EAL SU4.2, coolant samples exceeding the 2.0 &#xb5;Ci/gm dose equivalent I-13 lconcentration require prompt action by DAEC Technical Specifications and are representative of minor fuel cladding degradation.
Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications.
Escalation of the emergency classification level would be via ICs FAl or the Recognition Category R ICs. 100 ECL: Notification of Unusual Event Initiating Condition:
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-m i nute delay and decay of 1 Ci/sec. F or EAL SU4.2, coolant samples exceeding the 2.0 &#xb5;Ci/gm dose equivalent l-131concentration require prompt action by DAEC Technical Specifications and are representative of minor fuel cladding degradation.
Escalation of the emergency classification level would be via ICs FAl or the Recognition Category A-R I Cs. DevelepeF Netes: For EAL #1 Enter the radiation monitor(s) that may be used to readily identify 1.vhen RCS activity levels e>rneed Technical Specification allovvable limits. This EAL may be developed using different method s and sites should use existing capabilities to address it (e.g., de;, elopment of new capabilities is not required).
E>rnmples of e>C.isting methods/capabilities include:
* An installed radiation monitor on the letdown system or air ejector.
* A hand held monitor or deployed detector reading with pre calculated conversion values or readily implementable conversion calculation capability.
223 l>ffil 99 O 1 (Re*,isioR e) November 2012 The monitor reading values should eorrespond to an RCS aetivity leve l approximate l y at , Teehnieal Specification allowable limits. If there is no e>dsting method/capability for determining this EAL , then it should not be included.
IC evaluation will be based on EAL #2. For EAL#2 Developers may reword the EAL to include the reactor coolant activity parameter(s) specified in Teehnieal Specifieations and the assoeiated allowable limit(s) (e.g., va l ues for dose equ i valent I 131 and gross activity, time dependent or transient va l ues, ete.). If this approach is selected, all RCS aetivity allowable limits should be ineluded. EGL Assignment Attributes:
3 .1.1.A and 3 .1.1.B 224 NEI 99 Ql (RevisioA
: 6) November 2012 SU4SU5 ECL: Notification of Unusual Event Initiating Condition:
RCS leakage for 15 minutes or longer. Operating Mode Applicability:
RCS leakage for 15 minutes or longer. Operating Mode Applicability:
l, 2, 3 Emergency Action Levels: SUS Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
Po*.ver Operation , Startup , Hot Standby , Hot Shutdownl..,_Ll Exem13le Emergency Action Levels: (1 or 2 or 3) Note: S 5.1 $2 ~3 The Emergency Director should declare the Unusual E~vent promptl y upon determining that the applicable time 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:
RCS unidentified or pressure boundary leakage greater than (site specific Yalue) 10 gpm for 15 minutes or longer. RCS identified leakage greater than (site specific Yalue)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:
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).
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~ steam generator tube leakage in a PWR) or a location outside of containment.
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.
The leak rate values for each EAL were selected because they are usually observable with nonnal Control Room indications.
Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation).
Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation).
EAL SUS.I uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage. The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification.
EAL SUS.I uses a lower value that reflects the greater significance of unidentified or pressure boundary l eakage. 225 tJEI 99 Ql (RevisieR C:i) tl0 1 1em0er 2Q 12 The release of mass from the RCS due to the as-designed
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.
/expected operation of a relief valve does not warrant an emergency classification.
Escalation of the emergency classification level would be via ICs of Recognition Category R or F. 101 ECL: Notification of Unusual Event Initiating Condition:
For PWRs , an emergency classification would be required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flo 1 t't' cannot be isolated). For BWRs , aA 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 E AL. 226 227 ~ffil 99 01 (RevisioR
Automatic or manual scram fails to shutdown the reactor. Operating Mode Applicability:
: 6) ~fo*remeer 2012 NEI 99 Gl (Re,*isioA
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.
: 6) l>fo*,*emeer 2G 12 The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage , if possible.
SU6.l SU6.2 a. An automatic scram did not shutdown the reactor. AND b. ANY of the following manual actions taken at 1C05 are successful in lowering reactor power below 5% power
Escalation of the emergency classification level would be via ICs of Recognition Category A-R or F. Develof)eF Notes: Ei\L #1 For the site specific leak rate 1 1alue , enter the higher of l O gpm or the Yalue specified in the site's Technical Specifications for this type of leakage. EAL #2 For the site specific leak rate ,,atue, enter the higher of 25 gpm or the value specified in the site's Technical Specifications for this type of leakage. For sites that haYe Technical Specifications that do not specify a leakage type for steam generator tube leakage , developers should include an EAL for tube leakage greater than 25 gpm for 15 minutes or longer. EGL Assignment Attributes:
3.1. l .,", 228 
 
l>JEI 99 0 I (RevisioA
: 6) l>fo\'emeer 2012 SU5SU6 ECL: Notification of Unusual Event Initiating Condition:
Automatic or manual (trip [PWR] / scram [BWR]) fails to shutdown the reactor. Operating Mode Applicability:
Power Operationl,2 Nate: A manual action is any operator action , or set of actions , \Yhich 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.
Exemf)le Emergency Action Levels: (1 or 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.
SU6.1 a. An automatic (trip [PWR] / scram [B'.VR]) did not shutdown the reactor. S 6.2 AND b. ANY of the following manual actions taken at 1C05 are successful in lowering reactor power below 5% power a.
* Manual Scram Pushbuttons
* Manual Scram Pushbuttons
* Mode Switch to Shutdown
* Mode Switch to Shutdown
* Alternate Rod Insertion (ARI) a. A manual scram did not shutdown the reactor. AND b. EITHER of the following:
* Alternate Rod Insertion (ARI)A subsequent manual action taken at the reactor control consoles (I C05) is successful in shutting down the reactor. A manual trip ([P'.1/R]
: 1. ANY of the following subsequent manual actions taken at 1 COS are successful in lowering reactor power below 5% power
/ scram [B'.1/R])
did not shutdown the reactor. AND b. EITHER of the following: 1. -ANY of the following subsequent manual actions taken at l COS are successful in lowering reactor power below 5% power
* Manual Scram Pushbuttons
* Manual Scram Pushbuttons
* Mode Switch to Shutdown
* Mode Switch to Shutdown
* Alternate Rod Inse1iion (ARI) OR 2. A subsequent automatic scram is successful in shutting down the reactor. Definitions:
* Alternate Rod Insertion (ARI)A subsequent manual action taken at the reactor control console (1 C05)s is successful in shutting down the reactor. ---__ OR 230 Definitions:
None Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic scram is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant. Following the failure on an automatic reactor scram, operators will promptly initiate manual actions at the reactor control console to shutdown the reactor (e.g., initiate a manual reactor scram quickly fall to a level within the capabilities of the plant's decay heat removal systems. 102 If an initial manual reactor scram is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control console to shutdown the reactor (e.g., initiate a manual reactor scram using a different switch). Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. A manual action at the reactor control console is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies.
Basis: NEI 99 0 I (Re*,isioA
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".
: 6) ~lovemeer 2012 2. -A subsequent automatic (trip [PWR] / scram [BWR]) is successful in shutting down the reactor. This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor~ [PWR] I scram [BWR]) that results in a reactor shutdown , and either a subsequent operator manual action taken at the reactor control consoles or an automatic (trip [PWR] I scram [BWR]) is successful in shutting down the reactor._ This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant. Following the failure on an automatic reactor (trip [PWR] I scram [BWR]), operators will promptly initiate manual actions at the reactor control console s to shutdown the reactor (e.g., initiate a manual reactor (trip [PWR] I scram [BWR])). If these manual actions are successful in shutting dov,rn the reactor , core heat generation willscram quickly fall to a level within the capabilities of the plant's decay heat removal systems. 231 
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.
---------. ------------------~JBI 99 g I (ReYisioA e) November 2012 If an initial manual reactor (trip [PWR] I scram [BWR]) is unsuccessful , operators will promptly take manual action at another location(s) on the reactor control console s to shutdown the reactor (e.g., initiate a manual reactor (trip [PWR] I scram [BWR])) using a different switch). Depending upon several factors, the initial or subsequent effort to manually (trip [PWR] I scram [BW~]) the reactor , or a concurrent plant condition , may lead to the generation of an automatic reactor ftrtp [PWR] I scram [BWR]) signal. If a subsequent manual or automatic (trip [PWR] I scram [BWR]) is successful in shutting down the reactor , core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. A manual action at the reactor control console s 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 ftrtp [PWR] I scram [BW1"])). This action does not include manually driving in control rods or implementation of boron injection strategies.
* 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.
Actions taken at back-panels or other locations with in the Contro l Room , or any loc ation outs ide the Contro l Room , are not cons id ered to be " at the reactor control console s". Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. [BWR] The plant response to the failure of an automatic or manual reactor (trip [PWR] / scram [BWR]) will vary based upon several factors including the reactor power level prior to the event , availability of the condenser , performance of mitigation equipment and actions , other concurrent plant conditions , etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutt in g down the reactor , then the emerge n cy classification l evel wi ll escalate to an Alert via IC SA:)&sect;. Depending upon the plant response , escalation is a l so possible via IC FAl. Absent the plant conditions needed to meet either IC SA:)&sect; or FAI , 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).A reactor shutdovm is determined in accordance
* 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.
*.vith applicable Emergency Operating Procedure criteria.
103 ECL: Notification of Unusual Event Initiating Condition:
Should a reactor (trip [PWR] / scram [BWR]) signa l be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied. I
Loss of ALL onsite or offsite communications capabilities.
* If the signal causes a plant transient that should have included an automatic reactor ftrtp [PWR] / scram [BWR]) and the RPS fai l s to automatically shutdown the reactor , then this IC and the EALs are applicab l e, and should be eva lu ated. If the signa l does not cause a plant trans i ent and the (trip [PWR] / scram [BWR]) failure is determined through other means (e.g., assessment oftest results), then this IC and the EALs are not applicab l e and no classification is warranted.
Operating Mode Applicability:
De*1el013er Netes: This IC is applicable in any Mode in which the actual reactor power level could eJrneed the power level at which the reactor is considered shutdown.
1, 2, 3 . Emergency Action Levels: SU7.l Loss of ALL of the following onsite communication methods:
A PWR with a shutdown reactor power level that is less than or equal to the reactor power level which defines the lov,er bound of Pov,er Operation (Mode 1) will need to include Startup (Mode 2) in the Operating Mode Applicability.
* Plant Operations Radio System
For example , if the reactor is considered to be shutdov,rn at 3% and Power Operation starts at >5%, then the IC is also applicable in Startup Mode. Developers may include site specific EOP criteria indicative of a successful reactor shutdown in 232 aA EAL staterneAt , the Basis or aoth (e.g., a reactor pov,cer leYel). wm 99 Ql (ReYisioR a) NoYemeer 2Q12 The term " reactor coAtrol coAsoles" may ae replaced with the appropriate site specific term (e.g., rnaiA eoAtrol aoards). EGL AssigArneAt Attriautes:
3.1.1.A 233 
 
1>,'EI 99 QI (Re1, 1 isioA e) l>io't'emeer 2Q 12 SU6SU7 ECL: Notification of Unusual Event Initiating Condition:
Loss of al-I-ALL onsite or offsite communications capabilities. Operating Mode Applicability:
Power Operation , Startup, Hot Standby , Hot Shutdown.L..LJ Example Emergency Action Levels: (1 or 2 or 3) S 7.1 .fill.L 2 S 7.3 Basis: Loss of ALL of the following onsite communication methods: * (site specific list of communications methods) Plant Operations Radio System
* In-Plant Phone System
* In-Plant Phone System
* Plant Paging System (Gaitronics)
* Plant Paging System (Gaitronics)
SU7 SU7.2 Loss of ALL of the following offsite response organization communications methods: SU7.3 Basis:
Loss of ALL of the following GRGoffsite response organization communications methods: _* _(site specific list of communications methods) DAEC All-Call phone
* DAEC All-Call phone
* All telephone lines (PBX and commercial)
* All telephone lines (PBX and commercial)
* Cell Phones (including fixed cell phone system)
* Cell Phones (including fixed cell phone system)
* Control Room fixed satellite phone system
* Control Room fixed satellite phone system
* FTS Phone system Loss of ALL of the following NRC communications methods:
* FTS Phone system Loss of ALL of the following NRC communications methods: _* _(site specific list of communications methods) FTS Phone system
* FTS Phone system
* All telephone lines (PBX and commercial)
* All telephone lines (PBX and commercial)
* Cell Phones (including fixed cell phone system)
* 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.
* Control Room fixed satellite phone system This IC addresses a significant loss of on-site or offsite communications capabilities.
While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to offsite response organizations and the NRC. This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.). EAL SU7 .1 addresses a total loss of the communications methods used in support of routine plant operations.
While not a direct challenge to plant or personnel safety , this event warrants prompt notifications to GRGoffsite response organization s 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.). 235 
EAL SU7 .2 addresses a total loss of the communications methods used to notify all offsite resp'onse organizations of an emergency declaration.
 
The offsite response organizations referred to here are the State of Iowa, Linn County, and Benton County. EAL SU7 .3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
l>IEI 99 0 I (Re*,isioA
104 SA1 ECL: Alert Initiating Condition:
: 6) NoYemeer 2012 EAL SU7.l addresses a total l oss 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 GRGoffsite response organization s of an emergency declaration.
Loss of ALL but one AC power source to essential buses for 15 minutes or longer. Operating Mode Applicability:
The GRGoffsite response organization s referred to here are-the S t ate of Iowa, Linn County, and Benton County (see Developer Notes). ---EAL SU7.3 addresses a tota l loss of the communications methods used to notify the NRC of an emergency declaration.
1, 2, 3 Emergency Action Level: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
DeYelopeF Notes: EAL #1 The " site specific list of communications methods" should include all communications methods used for routine plant communications (e.g., commercial or site telephones , page party systems , radios , etc.). This listing should include installed plant equipment and components , and not items owned and maintained by individuals. EAL #2 The " site specific list of communications methods" should include all communications methods used to perform initial emergency notifications to OROs as described in the site Emergency Plan. The listing should include installed plant equipment and components, and not items ovmed and maintained by individuals.
SAl.1 a. AC power capability to 1A3 and 1A4 buses is reduced to a single power source for 15 minutes or longer. AND b. ANY additional single power source failure will result in a loss of ALL AC power to SAFETY SYSTEMS. Definitions:
Ei>cample methods are ring dovm/dedicated telephone lines , commercial telephone lines , radios , satellite telephones and internet based communications technology. In the Basis section , insert the site specific listing of the OROs requiring notification of an emergency declaration from the Control Room in accordance 1 Nith the site Emergency Plan , and typically within 15 minutes. EAL #3 The "site specific list of communications methods" should include all communications methods used to perform initial emergency notifications to the NRG as described in the site E,mergency Plan. The listing should include installed plant equipment and components , and not items ovmed and maintained b)' individuals. These methods are typically the dedicated Emergency Notification System (m-18) telephone line and commercial telephone lines. EGL Assignment Attributes:
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.
3 .1.1.C 237 
Basis: This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.
 
This IC provides an escalation path from IC SUl. An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.
1'JE I 99 QI (R ev i s i o R e) Novembe r 2Q 1 2 SU7 ECL: Notification of Unusual Event Initiating Condition:
* A loss *of all offsite power with a concurrent failure of all but one emergency power source ( e.g., an onsite diesel generator).
Failure to isolate containment or loss of containment pressure control. [P\"JR] Operating Mode Applicability:
* A loss of emergency power sources (e.g., onsite diesel generators) with a single train of essential buses being fed from an offsite power source. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. Escalation of the emergency classification level would be via IC SSL 105 SA3 ECL: Alert Initiating Condition:
Po,.*,er Operation, Startup, Hot Standby, Hot Shutdov.*n Example Emergency Action Levels: (1 or 2) 1 a. Failure of containment to isolate 'A'hen required by an actuation signal. AND b. ALL required penetrations are not closed within 15 minutes of the actuation signal. 2 a. Containment pressure greater than (site specific pressure).
UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.
AND b. Less than one full train of (site specific system or equipment) is operating per design for 15 minutes or longer. Basis: Th is IC addresses a failure of one or more containment penetrations to automatically isolate (close) '.\.'hen required by an actuation signal. It also addresses an event that results in high containment pressure ,,,ith a 239 NE I 9 9 0 I (R ev i s i o R a) No,*embe r 20 1 2 concurrent failure of containment pressure control systems. Absent challenges to another fission product barrier, either condition represents potential degradation of the level of safety of the plant. For EAL 1, the containment isolation signal must be generated as the result on an off normal/accident condition (e.g., a safety injection or high containment pressure);
a failure resulting from testing or maintenance does not warrant classification.
The determination of containment and penetration status isolated or not isolated should be made in accordance v.*ith the appropriate criteria contained in the plant AOPs and EOPs. The 15 minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.
EAL 2 addresses a condition
'Nhere containment pressure is greater than the setpoint at '.*Jhich containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15 minute criterion is included to allo'I-' operators time to manually start equipment that may not have automatically started, if possible.
The inability to start the required equipment indicates that containment heat removal/depressurization systems (e.g., containment sprays or ice condenser fans) are either lost or performing in a degraded manner. This event ,,.,ould escalate to a Site Area Emergency in accordance
,,.,ith IC FS1 if there ,.,.,ere a concurrent loss or potential loss 240 1>IBI 99 QI (Re,*isioA e) "!>lo'iemeer 2Q 12 of either the Fuel Clad or RCS fission product barriers.
Developer Notes: Enter the "site specific pressure" value that actuates containment pressure control systems (e.g., containment spray). Also enter the site specific containment pressure control system/equipment that should be operating per design if the containment pressure actuation setpoint is reached. If desired, specific condition indications such as parameter values can also be entered (e.g., a containment spray flow rate less than a certain value). EAL #2 is not applicable to the U.S. Evolutionary Po'..ver Reactor (EPR) design. Attributes:
241 ECL Assignment 3.1.1.A 242 ~m, 99 Ql (Re,*isioR e) ~fovemaer 2()12 Ne! 99 0 I (ReYisioA
: 6) :Jlolo 1 ,em0er 2012 SA1 ECL: Alert Initiating Condition:
Loss of al-l-ALL but one AC power source to emergency essential buses for 15 minutes or longer. Operating Mode Applicability:
Po 1 Ner Operation , Startup , Hot Standby , Hot Slrntdown.1.1..,_J.
Example Emergency Action Leve ls: Note: The Emergency Director should declare the Alert-event promptly upon determining that the applicable time 15_ minutes has been exceeded, or will likely be exceeded. S 1.1 a. AC power capability to (site specific emergency buses) 1A3 and 1 A4 buses is reduced to a single power source for 15 minutes or longer. AND b. Any /\NYANY additional single power source failure will result in a loss of alt ALL AC power to SAFETY SYSTEMS. Definitions:
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdovm condition.
including the EGGS. Systems classified as safety related. Basis: This IC describes a significant degradation of off site and onsite AC power so urces such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the so l e AC power source may be powering one, or more than one , train of safetyrelated equipment.
This IC provides an esca l ation path from IC SUL An "AC power source" is a source recognized in AOPs and EOPs , and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.
* A loss of all offsite power with a concurrent failure of all but one emergency power source ( e.g., an onsite diesel generator).
* A loss of all offsite power and loss of all emergency pO'Ner sources (e.g., onsite diesel generators) 1 Nith a single train of emergency buses being back fed from the unit main generator.
* A loss of emergency power sources (e.g., onsite diesel generators) with a sing l e train of essentialemergency buses being -eaek-fed from an offsite power source. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. Escalation of the emergency classification level would be via IC SSl. De*;eleper Notes: For a po 1 Ner source that has multiple generators , the EAL and/or Basis section should reflect the minimum number of operating generators necessary for that source to provide required power to 243 Nel 99 QI (Re\1 isioA a) l>lo*,*eme er 2Q 12 an AC emergency bus. For e>rnmple , if a backup pov,er source is comprised of two generators (i.e., two 50% capacity generators sized to feed 1 AC emergency bus), the EAL and Basis section must specify that both generators for that source are operating.
The " site specific emergency buses" are the buses fed by offsite or emergency
/\C po 1 Ner sources that supply power to the electrical distribution system that powers SAFETY 8&#xa5;8TEM8. There is typically 1 emergency bus per train of SAFETY 8&#xa5;8TEM8. Developers should modify the bulleted e>rnmples provided in the basis section , above , as needed to reflect their site specific plant designs and capabilities.
The EALs and Basis should reflect that each independent offsite pov,er circuit constitutes a single po:wer souroe. For e>rnmple, three independent 345kV offsite power circuits (i.e., incoming power lines) comprise three separate power sources. Independence may be determined from a revie\\' of the site specific UF8AR , 8BO analysis or related loss of electrical power studies. The EAL and/or Basis section may specify use of a non safety related power source provided that operation of this sou roe is recognized in AOPs and EOPs , or beyond design basis accident response guidelines (e.g., FLEX support guidelines).
8uch power sources should generally meet the " Alternate ac souroe" definition provided in 10 CFR 50.2. At multi unit stations , the E ALs may credit compensatory measures that are proceduralized and can be implemented within 15 minutes. Consider capabilities such as power source cross ties, " sv,*ing" generators , other power sources described in abnormal or emergency operating procedures , etc. Plants that have a proceduralized capability to supply offsite AC power to an affected unit Yia a cross tie to a companion unit may credit this power source in the EAL provided that the planned cross tie strategy meets the requirements of 10 CFR 50.63. EGL Assignment Attributes:
3.1.2.B 244 NEI 99 01 (Re\1 isioA e) ~fovemeer 2012 SA2SA3 ECL: Alert Initiating Condition:
UNPLANNED loss of Contro l Room indications for 15 minute s or lon ger wit h a s i gnificant transient in progress.
Operating Mode Applicability:
Operating Mode Applicability:
1, 2, 3 Emergency Action Level: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
Power Operation , Startup , Hot Standby , Hot Shutdovml.,_1,_l Example Emergency Action Level s: Note: The Emergency Director should declare the A-left-event promptly upon determining that the applicable time 15_ minutes has been exceeded , or w ill likely be exceeded.
SA3.1 a. An UNPLANNED event results in the inability to monitor one or more Table S-1 parameters from within the Control Room for 15 minutes or longer.
S 3.1 a. An UNPLANNED event results in the inability to monitor one or more Table S-1 parameters from within the Control Room for 15 minutes or longerAn UNPLANl'H3D e'rent results in the inability to monitor one or more of the follo 1 Ning parameters from within the Control Room for 15 minutes or longer. ......-Table S-1 Safety System Parameters
* Reactor power
* Reactor power
* RPV Water Level
* RPV Water Level
Line 892: Line 399:
* Primary Containment Pressure
* Primary Containment Pressure
* Suppression Pool Level
* Suppression Pool Level
* Suppression Pool Temperature AND b. ANY of the Table S-2 transient events are in progress.
* Suppression Pool Temperature PeweF RPV Water Level R1)V Pressure Primary Containment Pressure Suppression Pool Level Suppression Pool Temperature Suppression Pool Temperature AND -Reactor ___ b __ . __ ANY of the Table S-2 transient events are in progress.
T 1ble 5:0.2 'Sig'nificaht Transients:, -',.,"' ,, f.._ ' -_;.., "~;,,
245 
-Table S-2 Significant Transients
* Automatic or manual runback greater than 25% thermal reactor power
* Automatic or manual runback greater than 25% thermal reactor power
* Electrical load rejection greater than 0; 25% full electrical load C, :' 11
* Electrical l oad rejection greater than 25% full electrical load
* Reactor scram ECCS actuation R
* Reactor scram
* I'
* ECCS actuation
* Thermal power oscillations greater ti: than 10% 106 Definitions:
* Thermal power oscillations greater than 10% transient eyents in progress.
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.
}>JEI 99 QI (Re\*isioA
UNPLANNED:
: 6) }>Jo\*emeer 2Q 12 of the following Automatic or manual runback greater than 25% thermal reactor power Electrical load rejection greater than 25% full electrical load Reactor scram [BWR] / trip [PWR] EGGS (81) actuation Thermal power oscillations greater than 10% [BWR] 246 Definitions:  
~JEJ 99 0 I (Re\*isieA e) ~Je\1 eme er 2012 SAFETY SYSTEM: A system required for safe plant operation, coo l ing down the p l ant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.A system required for safe plant operation, cooling dov-m the plant and/or placing it in the cold shutdov,rn condition, including the EGGS. 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.
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. Basis: This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant. As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s  
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. 247 NEI 99 0 I (RevisioA
). 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.
: 6) ])ofo&#xa5;emeer 2012 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 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.
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.
In particular , emergency assessments necessary to implement abnormal operating procedures , emergency operating procedures , and emergency plan implementing procedures addressing emergency classification , accident assessment , or protective action decision-making.
This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV level and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition.
This EAL is focused on a selected subset of p l ant parameters associated with the key safety functions of reactivity control , core cooling [PWR] I RPV level [BWR] and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition.
In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for 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.
In addition , if all indication sources for one or more of the listed parameters are lost, then the ability to determ i ne the values of other SAFETY SYSTEM parameters may be impacted as well. For example , if the value for reactor vessel level [PW~] I RPV water level [BWR] cannot be determined from the indications and recorders on a main control board , the SPDS or the p l ant 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 I Cs FS 1 or IC AS+RS 1. Developer Notes: In the PWR parameter list column , the " site specific number" should reflect the minimum number of steam generators necessary for plant cooldown and shutdown. This criterion may also specif)' \\'Aether the level value should be vride range, narrow range or both , depending upon the monitoring requirements in emergency operating procedures. Developers may specify either pressurizer or reactor vessel level in the P'.VR parameter column entry for RCS Level. Developers should consider if the "transient events" list needs to be modified to better reflect site specific plant operating characteristics and eJ(pected responses.
Escalation of the emergency classification level would be via ICs FSl or IC RSl. 107 SAG ECL: Alert Initiating Condition:
The number, type , location and layout of Control Room indications , and the range of possible failure modes , can challenge the ability of an operator to accurately detennine , within the time period available for emergency classification assessments , if a specific percentage of indications have been lost. The approach used in this EAL facilitates prompt and accurate emergency classification assessments by focusing on the indications for a selected subset of parameters.
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:
By focusing on the availability of the specified parameter values , instead of the sources of those values , the EAL recognizes and accommodates the 1.,*ide variety of indications in nuclear power plant Control Rooms. Indication types and sources may be analog or digital, safety related or not , primary or alternate , individual meter value or computer group display , etc. 248 Ne! 99 QI (RevisioR e) Jlolovemeer 2Q 12 A loss of plant annunciators will be evaluated for reportability in accordance with 10 CPR 50.72 (and the associated guidance in NUREG 1022), and reported if it significantly impairs the capability to perform emergency assessments.
1, 2 Note:
Compensatory measures for a loss of annunciation can be readily implemented and may include increased monitoring of main control boards and more frequent plant rounds by non licensed operators.
* 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.
Their alerting function notwithstanding , annunciators do not provide the parameter values or specific component status information used to operate the plant , or process through AOPs or EOPs. Based on these considerations , a loss of annunciation is considered to be adequately addressed by reportability criteria , and therefore not included in this IC and EAL. With respect to establishing event severity , the response to a loss of radiation monitoring data (e.g., process or effluent monitor values) is considered to be adequately bounded by the requirements of 10 CPR 50.72 (and associated guidance in NUREG 1022). The reporting of this event will ensure adequate plant staff and NRG awareness , and drive the establishment of appropriate compensatory measures and corrective actions. In addition , a loss of radiation monitoring data , by itself , is not a precursor to a more significant event. Personnel at sites that have a Failure Modes and Effects Analysis (FMEA) included within the design basis of a digital J&C s;'stem should consider the FMEA information when developing their site specific EALs. Due to changes in the configurations of 8AFETY 8&#xa5;8TEM8 , including associated instrumentation and indications , during the cold shutdown , refue l ing , and defueled modes , no analogous IC is included for these modes of operation.
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
EGL Assignment Attributes:
3.1.2.B 249 
 
ECL: Alert }ffil 99 0 I (ReYisioA
: 6) }lo\1 emeer 2012 SA5SA6 Initiating Condition:
Automatic or manual (trip [P'.l/R] / scram [BWR]) fails to shutdown the reactor , and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor. Operating Mode Applicability:
Pov,er Operation.1.....1 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. l E,ample EmeFgeeeyEmergency Action Level s: 1 a. An automatic or manual (trip [PWR] / scram [BWR]) did not shutdown the reactor. AND b. ALL of the following manual actions taken at 1 COS are not successful in lowering reactor power below 5% power
* Manual Scram Pushbuttons
* Manual Scram Pushbuttons
* Mode Switch to Shutdown
* Mode Switch to Shutdown
* Alternate Rod Insertion (ARI) Definitions:
* Alternate Rod Insertion (ARI)Manual actions taken at the reactor control consoles (1 COS) are not 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 subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful.
Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor ftrtJ3 [PWR] I s cram [BWR]) that results in a reactor s hutdown , 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.
This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emer g ency declaration is required even if the reactor is s ubsequently 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 ftrtJ3 [PWR] I scram [BWR])). This action does not include manually driving in control rods or implementation of boron injection strategies.
If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control consoles ( e.g., locally opening breakers).
If this action(s) is unsuccessful , operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers).
Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles." Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. 108 The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shutdown the reactor is prolonged enough to cause a challenge to the RPV water level or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS6. Depending upon plant responses and symptoms, escalation is also possible via IC FSl. Absent the plant conditions needed to meet either IC SS6 or FSl, an Alert declaration is appropriate for this event. It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.
_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.,_''-:-251 NEI 99 01 (RevisioA a) l>fovemeer 2012 Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. [BWR] 252 
The reactor should be considered shutdown when it is producing less heat than the maximum decay heat load for which the SAFETY SYSTEMS are designed (typically 3 to 5% power). 109 SAS ECL: Alert Initiating Condition:
 
Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. Operating Mode Applicability:
NEI 99 0 I (RevisioA
1, 2, 3 Emergency Action Level: Notes: SA8.l
: 6) ]I,[ 0\'emeer 2012 The plant response to the failure of an automatic or manual reactor (trip [PWR] I scram [BWR]) will vary based upon several factors including the reactor power level prior to the event , availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions , etc. If the failure to shutdown the reactor is prolonged enough to cause a challenge to the core cooling [PWR] / RPV water level [BWR] or RCS heat removal safety functions , the emergency classification level will escalate to a Site Area E mergency via IC SS~&sect;. Depending upon plant responses and symptoms, escalation is also possible via IC FSl. Absent the plant conditions needed to meet either IC SS~&sect; or FSl, an Alert declaration is appropriate for this event. It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however , this IC and EAL are included to ensure a timely emergency declaration. The reactor should be considered shutdown when it is producing less heat than the maximum decay heat load for which the SAFETY SYSTEMS are designed (typically 3 to 5% power).A reactor shutdovm is determined in accordance
* 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.
*.vith applicable Emergency Operating Procedure criteria.
* 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.  
Develeper Netes: This IC is applicable in any Mode in which the actual reactor power level could exceed the pov,rer level at which t he reactor is considered shutdown.
: a. AND b. The occurrence of ANY of the Table S-3 hazardous events: Ir > T;>ble S-:}-Ha~ardo;,s Events~--1 1 !-----**-**----*---** . **---*--*-**-. --1.
A PWR *.vith a shutdown reactor power level that is less than or equal to the reactor power level which defines the lower bound of Pmver Operation (Mode l) will need to include Startup (Mode 2) in the Operating Mode Applicability. For example , if the reactor is considered to b e shutdovm at 3% and Power Operation starts at >5%, then the IC is also applicable in Startup Mode. Developers may include site specific EOP criteria indicative of a successful reactor shutdown in an EAL statement , the Basis or both (e.g., a reactor power level). The term " reactor control consoles" may be replaced v r ith the appropriate site specific term (e.g., main control boards). EGL Assignment Attributes
* Seismic event (earthquake)
: 3.1.2.B 254 
 
Nm 99 QI (ReYisioR
: 6) ~loYemller 2Q 12 SA9SA8 E CL: A l ert Initiating Condition:
Hazar d o u s eve nt affec tin g a SAFETY SYSTEM n ee d e d for th e c u rre nt o p e ratin g m o d e. Operatin g Mode Applicability:
Power Operation , Startup , Hot Standby , Hot Shutdovm.1..1.,_]_
Example Emergenc y A ction Le v el s: Notes: S 8.1
* If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardot1s event occurred, then this emergency classification is not warranted. -If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of the SAFETY SYSTEM, then this emergency classification is not warranted .
* a. AND b. The occurrence of ANY of the Table S-3 hazardous events: ----l Table S-3 Hazardous Events -* Seismic event (earthquake)
* Internal or external flooding event
* Internal or external flooding event
* High winds or tornado strike
* High winds o r t ornado strike
* FIRE
* FIRE
* EXPLOSION
* EXPLOSION
* Other events with similar hazard characteristics as determined by the Shift Manager or Emergency Director Event damage has caused indications of degraded performance in one train of a SAFETY SYSTEM needed for the current operating mode. AND 2. EITHER of the following:
* Othe r events w i th s i milar hazard c h aracteristics as determined by the Sh i ft Manager or Emergency Director Director 1. 2. Event damage has caused indications of degraded performance in one train of a SAFETY SYSTEM needed for the current operating mode. AND 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
256 
* The event has resulted in VISIBLE DAMAGE to the second train of a SAFETY SYSTEM needed for the current operating mode. 110 Definitions:
~IEJ 99 Q I (Re,,isioA
EXPLOSION:
: 6) *Novemeer 2Q 1 2
* Event damage has caused indications of degraded performance to a second train of t h e SAFETY SYSTEM needed for the current operating mode,
* The event has resulted in VfSTBLE DAMAGE to the second train of a SAFETY SYSTEM needed for the current operating mode.;: E 1 ,rent damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode, or The eyent has resu l ted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode. Loss of the safety function of a single train SAFETY SYSTEM. 257 
*------------------------Definitions:
NEI 99 0 I (ReYisioA e) 1-Jo,,ember 20 12 EXPLOSION:
A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization.
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.
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.
Such events may require a post-event inspection to determine if the attributes of an explosion are present. FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, inc l uding the ECCS. These systems are classified as safety-related.A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdov,rn condition, including the BCCS. Systems classified as safety related. VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis.
VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis.
The visua l impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.
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. EITHER of the following:
Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not VISIBLE DAMAGE. Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of perfonnance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded perfonnance for criteria SA8.l.b.l of this EAL; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events. An event affecting a single-train SAFETY SYSTEM (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the one train) would not be classified under SA8 because the two-train impact criteria that underlie the EALs and Bases would not be met. If an event affects a single-train SAFETY SYSTEM, then the emergency classification should be made based on plant parameters/symptoms meeting the EALs for another IC. Depending upon the circumstances, classification may also occur based on Shift Manager/Emergency Director judgement.
: 1. event damage has caused indications of degraded performance in at least one train of a SAFBTY 8Y8TBM needed for the current operating mode. OR 2. The e1,*ent has caused V18IBLB DAMAGE to a SAFBTY 8&#xa5;8TBM component or structure needed for the current operating mode. 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 SA98.1.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. An event affecting a single-train SAFETY SYSTEM (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the one train) would not be classified under SA8 because the two-train impact criteria that underlie the EALs and Bases would not be met. If 2 58 Ne! 99 01 (Re;1 isioR e) ~J o*temeer 2012 a n event affects a single-train SAFETY SYST E M, then the emergency cl ass ification should be made based on plant parameters
Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available.
/symptoms meeting the E ALs for another TC. Depending upon the circ u mstances, classificatio n may also occur based o n Shift Manager/Emergency Director judgement.
The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. 111 VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make this determination based on the totality of available event and damage report information.
Ind i cations of degraded performance addresses damage to a SAFETY SYSTEM tra i n that is i n serv i ce/operation s i nce indications for it wi ll be readily available.
This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. Escalation of the emergency classification level would be via IC FSl or RSl. 112 SS1 ECL: Site Area Emergency Initiating Condition:
The indications of degraded performance should be significa n t enough to cause concern regarding the operabi l ity or reliability of the SAF E TY SYSTEM train. 259 
Loss of ALL offsite and ALL onsite AC power to essential buses for 15 minutes or longer. Operating Mode Applicability:
 
1, 2, 3 Emergency Action Level: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
~JEI 99 QI (ReYisioA e) No,*ember 2Q 12 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.
SSl.1 Loss of ALL offsite and ALL onsite AC power to 1A3 and 1A4 buses for 15 minutes or longer. Definitions:
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. 261 Tl-IE! 99 0 I (Re,,*isioA
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.
: 6) November 2012 This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier , and therefore represents an actual or potential substantial degradation of the level of safety of the EAL l.b.l addresses damage to a SAFETY SYSTEM train that is in serYice/operation since indications for it 1,vill be readily available.
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.
The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. EAL 1.b.2 addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components.
This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via ICs RGl, FGl or SGl. 113 ECL: Site Area Emergency Initiating Condition:
Operators will make this determination based on the totality of available eYent and damage report information.
This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. Escalation of the emergency classification level would be via IC FS 1 or Ml-RS 1. Develeper Netes: 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 t>.&#xa5;0 or more separate and redundant trains of equipment in accordance with site specific design criteria.
EGL Assignment ,r\ttributes:
3.1.2.B 262 ECL: Site Area Emergency NEI 99 0 I (Re*,isioR
: 6) No*,emeer 2012 551 Initiating Condition:
Loss of ALLa-lt offsite and al-I-ALL onsite AC power to emergency essential-buses for 15 minutes or longer. Operating Mode Applicability:
Pov,*er Operation , Startup , Hot Standby , Hot Shutdownl....1.J Example Emergency Action Level s: Note: The Emergency Director should declare the Site Area Emergencyevent promptly upon determining that the applicable time 15 minutes has been exceeded , or will likely be exceeded.
S 1.1 Loss of ALL offsite and ALL onsite AC power to (site specific emergenC)' buses) 1 A3 and 1A4 buses for 15 minutes or longer. Definitions:
SAFETY SYSTEM: A s ystem 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.A system required for safe plant operation, cooling do 1.vn the plant and/or placing it in the cold shutdovm condition, including the EGGS. Systems cla s sified a s safety related. Basis: This IC addresses a total loss of AC power that compromises the performance of a ll 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. Esca l ation of the emergency c l assification l eve l wou ld be via I Cs AG+RG 1 , FG I or SG I. De*;eloper Notes: For a power source that has multiple generators , the EAL and/or Basis section should reflect the minimum number of operating generators necessary for that source to provide adequate power to an AC emergency bus. For e>rnmple , if a backup power source is comprised of two generators (i.e., tv,o 50% capacity generators sized to feed 1 AC emergency bus), the E AL and Basis section must specify that both generators for that source are operating.
The " site specific emergency buses" are the buses fed by offsite or emergency AC power sources that supply power to the electrical distribution system that powers SAFETY SYSTEMS. There is typically l emergenC)' bus per train of SAFETY SYSTEMS. 263 _J NBT 99 0 I (RevisioA
: 6) November 2012 The EAL and/or Basis section rnay specify use of a non safety related power source pro 1 1ided that operation of this source is controlled in accordance with abnorrnal or ernergency operating procedures , or beyond design basis accident response guidelines (e.g., FLEX support guidelines).
Such power sources should generally rneet the "Alternate ac source" definition provided in 10 GFR 50.2. At rnulti unit stations , the EA.Ls rnay credit compensatory rneasures that are proceduralized and can be irnplernented within 15 minutes. Consider capabilities such as power source cross ties, " swing" generators , other po 1.Yer sources described in abnormal or ernergency operating procedures , etc. Plants that have a proceduralized capability to supply offsite AG power to an affected unit via a cross tie to a cornpanion unit may credit this po 1.ver source in the EAL provided that the planned cross tie strategy meets the requirernents of 10 GFR 50.63. EGL Assignrnent Attributes:
3.1.3.B 264 ECL: Site Area Emergency  
~JEI 99 0 I (RevisioA
: 6) November 2012 SS8SS2 Initiating Condition:
Loss of ALL Vital DC power for 15 minutes or longer. Operating Mode Applicability:
Loss of ALL Vital DC power for 15 minutes or longer. Operating Mode Applicability:
1, 2, 3 Emergency Action Level: SS2 Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
l, 2, 3 Emergency Action Levels: Note: The Emergency Director should declare the Site Area Emergenc)'event promptly upon determining that the applicable time 15 minutes has been exceeded, or will likely be exceeded.  
SS2.l Indicated voltage is less than 105 VDC on BOTH Div 1 and Div 2 125 VDC buses for 15 minutes or longer. Definitions:
~S+=-2~. l __ Indicated voltage is less than (site specific bus voltage value) 105 VDC on ALL(site specific Vital DC busses) BOTH Div 1 and Div 2 125 VDC buses for 15 minutes or longer. Definitions:
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
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.A system required for safe plant operation, cooling do,Yn the plant and/or placing it in the cold shutdown condition, including the EGGS. S)'stems classified as safet)' related. Basis: This IC addresses a loss of Vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown , this condition involves a major failure of plant functions needed for the protection of the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via I Cs AG-1-RG 1 , FG 1 or SG2. 265 
* Basis: This IC addresses a loss of Vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via I Cs RG 1, FG 1 or SG2. 114 556 ECL: Site Area Emergency Initiating Condition:
~ml 99 Ql (RtwisioA a) No*,emaer 2Q12 SS5SS6 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:
Inability to shutdown the reactor causing a challenge to (core cooling [PWR] I RPV water level [BWR]) or RCS heat removal. Operating Mode Applicability:
1, 2 Emergency Action Levels: SS6.1 a. An automatic or manual scram did not shutdown the reactor. AND b. ALL of the following manual actions taken at 1 COS are not successful in lowering reactor power below 5% power:
Power OperationLl Examf)le Emergency Action Levels: S 6.1 a. b. An automatic or manual (trip [PWR] / scram [BV/R]) did not shutdown the reactor. AND ALL of the following manual actions taken at 1 COS are not successful in lowering reactor power be l ow 5% power:
* Manual Scram Pushbuttons
* Manual Scram Pushbuttons
* Mode Switch to Shutdown
* Mode Switch to Shutdown
* Alternate Rod Insertion (ARI) AND c. EITHER of the following conditions exist:
* Alternate Rod Insertion (ARI)AII manual actions to shutdown the reactor have been unsuccessful.
* RPV level cannot be restored and maintained above -25 inches. OR
AND c. EITHER of the following conditions exist: Definitions:
* HCL (Graph 4 of EOP 2) exceeded.
Basis: _* _(Site specific indication of an inability to adequately remove heat from the core) Reactor vessel 'NaterRPV level cannot be restored and maintained above -25 inches. * (Site specific indication of an inability to adequately remove heat from the -RGSjHCL (Graph 4 of EOP 2) exceeded.
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor ftFtp [PWR] I scram [BWR]) that results in a reactor shutdown , all subsequent operator actions to manually shutdown the reactor are unsuccessful , and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.
In some instances , the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition 266 NE! 99 QI (ReYisieR
: 6) "t-Je&#xa5;ember 2Q 12 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 shou l d 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). A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria. Esca lati on of the emergency c l assification le ve l would be via IC AG+-RG 1 or FG 1. De*,relof)er Notes: This IC is applicable in any Mode in which the actual reactor povt'er level could e,rneed the power level at vt'hich the reactor is considered shutdown.
A PWR with a shutdown reactor power level that is less than or equal to the reactor power level which defines the lower bound of Power Operation (Mode 1) will need to include Startup (Mode 2) in the Operating Mode Applicability.
For e,rnmple , if the reactor is considered to be shutdown at 3% and Power Operation starts at >5%, then the IC is also applicable in Startup Mode. Developers may include site specific EOP criteria indicative of a successful reactor shutdown in an EAL statement, the Basis or both (e.g., a reactor power level). Site specific indication of an inability to adequately remove h e at from the core: [BWR] Reactor vessel water level cannot be restored and maintained above Minimum Steam Cooling RPV Water Level (as described in the EOP bases). [PWR] Insert site specific values for an incore/core e,cit thermocouple temperature and/or reactor vessel water level that drives entry into a core cooling restoration procedure (or otherwise requires implementation of prompt restoration actions).
Alternately , a site may use incore/core e J lit thermocouple temperatures greater than l , 200 6 F and/or a reactor vessel water level that corresponds to apprmcimatel)
' the middle of active fuel. Plants vt'ith reactor vessel level instrumentation that cannot measure down to approximately the middle of active fuel should use the lowest on scale reading that is not above the top of active fuel. If the lo'Nest on scale reading is above the top of active fuel , then a reactor vessel level value should not be included. For plants that have implemented Westinghouse Owners Group Emergency Response Guidelines , enter the parameters used in the Core Cooling Red Path. Site specific indication of an inability to adequately remove heat from the RCS: [BWR] Use the Heat Capacit)' Temperature Limit. This addresses the inability to remove heat via the main condenser and the suppression pool due to high pool water temperature.
[PWR] Insert site specific parameters associated with inadequate RCS heat removal via the steam generators. These parameters should be identical to those used for the Inadequate Heat Removal threshold Fuel Clad Barrier Potential Loss 2.B and threshold RCS Barrier Potential Loss 2.A in the PWR EAL Fission Product Barrier Table. EGL Assignment Attributes:
3.1.3.B 267 SS8 ECL: Site Area EmergeRcy Initiating Condition:
Loss of all Vital DC power for 15 miRutes or loRger. --------------~ 1%1 99 GI (ReYisioA
: 6) November 2G 12 Of)erating Mode Af)f)lieability:
Power OperatioR , Startup , Hot 8taRdb;*, Hot 8hutdovm1, 2, 3, 4 Examf)le Emergeney Aetion Levels: Note: The EmergeRcy Director should declare the Site Area EmergeRcy promptly upoR determiRiRg that 15 miRutes has beeR e)weeded, or will likely be e)weeded.
l lAdicated voltage is less thaR (site specific bus voltage value)l 15 VOC OR ALL (site specific Vital DC busses) 1(2) D O l, D 02, D 03, aRd D 04 for 15 miRutes or loRger. Basis+ SAFETY 8Y8TEM: /*, system required for safe plaRt operatioR, cooliRg dowR the plaRt aRd/or placiRg it iR the cold shutdovm coRditioR, iRcludiRg the EGGS. Systems classified as safety related. This IC addresses a loss of Vital DC power which compromises the ability to moRitor aRd coRtrol SAFETY 8Y8TEM8. IA modes above Cold 8hutdowR , this coRditioR iRvolves a major failure of plaRt fuRctioRs Reeded for the protectioR of the public. FifteeR miRutes was s~lected as a threshold to e,wlude traRsieRt or momentary power losses. EscalatioR of the emergeRcy classificatioR level 1 Nould be via ICs AGlB.Ql , FGl or 808. DeYelof)er Notes: The "site specific bus Yoltage value" should be based OR the miRimum bus voltage Recessary for adequate operatioR of SAFETY 8Y8TEM equipmeRt.
This voltage value should iRcorporate a margiR of at least 15 miRutes of operatioR before the oRset of iRability to operate those loads. This voltage is usually Rear the miRimum voltage selected wheR battery siziRg is performed. The typical value for aR eRtire battery set is appFO>(imately 105 VDC. For a 60 cell striRg of batteries , the cell voltage is apprmdmately 1.75 Volts per cell. For a 58 striRg battery set , the miRimum voltage is approximately 1.81 Volts per cell. The " site specific Vital DC busses" are the DC busses that provide moRitoriRg aRd coRtrol capabilities for SAFETY 8Y8TEM8. EGL AssigRmeRt Attributes:
3.1.3.B 268 1'IBI 99 O I (ReYisioR e) NoYemeer 2012 SG1 ECL: General Emergency Initiating Condition:
Prolonged loss of al-l-ALL offsite and ALLalt onsite AC power to emergency essential buses. Operating Mode Applicability
: Power Operation , Startup , Hot Standby , Hot ShutdovmL..1.,__J Example Emergency Action Level s: Note: The Emergency Director should declare the General Emergencyevent promptly upon determining that (site specific hours) the applicable time 4 hours has been exceeded, or will likely be exceeded. a. Loss of ALL offsite and ALL onsite AC power to 1A3 and 1 A4 specific emergency buses). AND b. EITHER of the following:
Definitions:
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
_* _Restoration of at least one AC emergency essential bus in less than specific hours)4 hours is not likely. OR * (Site specific indication of an inability to adequately remo*,e heat from the serejRPV level cannot be restored and maintained above -25 inches. 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.A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the EGGS. Systems classified as safety related. Basis: Th i s IC addresses a prolonged loss of all power sources to AC emergency 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/press u re 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 capabi l ities may be degraded under these cond i tions. 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 offsite protective actions. Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency essential bus by the end of the analyzed 4 hour station blackout coping period. Beyond this time , plant responses and event trajectory are subject to greater uncertainty, and there is an increased l i kelihood of challenges to multiple fission product barriers.
269 
~JEI 99 QI (ReYisioA e) ~Je
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.
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.
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 definitions of these terms are provided below. CONFINEMENT BOUNDARY: (Insert a site speeific definition for this term.) Developer Note -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.
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.
For DAEC, this is considered to be Secondary Containment as required by Technical Specifications.
CONTAINMilff CLOSURE: (Insert a site speeific definition for this term.) DeYeloper Note Site speeific proeedurally defined aetions taken to seeure eontainment and its assoeiated struetures, systems, and eomponents as a funetional barrier to fission product release under e>(isting plant eonditions.
For DAECs, this is considered to be Secondary Containment as required by Teehnieal Specifieations.The proeedurally defined eonditions or aetions taken to secure containment (primary or secondary for BWR) and its assoeiated struetures , systems , and eomponents as a functional barrier to fission product release under shutdovm eonditions.
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.
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:
EXPLOSION:
A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization.
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.
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.
Such events may require a post-event inspection to determine if the attributes of an explosion are present. F/\ULTED:
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.
The term applied to a steam generator that has a steam leak on the seeondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to beeome eompletely depressurized.
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.
Developer Note This term is applicable to PV/Rs only. B-2 
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.
Ne! 99 0 I (RevisioR
* 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.
: 6) ~/0 1 ,emaer 2012 FIRE: Combustion characterized by heat and li ght. Sources of smoke such as slipping drive belts or overheated electrica l eq uipm ent do not constitute FIRES. Observation of flame is preferred but is NOT required if l arge quantities of smoke and heat a r e observed. HOSTAGE: A person(s) held as l everage aga in st the stat i on to ensure that demands will be met by the station. HOSTILE ACTION: An act toward a NP-llnuclear power plant or its personnel that includes the use of violent force to destroy equipment , take HOSTAGES , and/or intimidate the lic ensee to ac hi eve an end._ This includes attack by air , land, or water using guns , explos i ves , PROJECTILEs , veh i cles , or other devices used to de li ver destructive force. _Other acts that sat isf y the overa ll int ent may be inc luded. HOSTILE ACTION shou ld not be construed to include acts of civil disobedience or fe l on iou s acts that are not part of a concerted attack on the NPJ!nuclear power plant._ Non-terrorism-b ased 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: This term is typically taken to mean the site property owned by or otherwise under the control of the licensee.
* HOSTILE FORCE: One or more individuals who are engaged in a determined assau l t , overtly or by stea lth and deception , equ ipp ed with su it ab l e weapo n s capable of killing , maiming , or causing destruction.
PROJECTILE:
IMMINENT:
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.
The trajectory of events or conditions is s u ch that an EAL will be met within a re lativel y 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. NOR~4AL LEVELS: As applied to radiologieal IG/EALs , the highest reading in the past twenty four hours exeluding the eurrent peak value. OPERA TING BASIS EARTHQUAKE (OBE): An OBE is vibratory ground motion for which those features of a nuclear power plant necessary for continued operat i on without undue risk to the health and safety of the public will remain functional.
SECURITY CONDITION:
OWNER CONTROLLED AREA: (Insert a site speeifie definition for this term.) Developer Note This term is typically taken to mean the site property owned by , or otherw i se und er the co ntrol of , the licensee.
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:
Tn some cases , it may be appropriate for a lieensee to define a smaller area with a perimeter eloser to the plant Proteeted Area perimeter (e.g., a site with a large OGA where some portions of the boundary may be a signifieant distanee from the Proteeted Area). In these eases , developers should eonsider using the boundary defined by the Restricted or Seeured Ovmer Controlled Area (ROGfJSOGA). The area and boundar)'
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.
seleeted for seheme use must be eonsistent 1 ,vith the deseription of the same area and boundary eontained in the Seeurity Plan. PROJECTILE:
UNISOLABLE:
An object directed toward a NP-llnuclear power plant that could cause concern for its continued operability , rel i abi lit y, or personnel safety. B-3 NE! 99 QI (RevisieA e) ~Jevemaer 2Q 12 PROTECTED AREA:_ (Insert a site specific definition for this term.) De~relaper Nate This term is typically taken to mean ti he area under continuous access monitoring and control , and armed protection as described in the site Security Plan. REFUELING PATHWAY:_ (Insert a site specific definition for this term.) Develaper Nate This description should include all the cavities, tubes, canals and pools through which irradiated fuel may be moved , but not including the reactor vessel. Includes all the cavities, tubes, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel. RtJPTURE(D):
An open or breached system line that cannot be isolated, remotely or locally. UNPLANNED:
The condition of a steam generator in 1 , 1 ,rhich primar)' to secondary leakage is of sufficient magnitude to require a safety injection. Develaper Nate This term is applicable to PWRs only. SAFETY SYSTEM: A system required for safe plant operation , cooling down the plant and/or placing it in the cold shutdown condition , including the ECCS. These are typically systems are classified as safety-related.
Develaper Nate 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 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.
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 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.
_The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.
Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not VISIBLE DAMAGE. B-3 ATTACHMENT 3 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELATING TO LICENSE AMENDMENT REQUEST TSCR-166 UPDATED DEVIATIONS AND DIFFERENCES MATRIX 100 pages follow UPDATED DAEC DEVIATIONS AND DIFFERENCES MATRIX TABLE OF CONTENTS
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-4 B-5 NEI 99 g I (Re'lisieA e) NeYember 2012 NEI 99 0 I (Re,*isioA
 
: 6) ~tOY6fl'l06F 20 J 2 APPENDIX C PERMANENTLY DEFUElED STATION ICs/EAls Recognition Category PD provides a stand alone set of ICs/EALs for a Permanently Defueled nuclear power plant to consider for use in developing a site specific emergency classification scheme. For development , it was assumed that the plant had operated under a 10 CFR &sect; 50 license and that the operating company has permanently ceased plant operations.
==GENERAL COMMENT==
Further , the company intends to store the spent fuel within the plant for some period of time. When in a permanently defueled condition , the plant licensee typically receives approval from the NRG for eJ(emption from specific emergency planning requirements.
S ................................................................................................................................
These eJ(emptions reflect the lowered radiological source term and risks associated 1.vith spent fuel pool storage relative to reactor at power operation.
1 ABNORMAL RAD LEVELS/ RADIOACTIVE EFFLUENT ICS/EALS ...................................................................
Source terms and accident analyses associated
5 COLD SHUTDOWN/
\.Vith plausible accidents are documented in the station's Final Safety Anal)1 sis Report (FSAR), as updated. As a result, each licensee will need to develop a site specific emergency classification scheme using the :NRG approved exemptions , revised source terms , and revised accident analyses as documented in the station's FSAR. Recognition Category PD uses the same ECLs as operating reactors; however , the source term and accident analyses typically limit the ECLs to an Unusual Event and Alert. The Unusual Event ICs provide for an increased awareness of abnormal conditions 1.vhile the Alert ICs are specific to actual or potential impacts to spent fuel. The source terms and release motive forces associated with a permanently defueled plant would not be sufficient to require declaration of a Site Area Emergency or General Emergency.
REFUELING SYSTEM MALFUNCTION ICS/EALS ........................................................
A permanently defueled station is essentially a spent fuel storage facility 1.vith the spent fuel is stored in a pool of water that serves as both a cooling medium (i.e., removal of decay heat) and shield from direct radiation. These primary functions of the spent fuel storage pee I are th e focus of the Recognition Category PD ICs and EALs. Radiological effluent IC and EALs were included to provide a basis for classifying eyents that cannot be readily classified based on an obserYable e\.1 ents or plant conditions alone. Appropriate ICs and EALs from Recognition Categories A, C , F , H , and S 1 ,\.1 ere modified and included in Recognition Category PD to address a spectrum of the events that may affect a spent fuel pool. The Recognition CategOF)'
20 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS ....................................................
PD ICs and EALs reflect the relevant guidance in Section 3 of this document (e.g., the importance of avoiding both over classification and under classification).
36 FISSION PRODUCT BARRIER ICS/EALS .......................................................................................................
Nonetheless , each licensee will need to develop their emergency classification scheme using the NRG approved eJ(emptions , and the source terms and accident analyses specific to the licensee.
38 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS ..............................................
Security related events will also need to be considered.
47 SYSTEM MALFUNCTION ICS/EALS .............................................................................................................
C-1 l>lel 99 0 I (Re1t*isioR
63 APPENDIX A-ACRONYMS AND ABBREVIATIONS
: 6) l>Jo,*emeer 2012 Table PD 1: Reeognition Category "PD" Initiating Condition MatFix UNUSUAL EVENT PD ,A .. Ul Release of gaseous or liquid radioactivit)*
....................................................................................
greater than 2 tirnes the (site specific effluent release controlling docurnent) limits for 60 rninutes or longer. Op. },lodes: llfetApplicehlc PD AU2 UNPLA1'J1'ffiD rise in plant radiation levels. Op. A fede s: Ne tA ppli c ehle PD SUl UNPLA1'J1'ffiD spent fuel pool temperature rise. Op. },lodes: l*let Applicehlc PD HUl Confirrned SECURITY C01'JDITI0N or threat. Op. },lodes: ,VetAppliceblc PD HU2 Hazardous event affecting SAFETY SYSTEM equipment necessary for spent fuel cooling. Op. Modes: ,"hlet Appliceblc PD HUJ Other conditions e>dst which in the judgment of the Emergency Director *.varrant declaration of a (NO)UE. Op. },fedes: ,"hlet Applicehlc ALERT PD AAl Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrern TEDE or 50 rnrern thyroid CDE. Op. },fedes: ,"hletAppliceble PD AA2 UNPLA1'J1'mD rise in plant rad i at i on le;*els that irnpedes plant access required to maintain spent fuel integrity.
84 APPENDIX B -DEFINITIONS
Op. },fedes: , Vet Applirnble PDHAl HOSTIL E ACTION v,ithin the OW1'J:ER CONTROLLED AREA or airborne attack threat within 30 minutes. Op. },fedes: A'etApplicehle PD HAJ Other conditions e>dst which in the judgrnent of the Ernergency Director warrant declaration of an Alert. Op. },{edes: ,VetAppliceble
.......................................................................................................................
,-------------------, : Table intenaea for use by 1 1 BAL aevelopers.  
89 APPENDIX C -PERMANENTLY DEFUELED ICS/EALS ..................................................................................
: Inclu s ion in licensee C-2 I ..I , , ..I , uocurnents 1s not requ1reu. 1 L------------------J 
98 UPDATED DAEC DEVIATIONS AND DIFFERENCES MATRIX
---~------------ECL: Notification of Unusual Event ~/EI 99 Ql (Revision
 
: 6) ~lovomeor 2Ql 2 PD AU1 Initiating Cenditien:
==GENERAL COMMENT==
Release of gaseous or liquid radioactivit)
S Page 1 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 DAEC Change Justification Validatidn
' greater than 2 times the (site specific effluent release controlling document) limits for 60 minutes or longer. Netes: Operating Mede Applieability:
'. # .; ,* GLOBAL#l References to NEI 99-01 Replaced with DAEC Difference Convert generic guidance to DAEC specific.
Not Applicable Example Emergeney Aetien Levels: (l or 2)
None GL0BAL#2 Effective date Replaced with TBD, 2018 Difference Convert generic guidance to DAEC specific.
* The Emergency Director should declare the Unusual Event promptl)' upon determining that 60 minutes has been exceeded , or will likely be exceeded.
None GL0BAL#3 Defined terms in Appendix B; Defined terms in Appendix B; Difference All defined terms in Appendix B used in the Title Case Upper Case document are in upper case (CAPs} to None indicate that the terms are defined. GL0BAL#4 PWR specific references PWR references removed Difference DAEC is a BWR None GLOBAL#S Recognition Category A-Recognition Category R-Difference DAEC implemented the optional Abnormal Radiation Abnormal Radiation designation of "R" for radiological related Levels/Radiological Effluent Levels/Radiological Effluent items to maintain continuity with previous None category and Emergency Action category and Emergency Action practice at DAEC. Levels; AU, AA, AS, and AG Levels; RU, RA, RS, and RG GL0BAL#6 Permanently Defueled Section Deleted references to Difference Not Applicable to DAEC None Permanently Defueled Station GL0BAL#7 Acknowledgments, Notice and Deleted Difference Not Applicable to DAEC None Executive Summary GL0BAL#8 Parameters or indications listed Some parameters or indications Difference Tables or bullets were created to present in EALs listed in EALs were placed in DAEC-specific information in a manner None tables or bulletized lists. familiar to and desired by scheme users. GL0BAL#9 Site specific information or "Site specific information or Difference Compliance with intent of the guidance.
* If an ongoing release is detected and the release start time is unlmown , assume that the release duration has e>weeded 60 minutes.
indication statements indications" were replaced with None DAEC-specific information or indications where applicable.
* If the effluent flow past an effluent monitor is knovm to have stopped due to actions to isolate the release path , then the effluent monitor reading is no longer valid for classification purposes. (l) (2) Reading on ,A_._..""JY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactiYity 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 (site specific effluent release controlling document) limits for 60 minutes or longer. Basis: This IC addresses a potential decrease in the level of safety of the plant as indicated by a low level radiological release that e>weeds regulatory commitments for an extended period of time (e.g., an uncontrolled release).
GLOBAL#lO Operating Mode Applicability lists Operating Mode Applicability lists Difference Mode numbers used for consistency with mode names (i.e., Power mode numbers (i.e., 1, 2, etc.} DAEC procedures and training.
It includes any gaseous or liquid radiological release , monitored or un monitored , including those for which a radioactivity discharge permit is normally prepared.
None Operation, Startup} GLOBAL#ll Developer's Notes Developer's Notes deleted Difference Developer's notes are not reflected in the implementation of the EALs. None GL0BAL#12 Example EAL statement "Example" deleted from Difference In adopting the EAL, the "example" status statement is no longer applicable.
Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment.
None GL0BAL#13 The following terms: "all, any, or, Consistently capitalized and Difference Capitalized and balded conditional terms in either" are sometimes capitalized balded the following terms: "ALL, !Cs and EALs for consistency based on user None and/or balded in !Cs and EALs ANY, OR, EITHER" in !Cs and EALs. feedback.
Further , there are administrative controls established to prevent unintentional releases , and to control and monitor intentional releases.
GL0BAL#14 Defined terms are only listed in Defined terms are also listed as in Difference Aid to the user to present all needed APPENDIX B -DEFINITIONS separate section of each IC/EAL information within the same section of the None where the terms are used. Basis document.
The occurrence of an e>(tended , uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.
GLOBAL#lS Term "emergency buses" Replaced with "essential buses" Difference Changed to reflect DAEC nomenclature None 2 DAEC DEVIATIONS AND DIFFERENCES MATRIX , , Section NEI 99.:01 Rev. 6 DAEC Change Justification Validation " ', ', # " \ '. COVER PAGE Development of Emergency Duane Arnold Emergency Action Difference Changes made to adapt the generic NEI None Action Levels for Non-Passive Level Technical Bases Document guidance to a DAEC-specific document Reactors Introduction Acknowledgments, Notice and Deleted Difference Not Applicable to DAEC None Executive Summary TOC 1. Regulatory Background
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. C-1 NEI 99 o I (Re*,*isieR e) Jl,levemeer 2012 ti:! te the mes that a release pa ed *ter readiAgs assu . . l'Ae"'A te haYe stepp 8 ffiHOR! !HORI ffi t lflORl!eFIS
: 1. Basis for Emergency Action Difference Title change None Levels TOC 1.1 Operating Reactors 1.1 Regulatory Background Difference Title change None TOC 1.2 Permanently Defueled Station Deleted section Difference Not Applicable to DAEC None TOC 1.3 Independent Spent Fuel 1.2 Independent Spent Fuel Difference Re-numbered None Storage Installation (ISFSI) Storage Installation (ISFSI) TOC 1.4 NRC Order EA-12-051 1.3 NRC Order EA-12-051 Difference Re-numbered None TOC 1.5 Applicability of Advance and Deleted section Difference Not Applicable to DAEC None Small Modular Reactor Designs TOC 3.Design of the NEI 99-01 3. Design of the DAEC Emergency Difference Title Change None Emergency Classification Scheme Classification Scheme TOC 3.3 NSSS Design Differences Deleted section Difference Changes made to adapt the generic NEI None guidance to a DAEC-specific document TOC 3.4 Organization and Changed to 3.3 DAEC 3.4 Difference Changes made to adapt the generic NEI None Presentation of Generic Organization and Presentation of guidance to a DAEC-specific document Information Generic Information TOC 4.0 Site-Specific Scheme 4.0 DAEC Scheme Development Difference Title change None Development TOC 4.4; 4.5; 4.6; 4.8 Deleted sections Difference Changes made to adapt the generic NEI None guidance to a DAEC-specific document TOC 4.7 Developer and User Feedback None TOC Appendix (-Permanently Deleted section Difference Changes made to adapt the generic NEI None Defueled Station ICs/EALs guidance to a DAEC-specific document 1.1 Regulatory Background Regulatory Background Difference Changes made to adapt the generic NE! None guidance to a DAEC-specific document and removed developer information 1.2 Permanently Defueled Station Section deleted Difference Not Applicable to DAEC None 1.3 1.3 Independent Spent Fuel 1.2 Independent Spent Fuel Difference Re-numbered section. None Storage Installation (ISFSI) Storage Installation (ISFSI) 3 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 DAEC Change Justification Validation
' j" dfeF ClassifieatieR
# " .. ., ,-,, " '* 1.4 1.4 NRC Order EA-12-051 1.3 NRC Order EA-12-051 Difference Re-numbered and removed wording to add None these readings (DAEC installation completed).
_8*s: ~~:e efflHeRt flew past aR * "":,er rea8iRg is "" leRge, va , ,, AmeAt is estabhs!:ie . at!:i tl=ieA the efflueAt me eA~ ire . . late the release p ' 8HO !0 80!l0RS le 199 8' g 4 times *c: f A purpeses.
1.5 Applicability to Advanced and Section deleted Difference Not Applicable to DAEC None Small Modular Reactor Designs 2 KEY TERMINOLOGY USED IN NEI KEY TERMINOLOGY USED IN Difference Minor changes to reflect DAEC-specific None 99-01 DAEC EAL SCHEME implementation.
I a release exoee IA elassiNea-,e-
3 DESIGN OF THE NEI 99-01 DESIGN OF THE DAEC Difference Changes made to adapt the generic NEI None EMERGENCY CLASSIFICATION EMERGENCY CLASSIFICATION guidance to a DAEC-specific document SCHEME SCHEME 3.1 Assignment of Emergency Assignment of Emergency Difference Changes made to adapt the generic NEI None Classification Levels (ECLs) Classification Levels (ECLs) guidance to a DAEC-specific document, removed references to PWRs, and removed developer information.
-8 Fe, &deg;'""'P**, Aet be prerated er averag; . Releases sheHlf AHies Sees Rel !fleet the llnL. a<liatieR ffi0Riter release liffiits ie, 39"' e* aetivity releases that eause h~:*;::lflit.  
3.2 Types of Initiating Conditions and Types of Initiating Conditions and Verbatim None Emergency Action Levels Emergency Action Levels 3.3 Text referring to NSSS design Deleted Difference Guidance is now DAEC specific None differences for various types or plants; Developer guidance 3.4 Organization and Presentation of DAEC-Specific Organization and Difference Renumbered to 3.3, made DAEC-specific, None Generic Information Presentation of Generic and deleted developer information Information 3.5 Mode of Applicability Matrix; Deleted "Permanently Defueled" Difference Renumbered to 3.4, removed PWR Vl Typical BWR Operating Modes section of matrix; replaced information, removed permanently Typical BWR Operating Modes defueled, and inserted DAEC Operating with DAEC-specific Operating Modes to comply with the document intent. Modes 4 Site Specific Scheme Development of the DAEC Difference Upda.ted to reflect DAEC specific scheme None Development Guidance Emergency Classification Scheme development process. 5 GUIDANCE ON MAKING GUIDANCE ON USING THE DAEC Difference Guidance is now DAEC specific None EMERGENCY CLASSIFICATIONS EALS 6-11 Recognition Category IC/EAL removed Difference Matrixes were intended for use by EAL None Matrixes developers.
+his eAL B' L #1 +his on . . lal,lishe8 By a ro ,0 ' tiRH0HS release pa-" c A :b addresses ra 10 d. aot 11*1ty d1so ar t!:i.,*ays n 2
Inclusion in licensee scheme is not desired. 4 DAEC DEVIATIONS AND DIFFERENCES MATRIX ABNORMAL RAD LEVELS/ RADIOACTIVE EFFLUENT ICS/EALS 5 DAEC DEVIATIONS AND DIFFERENCES MATRIX Recognition Category:
* s the l1m1t es ffem nen readiAgs te e>,oeed t1~e~ .. :ti:! planAed bato!:i releases "
AUl RUl Difference Global Comment #5 None Initiating Condition:
* iealiy Be as,eeiate n I 8 B " w1ll l}jl ,.. te gas). . that are Seteete ) (e.g., ra8waste, ..as eR!rnlle8 gaseeus er liq~,e ~'::~::,.ys (e.g., spills ef . g, L addresses""" Rlfl0R1te,e ) B* L #2 +his~ I r>'eys paFtieularly 0R * . . *er>**atersystelfls , ete .. , r
Release of Release of gaseous or liquid Difference Global Comment #9 None gaseous or liquid radioactivity radioactivity greater than 2 times greater than 2 times the {site-the ODAM limits for 60 minutes .-i ::::> specific effluent release or longer. <t controlling document}
* eAta su-. ' I I e IA rn sample aAalyses er eAv1reAm draiAs heat e>wl=iaAgerea rng ra<lieaetive liquids iRte stem, ' . . ... 18 Be via IC PD AA!. esoalatien ef the emer geAoy olassifioatieA le,,el weu N t s* n. h Ra<lielegieal Effl**Rt D""eleper e;,
limits for 60 minutes or longer. Operating Mode of Applicability:
* eeRtrnlliRg 8ee~ffieRt 19 t
Operating Mode of Applicability:
* C:eRerie Leiter 89 91 , +he "site Sfl**ifie effiueRt ,el~aseplaRts that have 11F1pie1HeRte8 le1HeRt "'gulatieRs related . Ii ati eRs (Rll+S) **, <<>* MJ +hese 8eeulfleRts
Verbatim None All All 6 DAEC DEVIATIONS AND DIFFERENCES MATRIX I : DA.EC *"''-,.}
'"'P. l ' *l'JJrnp,iate, the +eehRieal Si**: ~aleulatieR MaRual (O~C;;.9 ~FR Part 59 , AppeR81K Lt;,:shtles ier this IC. the Offs1te es E 10 CFR Part 2 aA t blisl=iiAg the memter 11 1 eeRtrnls e.g., 18 B *see Jer es* M te ef HOR he8elegy sheu -e -. 1Uff8 er 00C~. RB+8 er ODCM !flat *10,s 8eseriBe8 IR the
* Cf:tange 1
* e the offiueRt ffieRI . I . RiteFS sh0Hl8 iRelu e . te8 with ether pete*ti* Listed lfl0 . . t lied lfl0RiteFS ass*~** e e B' L values ier i8er i**iu81Rg IRS a {;M"H If IRelu * , fl e . Develepers
* Justification*
"'"Y al:::~::eriBe8 i* tRe llll+8 er~~.;;,~
: * *;,:* I V~lidatio9
Sesokelease lilflit:,v:*::::*
# -I (1) Reading on ANY effluent (1) Reading on ANY Table R-1 Difference See Global Comments #8, 9, 12, & 13. V2 radiation monitor greater effluent radiation monitor than 2 times the (site-greater than column "NOUE" Reworded EAL statement to remove specific effluent release for 60 minutes or longer: operator confusion as to whether they controlling document) needed to multiply the values of the limits for 60 minutes or [inserted Table R-1 of DAEC-following table by 2 or if the value provided longer: (site-specific specific radiation monitors already was 2X. Wording now matches monitor list and threshold and threshold values] wording of RSl and RGl allowing for easier values corresponding to 2 operator progression through the EALs. times the controlling document limits) (2) Reading on ANY effluent (2) Reading on ANY effluent Difference Global Comment #13 None radiation monitor greater radiation monitor greater than 2 times the alarm than 2 times the alarm -. ...: setpoint established by a setpoint established by a = 0 current radioactivity current radioactivity '-' discharge permit for 60 discharge permit for 60 'l"'"'I minutes or longer. minutes or longer. (3) Sample analysis for a (3) Sample analysis for a gaseous Difference Global Comment #9 None gaseous or liquid release or liquid release indicates a indicates a concentration concentration or release rate or release rate greater greater than 2 times the than 2 times the (site-ODAM limits for 60 minutes specific effluent release or longer. controlling document) limits for 60 minutes or longer. Intent and meaning of the EALs are not altered. 7 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section . ! .. NEI 99-01 Rev. 6 ,, . " '~.' .. 1 . DAEC Change I . Ju~tification I Valida~ion
tR>>*ays tRat are
#. I Recognition Category:
* the 1H0st "PP 1 , Be Belew """ efflueR! pa--.. R 18 Be 8eterlfliAe8 usmg *-I lated BAL value "'a) : R r t Alse , selfle these Fll0Rll0FS s *:i It is reeegaiaee tRat a ea **.1 ROOS 10 Be iRelue.ee IR t * .::.d ,elated the llll'.f8 e, ,?~Ci~ ;Rat ease, the lfl0~te, a~":;e:i!ieatieRs e, ethe~ ~**.:s:.::/eR elearly i8eRtify lfl0R1te, eaR ' .. med By +ee ***
AU2 RU2 Difference Global Comment #5 & 14 None Initiating Condition:
* 1 e BAL aR '"'' meRiteFS "'"Y *et B: ge,i: is i1Hp0FtaRt that the assee,~~FS . ts* t!:iere43re , bT . f tliese ffi0R
UNPLANNED UNPLANNED loss of water level Verbatim None loss of water level above above irradiated fuel. irradiated fuel. Operating Mode of Applicability:
* 1 reqH1re1HeR--, --h seer avaihH ,ty e . . I sos with sepa,a-e a*y li1Hitati0*s
Operating Mode of Applicability:
** t *
Verbatim None All All (1) a. UNPLANNED water level (1) a UNPLANNED water level Difference Global Comment #9, 12 & 13 V3 drop in the REFUELING drop in the REFUELING PATHWAY as indicated by PATHWAY as indicated by ANY of the following:
* address gaseeus aR8 hqu,e re** 8eme sites m8) . , fiAd it advaAtageeus te u ,..,,,...,...,,.,, ., ~7 {on *'*he 'Feehn,eo.~
ANY of the following:
{; ' l*o/ P,og,*""'
N (site-specific level
' I** so,Ha*(*)
* Report to control ::::) <C indications).
wh"* , __ ,. '""''""''"""'
room (visual observation)
-~*1 ."'; ',,;;:,,.1
* 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
* 10 /he P,*~-*""'',; *** ;. 1h, ,;,. .... ,.. ** , ' ; ', Ojfeile Do,e c.,"'"'&deg;''":'" flh effi,,at memlo,s '" .. *a O By l>JPO ,.1a1, l,re ........ 8 0 h o ***** , ,,.., O Thi, iAol,Oa, ";~**~FR >o.47(1,)(8)
' Justification V~lidatio1V#
,.; (9). I ofl O CFR SO .S4 (q) .,; H ~:**ilo,s.
: b. UNPLANNED increase in b. UNPLANNED rise in area Difference Global Comments #9 & 13 V4 area radiation levels as radiation levels as indicated by ANY of the indicated by ANY of the following radiation following radiation monitors.
the reE!HiremeRtS e J iR miRe the FeE!HIF~lfle~
monitors. (site-specific list of area
S h aeeitieR efether efffoeR H De**elepers sheule ceep . Rt , .. heR eeRs1eermg t e . ro * ..,, . .,..., ,e ' . 'S eifieoi,on, ** ' . h e; ss . , $ 'fiec1tief'ls il'I .'he . , . Bl Ejfh:e:?t Teehme~, ~e;t;:, edt!rnl Detc1i!s &j RET. le ti C,ntrelsfer R86He,egie_ . dtfie Releec1ltel'I 8.rr. ee
* Spent Fuel Pool Are~ radiation monitors)
* EfHipme geRey respeRse e te emer C-2 NEI 99 0 I (Re*,isioA e) Jl,lo*,emeer 2012 Radiati_on mo.niter readings should reflect va l ues that correspond to a radiolo ical release exceedmg 2_times a releas_e ~ontrol limit. The controlling document typical I)~ describes
Rl-9178
~eetho:olo~ies fo~ dete~mmmg effluent radiat i on monitor setpoints; these methodologies should .~se to etermme En~ values. In cases where a methodology is not adequate! , defined :***1';1'"" skeelB Beterm1Ae valees eeAsisteAt with effleeAt eeAtrel ,egelatieAs
* North Refuel Floor, RI-9163
(/g lQ cf~ art and IO CFR Part 50 Appendix T) and related guidance.  
* New Fuel Vault Area, -Rl-9153 ...; South Refuel Floor, RI-C
* ., . ~o: EA_L #1 Values in this EAL should be 2 times the setpoint established b , the ::::**elivity 8 15 ei,arge P*Fffi it le wam ef * ,el ease tkat is Aot iA eompl iaAee with tke ;peeifieB ts: Inde)ong the *~alue m this manner ensures consistency betv~*een the EAL and th t
* 0 9164 N NW Drywell Area Hi ::::, * <( Range Rad Monitor, RIM-9184A
* established by a speeific discharge permit. e se13omt Developers should researeh radiati~n monitor design documents or other information s~uree~ to ensure th~t I) the EAL value bemg considered is within the usable response and displ_a) rang~ of ~he 1~strument , and 2) there are no automatic features that may render the momtor readmg mvahd (e.g., an auto purge feature triggered at a particular indication level). ffl It is recognized that the condition described by this IC may result in a radio l ogical e uen~ :*~ue ~eyond the operating or display range of the installed effluent monitor In those case~~ 1' ;.a u~s sho~ld be determined with a margin sufficient to ensure that an ac.curate mom-or rea-mg 1s a"a1lable Fe I c AL *
* South Drywell Area Hi Range Rad Monitor, RIM-91848 Intent and meaning ofthe EALs are not altered. 9 DAEC DEVIATIONS AND DIFFERENCES MATRIX section J ' .'"NEl,99~01 Rev.* G * '** DAEC :-' ',, Change '' .. ' ...
* f .* ..r ex.amp e , an en momtor readmg might be set at 90% to 95% o~~he h1g7est accurate m~mtor re~dmg. This provision notwithstanding , if the est1matedrcalculated monitor read mg is greater than approx i mately 110% of the h" h t *t d. h ' 1g es accurate ~oni:~r rea mg , t en developers may choose not to include the monitor as an indication and 1 ent1 'an alternate EAL threshold. , _Indications from a real _time do~~ projection system are not included in the eneric EALs t~aA) heeR;ees Bo AO! ka*,e th!S eapah,hty.
* Justifica~ion , Validatioh
eor those that Bo, tke eapal,ility may A:t ~e ***itkiA * --e scope o-the p l t T h
# ., I' .. ' *::. .. Recognition Category:
* I 8 *fi
AAl RAl Difference Global Comment #5 & 14 None Initiating condition:
* 1-~ 1 f d~n :ec mcapec1 cations. A licensee may request to inelude an EAL using rea ,me ose prOJectJon s ystem results; approval ',&#xa5;ill be considered on a case by case ;asis. C f; 4 /ndications from a perimeter monitoring system are not included in the generic EA Ls
Release of Release of gaseous or liquid Verbatim None gaseous or liquid radioactivity radioactivity resulting in offsite resulting in offsite dose greater dose greater than 10 mrem TEDE than 10 mrem TEDE or 50 mrem or 50 mrem thyroid CDE. thyroid CDE. Operating Mode of Applicability:
* any 1censees d~ no~ have this capability.
Operating Mode of Applicability:
for those that do , these monitors may not
Verbatim None All All (1) Reading on ANY of the (1) Reading on ANY Table R-1 Difference Global Comment #8, 9, 12 & 13 vs following radiation radiation monitor greater monitors greater than the than column "Alert" for 15 reading shown for 15 minutes or longer: minutes or longer: (site-specific monitor list [inserted Table R-1 of DAEC-and threshold values) specific radiation monitors and threshold values] .-1 (2) Dose assessment using (2) Dose assessment using actual Difference Global Comment #9 None <C <C actual meteorology meteorology indicates doses Added bracketed
* T:::o~le~
'Preferred' to reinforce indicates doses greater greaterthan 10 mrem TEDE the 4th Note of the IC than 10 mrem TEDE or 50 or 50 mrem thyroid CDE at or mrem thyroid CDE at or beyond the SITE BOUNDARY.
:nd ;amt~med to the ~~me I eve~ as plant equipment , or within the scope of the plant mca 6 ~eCI cations. In add1t10n , readmgs may be influenced by environmental or other actor~. 1 1 1:.~1censee m_ay request to include an EAL using a perimeter monitoring system; a ppro1-a n ii I be considered on a case by case basis. EGL Assignment Attributes:
beyond (site-specific dose [Preferred]
3.1.l.B C-3 I-8 "cVV Od: JO l \IV 0d :)l l'l!A eq pf ROA\ f9A9f UO!ll'lO!J!SSBfO
receptor point). (3) Analysis of a liquid (3) Analysis of a liquid effluent Difference Global Comment #9 effluent sample indicates sample indicates a a concentration or release concentration or release rate rate that would result in that would result in doses doses greater than 10 greater than 10 mrem TEDE mrem TEDE or 50 mrem or 50 mrem thyroid CDE at or thyroid CDE at or beyond beyond the SITE BOUNDARY (site-specific dose for one hour of exposure.
,(ouei5J9UJ9 94lj0 UO!ll'JfBOSff "Sf 13!J9ll3W 9lSl3A\ 9A!l0l'lO!Pl'lJ JO weweAOUJ pua S90JROS 0!4de~O!PBJ JO esn se 4ons se9!A!lOl'l peuue1d WOJJ l[RSeJ ll'l4l seseeJOU!
receptor point) for one hour of exposure.
teAet UO!ltJ!PBJ sepn1oxe c# 1\/!I *doJp 1eAet J9ll3M G!IN:N:V1d:Nfl ua Ol enp S! i5U!pl'leJ pelBAete e4l eJe4"A sestJo LI! ,(1uo e1qeo!Jdde S! t# 1\/'3 ll34l 9lON *peJep!suoo eq p1no4s suO!lRfOAe peuutJjdJO Sl99JJ9 e4+/- "SUO!ll'JOOJ eS04l U! SJOl!UOW ,(q pel09l9P eq UB9 lB4l Sl'JeJe lUeoefpl'lJO Sf9A9f UO!ll'l!Pl'lJ e41 U! 9Sl'J9J0U!
10 DAEC DEVIATIONS AND DIFFERENCES MATRIX I ' ' DAEC Change
ue esneo OSfl'l Al'JUJ f9A9f J9ll3A\ e41 LI! doJp lUl39!J!Ui5!S y *Ee1qef!BABJ!)
* I' .. ' Justification (4) Field survey results (4) Field survey results indicate Difference Global Comment #9 None indicate EITHER of the EITHER of the following at or following at or beyond beyond the SITE BOUNDARY: (site-specific dose
suO!ll'JAJesqo tJJeweo oep!A JO f9UU0SJed lUl3fd UJOJj s&#xb5;odeJ epRfOU! , (l'JUJ SUO!ll'lO!PU!
* Closed window dose receptor point): rates greater than 10 -* Closed window dose mR/hr expected to .... rates greater than 10 continue for 60 minutes s: 0 mR/hr expected to or longer. .-4 continue for 60 Analyses of field survey
f0A9f JO S09JROS J94l0 "UO!ll'JlU0WR.llSU!
* minutes or longer. samples indicate thyroid
f9A9f e1qef !l'lAl'l UJOJJ suO!lBO!PU!
* Analyses offield survey COE greater than SO samples indicate mrem for one hour of thyroid CDE greater inhalation.
,(q peu!UJJelep
than SO mrem for one hour of inhalation.
, (f!JBW!Jd eq lf!A*, estJeJoep f9A9f J9ll'J,'A y "lUl3Jd 04lj0 Al9jl3S JO J9A0J 04l U! UO!lBPBJi50p J13!lU9lOd 13 S! UO!l!PUOO J94l!3 " SJ13!J9ll3UJ 9A!l0130!Pl'lJ JO lUl3Jd e41 U!4l!N. SJ9A9f UO!ll3!Pl3J fOJlUOO Ol ,(l!f!ql'l e41 U! sso1 JOU!lli 13JO 9A!ll'l0!PU!
Intent and meaning of the EALs are not altered. 11 
eJe s1eAet UO!lB!P13J peseeJOU!
----------
e4+/- *slueAe G'3N:N:V1d:Nfl Je410 JO 1enJ E1ueds) petB!Ptl"!
-DAEC DEVIATIONS AND DIFFERENCES MATRIX DAEC I" Change , I *' Justification I V~lidation
eAoqtJ 1eAe1 JeltJNr LI! estJeJoep tJ ,(q pesntJo SfeAef UO!ll'l!Pl'lJ lUBfd peltJAete sesseJppl'l
# I Recognition Category:
:)I S!4+/- "81!IA311Vt"q1f0N JQAO J4ffllli &sect;cJO es!J G3NNV1d:Nfl ue sell'lO!PU!
AA2 RA2 Difference Global Comment #5 & 14 None Initiating Condition:
lfRSeJ , (e,uns JO i5u!peeJ JOl!UOUJ UO!lll!Pl'lJ tJeJv "SJOl!UOW UO!ll'l!PtlJ i3U!MOjf Oj e4uo _;\~TV' , (q pell'JO!PU!
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:
Sl'J SJ9A9f UO!ll'l!Pl'lJ l'J9Jl'J U! 9S!J G3NNV1dN:fl
Operating Mode of Applicability:
:i5U!A\OIIOJ e4uo :A.NV ,(q pell'lO!PU!
Verbatim None All All (1) Uncovery of irradiated fuel (1) Uncovery of irradiated fuel in Verbatim None in the REFUELING the REFUELING PATHWAY. PATHWAY. (2) Damage to irradiated fuel (2) Damage to irradiated fuel Difference Global Comment #8, 9, 12 & 13 V6 resulting in a release of resulting in a release of radioactivity from the fuel radioactivity from the fuel as as indicated by ANY of the indicated by Hi Rad alarm for following radiation ANY of the following ARMs: monitors:
stJ 100d 1eRJ 1ueds e4l LI! doJp 1eAet Je1e;A G3:N:N:V1d:Nfl "t) Ee) Et) lU9Aff jl3RSRUf1 JO UO!lBO!J!lON_
* Spent Fuel Pool Area, RI-9178 (site-specific listing of radiation
:'}.)';I znv ad z l oz JaqwaAON:
* North Refuel Floor, Rl-9163 N monitors, and the associated
(~ IW!S!Aa'd)
* New Fuel Vault Area, RI-readings, setpoints and/or 9153 alarms)
LO 66 13:!"1:
* 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
Develof)eF Notes: NEI 99 Ql (Re\'isioR e) "!>Jove me er 2Q 12 For EAL #1 Site specific indications may include instrumentation values such as water level and area radiation monitor readings , and personnel reports. If available , video cameras may allow for remote observation.
* NW Drywell Area Hi Range monitors during plant operation, and Rad Monitor, RIM-9184A mechanical damage to a fuel assembly in
Depending on available in strumentation , the declaration may also be based on indications of water makeup rate and/or decreases in the lev el of a water storage taflb For EAL #2 The specified value of 25 mR/hr may be set to another value for a specific application with appropriate justification.
* South Drywell Area Hi the vessel can only happen with the reactor Range Rad Monitor, RIM-head removed (Mode 5). 91848 (3) Lowering of spent fuel pool (3) Lowering of spent Difference Global Comment #9 V7 level to (site-specific Level fuel pool level to 2 value). [See Developer 25.17 feet Intent and meaning of the EALs are not Notes altered. 12 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section "NEI 99-01 Rev. 6 DAEC ~hange Justification i Validation
EGL Assignment Attributes:
# I Recognition Category:
3.1.1.B B-2 
AA3 RA3 Difference Global Comment #5 & 14 None Initiating Condition:
~/El 99 QI (ReYisioR
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.
: 6) November 2012 ECL: Notification of Unusua:I;, E:*,e=n~t
plant operations, cooldown or shutdown.
--------------~PRO SU 1 Initiating Cenditien:
Operating Mode Applicability:
UNPLA1'H-ffi9 spent fuel pool temperature rise OpeFating Mede A I' . .
All Operating Mode Applicability:
* af)f) iealllh*"*
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:
Not A 1* hi ~J
* Control Room
* 1tpp-1cat1-e Example EmeFgeney Aetien Levels: (1) UNPLANNE9 spent fuel I pee temperature rise to greater than (s1*te s *.c: o peClt1C Fr-This IC addresses a condition that. potential degradation in the level o~ s:f:te~ursor to a more serious event and represents
* Control Room ARM (RM-* Central Alarm Station 9162} * (other site-specific
: a. occur , and result in a loss of pool level a:d ~fthe plant. T~u.ncorrected , soiling in the pool will . mcreased rad1at1on levels. Escalation of the emer e . g nc~* classification level *would ee via IC pg A A 1 >> .. eloper Notes: ' ** PD AA2. Th
* Central Alarm Station (by areas/rooms) survey) rtl (2) An UNPLANNED event Not used at DAEC Difference EALs RA3 and HAS are not applicable to V8 results in radiation levels DAEC because an evaluation has shown that prohibit or impede that there are no rooms or areas that access to any ofthe contain equipment which require a following plant rooms or manual/local action as specified in areas: operating procedures used for normal plant operation, cooldown and shutdown.
* e site specific temperature should e EGL A
All (site-specific list of plant rooms areas outside the Control Room that or areas with entry-related mode contain equipment necessary for normal applicability identified) plant operation, cooldown and shutdown do not require physical access to operate. Intent and meaning of the EALs are not altered. 13 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 DAEC j . Change j .. Justification I Validation
* n.ss1gnment Attrieutes:
# I Recognition Category:
3 .1.1.A B-3 ECL: Notification of Unusual Event Initiating Cenditien:
ASl RSl Difference Global Comment #5 & 14 None Initiating Condition:
Confirmed SECURITY CONDITION or threat. Operating Mede ,\pplieability:
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:
Not Applicable Example Emergeney Aetien Le~1 els: (I or 2 or 3) NEI 99 QI (ReYisieA
All Operating Mode Applicability:
: 6) l>le&#xa5;ember 2Ql 2 PD HU1 (1) A SECURITY CONDITiffi,1 that does not involi,re a HOSTILE ACTION as reported by the (site specific security shift superi,rision).  
All Verbatim None (1) Reading on ANY of the (1} Reading on ANY Table R-1 Difference Global Comment #8, 9, 12 & 13 V9 following radiation effluent radiation monitor monitors greater than the greater than column "SAE" reading shown for 15 for 15 minutes or longer. minutes or longer: .-I (site-specific monitor list and [inserted Table R-1 of DAEC-V) <( threshold values) specific radiation monitors and threshold values] (2) Dose assessment using (2) Dose assessment using actual Difference Global Comment #3 & 9 None actual meteorology meteorology indicates doses Added bracketed
(2) (3) }ofotification of a credible security threat directed at the site. A validated notification from the NRG providing information of an aircraft threat. Thi_s I~ addre~ses events that pose a threat to plant personnel or the equipment necessary to mamtam cool_mg of spent ~el , and thus represent a potential degradation in the level of plant safet_y. Security events v,rh1ch do not meet one of these EALs are adequately addressed by the requirements of 10 CFR &sect; 73.71 or 10 CFR &sect; 50.72. Security events assessed as HOSTILE ACTiffi,lS are classifiable under IC PD HAl. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security related event. Classification of these events will initiate appropriate threat related notifications to plant personnel and OROs. Security plans and terminology are based on the guidance provided by NEI 03 12 , Temptatcfor the Security Ptan , Treining e,9d QHelificetion Plen , Se.fcgMerds Contingency Plen {end Independent Spent-Fuel St-orege Instell8tio1~
'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.
Security Progrmn}.
dose receptor point). [Preferred]
EAL #1 references (site specific security shift supervision) because these are the individuals trained to confirm that a security event is occurring or has occurred.
14 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 DAEC Change Justification Valjdation
Training on security event confirmation and classification is controlled due to the nature of Safeguards and 1 O CFR &sect; 2.39 information. EAL #2 addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with (site specific procedure).
# (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):
EAL#~ addresses the threat from the impact of an aircraft on the plant. The NRG Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft.
* Closed window dose
The status and size of the plane may also be provided by NORAD through the NRG. Validation of the threat is performed in accordance with (site specific procedure). Emergency plans and implementing procedures are public documents; therefore , EALs should not incorporate gecurity sensitive information.
* 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.
This includes information that may be
* Analyses of field survey
* B-4 NEI 99 QI (ReYisieA
* 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.
: 6) ~Jeyemeer 2Q 12 advantageous to a potential adversary , such as the particulars concerning a specific threat or threat location.
hour of inhalation.
Security sensitive information should be contained in non public documents such as the Security Plan. Escalation of the emergency classification le*,el would be via JC PD HAI. DevelepeF Netes: The (site specific security shift supervision) is the title of the on shift individual responsible for supervision of the on shift security force. The (site specific procedure) is the procedure(s) used by Control Room and/or Security personnel to determine if a security threat is credible , and to validate receipt of aircraft threat information.
0 .... V) <( Intent and meaning of the EALs are not altered. 15 L ____ -
Emergency plans and implementing procedures are public documents; therefore , EALs should not incorporate Security sensitive information.
DAEC DEVIATIONS AND DIFFERENCES MATRIX Secfion "NH99.:.01 Rev~ 6 DAEC: . Change Justification Validation
This includes information that may be advantageous to a potential adversary , such as the particulars concerning a specific threat or threat location.
#. Recognition Category:
Security sensitive information should be contained in non public documents such as the Security Plan. With due consideration given to the above developer note , EALs may contain alpha or numbered references to selected events described in the Security Plan and associated implementing procedures.
AS2 RS2 Difference Global Comment #5 None Initiating Condition:
Such references should not contain a recognizable description of the event. For e>rnmple , an EAL may be 't't'Orded as " Security event #2 , #5 or #9 is reported by the (site specific security shift superYision)." EGL Assignment Attributes:
Spent fuel Spent fuel pool level at 16.36 feet Difference Global Comment #9 VlO pool level at (site-specific Level 3 description).
: 3. I. I .A B-5 Nel 99 o I (ReYisioH G) No,*ember 2012 PD HU2 ECL: Notification of Unusual Event Initiating Cenditien:
Operating Mode Applicability:
Hazardous event affecting SAFHTY SYSTEM equipment necessary for spent fuel cooling. Operating Mede Applieability:
All Operating Mode Applicability:
Not Appl i cable Example Emergeney A,etien Le,;els: (]) a. b. C. The occurrence of i+ ...... ~Y of the following hazardous events:
All Verbatim None N (1) Lowering of spent fuel pool (1) Lowering of spent fuel pool Difference Global Comment #9 & 12 VlO Ill <C level to (site-specific Level level to 16.36 feet 3 value). Intent and meaning ofthe EALs are not altered. 16 DAEC DEVIATIONS AND DIFFERENCES MATRIX 1:* .* Section I
* NEI 99-01.Re\r; . 6
* I DAEC .. , Change I ... , Justification*
1 vaUdatip'n
# 1 Recognition Category:
AGl RGl Difference Global Comment #5 & 14 None Initiating Condition:
Release of Release of gaseous radioactivity Verbatim None gaseous radioactivity resulting in resulting in offsite dose greater offsite dose greater than 1,000 than 1,000 mrem TEDE or 5,000 mrem TEDE or 5,000 mrem mrem thyroid CDE. thyroid CDE. Operating Mode Applicability:
All Operating Mode Applicability:
All Verbatim None (1) Reading on ANY of the (1) Reading on ANY Table R-1 Difference Global Comment #8, 9, 12 & 13 V9 following radiation effluent radiation monitor monitors greater than the greater than column "GE" for .... reading shown for 15 15 minutes or longer . minutes or longer: <C (site-specific monitor list and [inserted Table R-1 of DAEC-threshold values) specific radiation monitors and threshold values] (2) Dose assessment using (2) Dose assessment using actual Difference Global Comment #3 & 9 None actual meteorology meteorology indicates doses Added bracketed
'Preferred' to reinforce indicates doses greater greater than 1,000 mrem the 4th Note of the IC than 1,000 mrem TEDE or TEDE or 5,000 mrem thyroid 5,000 mrem thyroid CDE at CDE at or beyond the SITE or beyond (site-specific BOUNDARY.
[Preferred]
dose receptor point). 17 --~
DAEC DEVIATIONS AND DIFFERENCES MATRIX DAEC * , . Change , . Justification I Validati(;)n
# I (3) Field survey results (3) Field survey results indicate Difference Global Comment #3 & 9 None indicate EITHER of the EITHER of the following at or following at or beyond beyond the SITE BOUNDARY: (site-specific dose receptor
* Closed window dose point): rates greater than 1,000
* Closed window dose mR/hr expected to rates greater than 1,000 continue for 60 minutes mR/hr expected to or longer. continue for 60 minutes
* Analyses of field survey or longer. samples indicate thyroid -* Analyses of field survey CDE greater than 5,000 ... samples indicate thyroid mrem for one hour of C 0 CDE greater than 5,000 inhalation.
.-l mrem for one hour of C, <C inhalation.
Intent and meaning of the EALs are not altered. 18 DAEC DEVIATIONS AND DIFFERENCES MATRIX I*: Section *f :.:. NEI 99~01 Rev: 6 Ju~tification J Validatio.n
'# I .. . ' Recognition Category:
AG2 RG2 Difference Global Comment #5 None Initiating Condition:
Spent fuel Spent fuel pool level cannot be Difference Global Comment #9 VlO pool level cannot be restored to restored to at least 16.36 feet for at least (site-specific Level 3 60 minutes or longer. description) for 60 minutes or longer. Operating Mode Applicability:
All Operating Mode Applicability:
All Verbatim None N (1) Spent fuel pool level cannot (1) Spent fuel pool level cannot Difference Global Comment #9 & 12 VlO <( be restored to at least (site-be restored to at least 16.36 specific Level 3 value) for 60 feet for 60 minutes or longer. minutes or longer. Intent and meaning of the EALs are not altered. 19 DAEC DEVIATIONS AND DIFFERENCES MATRIX COLD SHUTDOWN/
REFUELING SYSTEM MALFUNCTION ICS/EALS 20 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section .NEI 99-01 Rev: 6 DAEC Change Justification j Validation
# [ Recognition Category:
CUl CUl Verbatim Global Comment #11, 14 None Initiating Condition:
UNPLANNED UNPLANNED loss of RPV Difference Global Comment #4 None loss of (reactor vessel/RCS
[PWR] inventory for 15 minutes or or RPV [BWR]) inventory for 15 longer minutes or longer. Operating Mode Applicability:
Operating Mode Applicability:
4, Difference Global Comment #10 None Cold Shutdown, Refueling 5 (1) UNPLANNED loss of (1) UNPLANNED loss of reactor Difference Global Comment #4 & 12 None reactor coolant results in coolant results in RPV level (reactor vessel/RCS
[PWR] less than a required lower or RPV [BWR]) level less limit for 15 minutes or than a required lower limit longer. .-1 for 15 minutes or longer. ::> u (2) a. (Reactor vessel/RCS
[PWR] (2) a. RPV level cannot be Difference Global Comment #4 None or RPV [BWR]) level cannot monitored.
be monitored.
AND AND b. UNPLANNED increase in b. UNPLANNED level rise in Difference Global Comment #9 None (site-specific sump and/or Drywell/Reactor Building tank) levels. Equipment or Floor Drain sump, or Suppression Pool. Intent and meaning of the EALs are not altered. Recognition Category:
CU2 CU2 Verbatim Global Comment #11, 14 None Initiating Condition:
Loss of all Loss of all but one AC power Difference Global comment #15 None but one AC power source to source to essential buses for 15 emergency buses for 15 minutes minutes or longer. or longer. N Operating Mode Applicability:
Operating Mode Applicability:
4, Difference Global Comment #10 None ::> u Cold Shutdown, Refueling, 5, Defueled Defueled (1) a. AC power capability to (1) a. AC power capability to Difference Global Comment #9, 12, & 13 Vll (site-specific emergency 1A3 and 1A4 buses is buses) is reduced to a reduced to a single power 21 single power source for 15 source for 15 minutes or minutes or longer. longer. AND AND b. Any additional single b. Any additional single power source failure will power source failure will result in loss of all AC result in loss of ALL AC power to SAFETY SYSTEMS. power to SAFETY SYSTEMS. Intent and meaning of the EALs are not altered. 22 DAEC DEVIATIONS AND DIFFERENCES MATRIX . Section NEI 99-01 Rev. 6 DAEC -I Change . j iustification I Validation
# I Recognition Category:-CU3 CU3 Verbatim Global Comment #11, 14 None Initiating Condition:
UNPLANNED UNPLANNED increase in RCS Verbatim None increase in RCS temperature.
temperature.
Operating Mode Applicability:
Operating Mode Applicability:
4, Difference Global Comment #10 None Cold Shutdown, Refueling 5 (1) UNPLANNED increase in (1) UNPLANNED increase in RCS Difference Global Comment #9 & 12 Vl RCS temperature to temperature to greater than greater than (site-specific 212&deg;F Technical Specification cold shutdown temperature limit). (2) Loss of ALL RCS (2) Loss of ALL RCS temperature Difference Global Comment #4 & 13 None temperature and (reactor and RPV level indication for m vessel/RCS
[PWR] or RPV 15 minutes or longer ::::, [BWR]) level indication for u 15 minutes or longer. Intent and meaning of the EALs are not altered. 23 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section *. ** NEI 99-01 Rev: 6 DAEC I Change Justification I Validation
# I Recognition Category:
CU4 CU4 Verbatim Global Comment #11, 14 None Initiating Condition:
Loss of Vital Loss of Vital DC power for 15 Verbatim None DC power for 15 minutes or minutes or longer. longer. Operating Mode Applicability:
Operating Mode Applicability:
4, Difference Global Comment #10 None Cold Shutdown, Refueling 5 ,::I' (1) Indicated voltage is less (1) Indicated voltage is less than Difference Global Comment #9, 12, 13 V12 ::::, than (site-specific bus 105 VDC on BOTH Div 1 and u voltage value) on required Div 2 125 VDC buses for 15 Vital DC buses for 15 minutes or longer minutes or longer. Intent and meaning of the EALs are not altered. 24 DAEC DEVIATIONS AND DIFFERENCES MATRIX Settion NEI 99-Ql Rev. 6
* DAEC , .. Change . J&#xb5;stification I Validation
# I Recognition Category:
CU5 CU5 Verbatim None Initiating Condition:
Loss of all Loss of all onsite or offsite Verbatim None onsite or offsite communications communications capabilities.
capabilities.
Operating Mode Applicability:
Operating Mode Applicability:
4, Difference Global Comment #10 None Cold Shutdown, Refueling, 5, Defueled Defueled (1) Loss of ALL of the following (1) Loss of ALL of the following Difference Global Comment #9, 12 & 13 V13 onsite communication onsite communication methods: methods: (site-specific list of
* Plant Operations Radio communications methods) System
* In-Plant Phone System
* Plant Paging System in :::::, (Gaitronics) u (2) Loss of ALL of the following (2) Loss of ALL of the following Difference Global Comment #9 & 13 V13 ORO communications offsite response organization V14 methods: communications methods: (site-specific list of
* DAEC All-Call phone communications methods)
* All telephone lines (PBX and commercial)
* Cell Phones (including fixed cell phone system)
* Control Room fixed satellite phone system
* FTS Phone system 25 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 DAEC
* I change Ju~tification I Validation
# I (3) Loss of ALL of the following (3) Loss of ALL of the following Difference Global Comment #9, 12 & 13 V13 NRC communications NRC communications methods: methods: (site-specific list of
* FTS Phone system -communications methods)
* All telephone lines (PBX +: C and commercial) 0 Cell Phones (including in * :::> fixed cell phone system) u
* Control Room fixed satellite phone system Intent and meaning of the EALs are not altered. 26 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section < .NEI 99-01. Rev. *6 DAEC Change Justification Validation
# Recognition Category:
CAl CAl Verbatim Global Comment #11, 14 None Initiating Condition:
Loss of Loss of RPV inventory.
Difference Global Comment #4 None (reactor vessel/RCS
[PWR] or RPV [BWR]) inventory.
Operating Mode Applicability:
Operating Mode Applicability:
4, Difference Global Comment #10 None Cold Shutdown, Refueling 5 (1) Loss of (reactor vessel/RCS (1) Loss of RPV inventory as Difference Global Comment #4, 9 & 12 V15 [PWR] or RPV [BWR]) indicated by level less than inventory as indicated by 119.5 inches level less than (site-specific level). (2) a. (Reactor vessel/RCS
[PWR] (2) a. RPV level cannot be Difference Global Comment #4 None or RPV [BWR]) level cannot monitored for 15 minutes .... be monitored for 15 or longer <C u minutes or longer AND AND Difference Global Comment #4, 9 & 13 None b. UNPLANNED increase in b. UNPLANNED level rise in (site-specific sump and/or Drywell/Reactor Building tank) levels due to a loss of Equipment or Floor Drain (reactor vessel/RCS
[PWR] sump, or Suppression Pool or RPV [BWR]) inventory.
due to a loss of RPV inventory.
Intent and meaning ofthe EALs are not altered. 27 DAEC DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev: *6 DAEC Change Justification
! Validation
# ! Recognition Category:
CA2 CA2 Verbatim Global Comment #11, 14 None Initiating Condition:
Loss of all Loss of all offsite and all onsite Difference Global Comment #15 None offsite and all onsite AC power to AC power to essential buses for emergency buses for 15 minutes 15 minutes or longer. or longer. Operating Mode Applicability:
Operating Mode Applicability:
4, Difference Global Comment #10 None N <( Cold Shutdown, Refueling, 5, Defueled u Defueled (1) Loss of ALL offsite and ALL (1) Loss of ALL offsite and ALL Difference Global Comment #9, 12 & 13 V12 onsite AC Power to (site-onsite AC Power to 1A3 and specific emergency buses) 1A4 for 15 minutes or longer. for 15 minutes or longer. Intent and meaning of the EALs are not altered. 28 
----------------------------------------
DAEC DEVIATIONS AND DIFFERENCES MATRIX Recognition Category:
CA3 CA3 Verbatim Global Comment #11, 14 None Initiating Condition:
Inability to Inability to maintain the plant in Verbatim None maintain the plant in cold cold shutdown.
shutdown.
Operating Mode Applicability:
Operating Mode Applicability:
4, Difference Global Comment #10 None Cold Shutdown, Refueling 5 (1) UNPLANNED increase in (1) UNPLANNED increase in RCS Difference Global Comment #9 & 12 Vl RCS temperature to temperature to greater than greater than (site-specific 212&deg;F for greater than the Technical Specification duration specified in Table C-cold shutdown 2. temperature limit) for greater than the duration specified in the following table. Table: RCS Heat-up Duration Threi h~lrlr .c C-2 RCS Heat-up Duration Th Difference Global Comment #4 None RCS Status Containment sea -up C'I) Closure Status Duration . Containment Closure Changed "RCS Status" to "RCS Integrity" to <( u Intact (but not at *-** -a, ... Status match current site nomenclature reduced inventory Not applicable i:;n m;~ .. +nr* [PWR]) Intact Not Applicable Established N%iffirncJtes*
Established Not intact (or at reduced inventory
[PWR]) Not Established 0 minutes Not Established
* If an RCS heat removal system is in ope ratio
* If a&i RCS heat removal system is in operatior l wy:fam t JMi'R1cs emperature is being reduced, frame and RCS temperature is being reduce , tntpjfJB1i?t applicable.
(2) UNPLANNED RCS pressure (2) UNPLANNED RCS pressure Difference Global Comment #4 & 9 V16 increase greater than (site-increase greater than 10 psig Added "due to a loss of RCS cooling" to specific pressure reading).
due to a loss of RCS cooling. clarify the intent of the EAL (This EAL does not apply during water-solid plant conditions.  
[PWR]) Intent and meaning of the EALs are not altered. 29 Recognition Category:
CA6 Initiating Condition:
Hazardous event affecting a SAFE1Y SYSTEM needed for the current operating mode. Operating Mode Applicability:
Cold Shutdown, Refueling (1) a. The occurrence of ANY of the following hazardous events:
* Seismic event (earthquake)
* Seismic event (earthquake)
* Internal or external flooding event
* Internal or external flooding event
Line 1,313: Line 1,033:
* EXPLOSION  
* EXPLOSION  
* (site specific hazards)
* (site specific hazards)
* Other events with similar hazard characteristics as determined by the Shift Manager DAEC DEVIATIONS AND DIFFERENCES MATRIX CA6 Hazardous event affecting a SAFE1Y SYSTEM needed for the current operating mode. Operating Mode Applicability:
* Other events vrith similar hazard characteristics as determined by the Shift Manager The event has damaged at least one train of a SAFETY SYSTEM needed for spent fuel cooling. AND The damaged SAFETY SYSTEM train(s) cannot , or potentially cannot , perform its design function based on EITHER:
4, 5 (1) a. The occurrence of ANY of the Table C-3 hazardous events:
* Indications of degraded performance
* Seismic event (earthquake)
* VISIBLE; DAM.AGE This IC addresses a hazardous event that causes damage to at least one tra i n of a SAFHTY SYSTEM needed for
* Internal or external flooding event High winds or tornado strike FIRE EXPLOSION Other events with similar hazard characteristics as determined by the Shift Manager or Emergency Director 30 Verbatim Global Comment #11, 14 Verbatim Difference Global Comment #10 Difference Global Comment #9, 12 & 13 None None None None DAEC DEVIATIONS AND DIFFERENCES MATRIX AND AND b. EITHER of the following:
: b. 1. Event damage has Deviation Adopted the revised EAL wording provided V17 1. Event damage has caused indications of in approved EAL FAQ 2016-02. caused indications of degraded degraded performance performance in one in at least one train of a train of a SAFETY SAFETY SYSTEM needed SYSTEM needed for for the current the current operating operating mode. mode. AND OR 2. EITHER of the Deviation Adopted the revised EAL wording provided V17 1. The event has caused following:
in approved EAL FAQ 2016-02. VISIBLE DAMAGE to a
* Event damage has SAFETY SYSTEM caused indications Difference Added the following clarification to the V18 component or structure of degraded Basis from EALFAQ 2018-04: -needed for the current performance to a An event affecting a single-train SAFETY ...; operating mode. second train of the SYSTEM (i.e., there are indications of s:: 0 SAFETY SYSTEM degraded performance and/or VISIBLE U) needed for the DAMAGE affecting the one train) would not <( u current operating be classified under SA8 because the two-mode, or train impact criteria that underlie the EALs
* The event has and Bases would not be met. If an event resulted in VISIBLE affects a single-train SAFETY SYSTEM, then DAMAGE to the the emergency classification should be second train of a made based on plant SAFETY SYSTEM parameters/symptoms meeting the EALs needed for the for another IC. Depending upon the current operating circumstances, classification may also occur mode. based on Shift Manager/Emergency Director judgement.
Intent and meaning of the EALs are not altered. 31 DAEC DEVIATIONS AND DIFFERENCES MATRIX Recognition Category:
CSl CSl Verbatim Global Comment #11, 14 None Initiating Condition:
Loss of Loss of reactor vessel/RCS Difference Global Comment #4 None (reactor vessel/RCS
[PWR] or RPV inventory affecting core decay [BWR]) inventory affecting core heat removal capability.
decay heat removal capability.
Operating Mode Applicability:
Operating Mode Applicability:
4, Difference Global Comment #10 None Cold Shutdown, Refueling 5 (1) a. CONTAINMENT CLOSURE (1) a. CONTAINMENT CLOSURE Difference Global Comment #9 & 12 V19 not established.
not established.
.... AND AND II) b. (Reactor vessel/RCS
[PWR] b. RPV level less than +64 u or RPV [BWR]) level less inches than (site-specific level). (2) a. CONTAINMENT CLOSURE (2) a. CONTAINMENT CLOSURE Difference Global Comment #4 & 9 V19 established.
established.
AND AND b. (Reactor vessel/RCS
[PWR] b. RPV level less than +15 or RPV [BWR]) level less inches than (site-specific level). 32 I I L -.... s:: 0 .... V) u (3) a. b. (Reactor vessel/RCS
[PWR] or RPV [BWR]) level cannot be monitored for 30 minutes or longer. AND Core uncovery is indicated by ANY of the following:
* (Site-specific radiation monitor) reading greater than (site-specific value)
* Erratic source range monitor indication
[PWR]
* UNPLANNED increase in (site-specific sump and/or tank) levels of sufficient magnitude to indicate core uncovery * (Other site-specific indications)
DAEC DEVIATIONS AND DIFFERENCES MATRIX (3) a. RPV level cannot be Difference Global Comment #4 None monitored for 30 minutes or longer. AND b. Core uncovery is indicated Difference Global Comment #9 &13 V6 by ANY of the following:
* Drywell Monitor (9184A/B) reading greater than 5.0 R/hr
* UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool of sufficient magnitude to indicate core uncovery Intent and meaning of the EALs are not altered. 33 DAEC DEVIATIONS AND DIFFERENCES MATRIX Recognition Category:
CGl CGl Verbatim Global Comment #11, 14 None Initiating Condition:
Loss of Loss of reactor vessel/RCS Difference Global Comment #4 None (reactor vessel/RCS
[PWR] or RPV inventory affecting fuel clad [BWR]) inventory affecting fuel integrity with containment clad integrity with containment challenged.
challenged.
Operating Mode Applicability:
Operating Mode Applicability:
4, Difference Global Comment #10 None Cold Shutdown, Refueling 5 (1) a. (Reactor vessel/RCS
[PWR] (1) a. RPV level less than +15 Difference Global Comment #4, 9, 12 & 13 V19 or RPV [BWR]) level less inches for 30 minutes or ..-1 than (site-specific level) for longer. u 30 minutes or longer. AND AND b. ANY indication from the b. ANY indication from the Containment Challenge Containment Challenge Table (see below). Table (see below). (2) a. (Reactor vessel/RCS
[PWR] (2) a. RPV level cannot be Difference Global Comment #

Revision as of 11:50, 21 September 2018

Response to Request for Additional Information Regarding License Amendment Request (TSCR-166), Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors. (Pa
ML18212A231
Person / Time
Site: Duane Arnold NextEra Energy icon.png
Issue date: 07/26/2018
From:
NextEra Energy Duane Arnold
To:
Office of New Reactors
References
NG-18-0090
Download: ML18212A231 (125)
v • d • e


Text

NEI 99 QI (RevisioA

6) }>Joye me er 2Q 12 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS 17 9

}>ffil 99 QI (RevisioR e) November 2012 Tal>le H 1: Reeognition CategoFY "H" Initiating Condition MatFix UNUSUAL EVENT HUl Confirmed SECURITY CONDITION or Op. A1edes: All HU2 Seismic e 1 ,rent greater than OBE Op. A1edes: All HUJ Hazardous e¥eflt Op. A1edes: All HU4 FIRE potentially degrading the level of safety of the plant. Op. },1edes: All ALERT HAl HOSTILE ACTION within the OWNER CO~ffROLLED AREA or airborne attack threat *within 30 minutes. Op. },fades: All HA5 Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdo 1 Nn. Op. ,\1edes: All Hf...(i Control Room evacuation resulting in transfer of plant control to alternate locations.

Op. ,\1edes: All SITE AREA EMERGENCY HSl HOSTILE ACTION within the PROTECTED AREA. Op. A1edes: All HS(i Inability to control a key safety function from outside the Control Room. Op. },1edes: All 180 I GENERAL EMERGENCY HGl HOSTILE ACTION resulting in loss of physical control of the facility.

Op. ,\1edes: All Table intended for use by 1 EAL deYelopers.

lnelusion in lieensee : doeurnents is not required.

L------------------J UNUSUAL EVENT HU7 Other conditions exist which in the judgment of the Emergency Director 1tvarrant deelaration of a (NO)UE. Op. J,1odes: All ALERT HA7 Other conditions e>,ist which in the judgment of the Emergency Director warrant deelaration ofan Alert. Op. A/odes: All ECL: Notification of Unusual Event SITE AREA EMERGENCY HS7 Other conditions e>,ist which in the judgment of the Emergency Director warrant deelaration of a Site Area Emergency.

Op. A1ode s: All Initiating Condition:

Confirmed SECURITY CONDITION or threat. Operating Mode Applicability:

All Emergency Action Levels: NE! 99 01 (RevisioR

6) November 2012 GENERAL EMERGENCY HG7 Other conditions e>,ist which in the judgment of the Emergency Director warrant declaration of a General Emergency.

Op. ,\1odes: All HU1 E xample Emergency Action Levels: (1 or 2 or 3) A SECURJTY CONDITION that does not involve a HOSTILE ACTION as reported by the (site specific security shift supervision).DAEC Security Shift Supervision. Notification of a credible security threat directed at the siteDAEC. 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 S E CURITY 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. -terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). 181

}ffil 99 0 I (Re,*isioA 6) }fo 1 remeer 2012 SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including theincluding 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 f-73.71 or 10_-CFR--§ 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HAI, HSI and HGl. Timely and accurate communications between DAEC Security Shift Superv i sion 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 personne l and GRGoffsite response organization

s. 182 NEI 99 QI (RevisieA a) !>foveme er 2012 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}.

183 NEI 99 Q 1 (ReYisioA

  • i) l'loYeFAeeF 2Q12 EAL HUI.I references (site specific security shift supervision)DAEC Security Shift Supervision because these are the indi vidua l s trained to co nfirm that a security event i s occurring or has occ urr ed. Training on security event confirmation and classification i s controlled due to the nature of Safeguards and 10 CFR § 2.39 Q inform atio n. EAL HUI.2 addresses the receipt of a credib l e security threat. The credibi lity of the threat is assessed in accordance with Abnormal Operating Procedure (AOP) 914, Security Events. specific procedure). EAL HU 1.3 addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) wi ll comm unicat e to the li censee i f the threat inv olves an aircraft.

T h e status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat i s performed in accordance w ith (site specific procedure)

Abnormal Operating Procedure (AOP) 914, Security Events . .,. Emergency plans and implementing procedures are public documents; therefore , EALs should do not in corporate Security-sensitive in formation. This includ es information that may be advantageous to a potential adversary , such as the particulars concerning a specific threat or threat l ocat ion. Security-sensitive information should beis contained in non public documents such as the Security Plan. Esca l ation of the emerge n cy class ific ation l eve l wou ld be via IC HA 1. Developer Notes: The (site specific security shift supervision) is the title of the on shift individual responsible for supervision of the on shift security force. The (site specific procedure) is the procedure(s) used by Control Room and/or Security personnel to determine if a security threat is credible, and to validate receipt of aircraft threat information.

Emergency plans and implementing procedures are public documents; therefore, E/\Ls should not incorporate Security sensitive information. This includes information that may be advantageous to a potential adversary , such as the particulars concerning a specific threat or threat location.

Security sensitive information should be contained in non public documents such as the Security Plan. With due consideration given to the above developer note, EALs may contain alpha or numbered references to selected events described in the Security Plan and associated implementing procedures.

Such references should not contain a recognizable description of the event. For example , an EAL may be worded as "Security event #2, #5 or #9 is reported by the (site specific security shift supervision)

." EGL Assignment Attributes:

3 .1.1.A 184 NEI 99 0 I (Re 1 1isioR 6) }io¥ember 2012 ECL: Notification of Unusual Event Initiating Condition:

Seismic event greater than OBE levels. Operating Mode Applicability:

All Exa1Rple Emergency Action Levels: H 2.1 Seismic event greater than Operating Basis Earthquake (OBE) as indicated by+ HU2 --+----receipt of the Amber Operating Basis Earthquake Light and the wailing seismic alarm on IC35. 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 t. An earthquake greater than an OBE but less than a Safe ShutdownDesign Basis Ear thquake (SSeDBE)i 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 under sta nd any impacts , this event represents a potential degradation of the level of safety of the plant. Eve nt verification with external sources s hould not be necessary durin g or following an OBE. Ea rthquakes of this magnitude should be readily felt by on-site personnel and recognized as a se ismic event (e.g., typical lateral accelerations are in e>wess of 0.08g). The Shift Manager or E mergenc y Director ma y seek external verification if deemed appropriate ( e.g., a call to the + AR OBE is ¥ibratory grouRd motioR fur whieh those features ofa Ruelear power plaRt Reeessary for eoRtiRued operatioR without uRdue risk to the health aRd safety of the publie will remaiR fuAetioAal.

! AR SSE is Yibratory grouRd motioR fur whieh eertaiR (geRerally, safety related) struetures , systems , aAd eompoReRts must be desigRed to remaiR fuAetioRal.

185 NEI 99 0 I (Re\*ision

6) NoYember 2012 USGS , check internet news so urces , 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 .£.A.9SA8. De"-1 el0per Netes: This "site specific indication that a seismic event met or exceeded QBE limits" should be based on the indications, alarms and displays of site specific seismic monitoring equipment.

Indications described in the EAL should be limited to those that are immediately available to Control Room personnel and which can be readily assessed. Indications available outside the Control Room and/or 1.vhich require lengthy times to assess (e.g., processing of scratch plates or recorded data) should not be used. The goal is to specify indications that can be assessed within 15 minutes of the actual or suspected seismic event. For sites that do not have readily assessable QBE indications within the Control Room, developers should use the following alternate EAL (or similar 1.vording). (1) a. b. Control Room personnel feel an actual or potential seismic event. The occurrence of a seismic event is confirmed in manner deemed appropriate by the Shift Manager or Emergency Director.

The EAL l .b statement is included to ensure that a declaration does not result from felt vibrations caused by a non seismic source (e.g., a dropped heavy load). The Shift Manager or Emergency Director may seek e><ternal verification if deemed appropriate (e.g., a call to the USGS, check internet nev,rs sources, etc.); howe*,rer, the verification action must not preclude a timely emergency declaration.

It is recognized that this alternate EAL *.vording may cause a site to declare an Unusual Event *while another site , similarly affected but with readily assessable QBE indications in the Control Room, may not. The above alternate

  • n<ording may also be used to develop a compensatory EAL for use during periods vrhen a seismic monitoring system capable of detecting an QBE is out of service for maintenance or repair. EGL Assignment Attributes:

3.1.1.A 186 NEI 99 Q 1 (ReYision

6) No 1 1ember 2Ql2 HU3 ECL: Notification of Unu s ual Event Initiating Condition:

Hazardous event.§ Operating Mode Applicability:

All Emergency Action Levels: Example EmeFgeney Aetion Levels: (1 or 2 or 3 or 4 or 5 or 6) Note: E AL HU3.4 does not apply to routine traffic impediments such as fog , snow , ice , or vehicle breakdown s o r accident s. ~1 !k 2 H~J3.3 H LJ 3.4 A tornado strike within the PROTECTED AREA. Internal room or area floodin g 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 PROTECT E D AREA is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).

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

PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related

.A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the EGGS. Systems classified as safety re l ated. Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. EAL HU3.1 addresses a tornado striking (touching down) within the Protected Area. E AL HU3.2 addresses floodin g of a building room or area that results in operators isolating power to a SAFETY SYST E M 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 SAF E TY SYSTEM component from its power source ( e.g., a breaker or relay trip). To 187 Jloffil 99 01 (Revision a) Jl,fo,,cember 2012 warrant classification , operability of the affected component must be required by Technical Specifications for the current operating mode. EAL HU3.3 addresses a hazardous materials event orig in ating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA. 188 NEI 99 01 (RevisioR e) NoYeR'leer 2012 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 fai lur e , 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 ve hicl e breakdowns or accidents , but rather to more s i gnificant conditions such as the Hurricane Andrew strike on Turkey Point in 1992 , the flooding around the Cooper Stat i on during the Midwest floods of 1993, or the flooding around Ft. Ca lh oun Station in 2011. EAL .H..!LL5 addresses (site specific description). . Escalation of the emergency classification l eve l wou ld be based on I Cs in Recognition Categories AR , F, Sor C. De'i'elepeF Netes: The "Site specific list of natural or technological hazard events" should include other events that may be a precursor to a more significant event or condition , and that are appropriate to the site location and characteristics. Notwithstanding the events specifically included as EALs above , a " Site specific list of natural or technological hazard events" need not include short lived events for which the extent of the damage and the resulting consequences can be determined 1.vithin a relatively short time frame. In these cases , a damage assessment can be performed soon after the event, and the plant staff 1.vill be able to identify potential or actual impacts to plant systems and structures.

This 1.vill enable prompt definition and implementation of compensatory or corrective measures with no appreciable increase in risk to the public. To the e>, tent that a short lived event does cause immediate and significant damage to plant systems and structures, it will be classifiable under the Recognition Category f, S and C ICs and EALs. Events of lesser impact would be e>,pected to cause only small and localized damage. The consequences from these types of events are adequately assessed and addressed in accordance with Technical Specifications.

In addition , the occurrence or effects of the event may be reportable under the requirements of 10 CFR 50.72. EGL Assignment Attributes:

3.1.1.A and 3.1.1.C 189 NEI 99 Q l (RtwisieR e) Ne¥ember 2012 HU4 ECL: Not ificati on of U nu s u a l Eve nt Initiating Condition:

F I RE potentially degrading the lev el of safety of the plant. Operating Mode Applicability:

All Emergency Action Levels: Example Emergeeey Aetiee Levels: (1 or 2 or 3 or 4) Note~:

  • T h e E m ergency Director s hould declar e the Unusual Eventevent promptl y upon determining that the app li ca ble time ha s b ee n exceede d , or will lik e l y b e excee ded. H LJ 4.l a. A FIRE i s NOT ext in guis h ed w ithin 15-minutes of ANY of the follow in g FIRE d e t ectio n indi catio ns: H 4.2 H 4.4
  • Report from the field (i.e., visual observation)
  • Receipt of multiple (more than 1) fire alarms or indi cat i o n s
  • Fie ld verificat ion of a s in g l e fire a l ar m AND b. The FIRE i s l oca ted w ithin ANY of the followingTable H-1 plant rooms or areas+ a. (site specific list of plant rooms or areas) Receipt of a sing l e fire alarm ~with n o other indications of a FIRE j. ---AND b. The FIRE i s l ocate d w ithin ANY of the followingTable H-1 plant roo m s or areas (site specific I ist of plant rooms or areas) ----AND c. The ex i ste nc e of a FIRE is not verified within 3 0-minut es of a larm receipt. A FIRE w ithin the plant or ISFSI [forplemts wit.Li an ISFSI eutside t,l1ep/a,9t Preteeted Afe.at-PROTECTED AREA n ot ext in g ui s h ed within 60-minutes of the initi a l report , alar m or indication.

A FIRE w ithin the pl ant or ISF SI [for plants wit.Li an ISFSI eutside the plant Pfflteeted Afe.at-PROTECTED AREA that requires firefighting s upp ort by an offs it e fire response agency to extinguish.

Table H-1 Fire Areas 190

  • I G31 DG and Day Tank Rooms,
  • I G21 DG and Day Tank Rooms,
  • Battery Rooms,
  • Essential Switchgear Rooms,
  • Cable Spreading Room
  • Torus Room * ~Intake Structure;
  • Pumphouse
  • Drywell,
  • Torus
  • NE. NW. SE Corner Rooms,
  • RHR Va l ve Room,
  • North CRD Area,
  • South CRD Area,
  • Control Building,
  • Panel ICSS/56 Area;
  • SBGTRoom 191 Jl,ffil 99 g I (ReyisioR e) jl,fo,,cemeer 2('.)12 Definitions:

NEI 99 0 I (ReYisioR

6) }fo 1 i8FR88F 20] 2 FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.

PROTECTED AREA: 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. E AL HU4.1 The intent of the 15-minute duration i s to size the FIRE and to discriminate against small FIRES that are readi l y extinguished (e.g., smoldering waste paper basket). In addit ion to alarms , other indications of a FIRE cou ld 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 ofreceipt of the initial alarm , indication or report. EAL HU4.2 This EAL addresses receipt of a sing l e fire a l arm , 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 a l arm. 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 , additiona l time is allowed to verify the valid it y 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.

192

~IBI 99 01 (ReYisioA

6) ~Jovemeer 20 I 2 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.

EAL HU4.3 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. [Sentence for plants with an ISFSI outside the plant Protested i\rea] EAL HU4.4 If a FIRE within the plant or ISFSI [for plants with tm ISFSI eutside the plant PretectedArea]

PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency ( e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish.

Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions. Basis-Related Requirements from Appendix Rand NFPA-805 Criterion 3 of Appendix A to 10 CFR 50 states in part that "structures, systems, and components important to safety shall be designed and located to minimize.

consistent with other safety requirements, the probability and effect of fires and explosions

." The Nuclear Safety Goal (" NSG") in NFPA 805, Section 1.3.1 states, "The nuclear safety goal is to provide reasonable assurance that a fire during any operational mode and plant configuration will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition." When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance because a safe shutdown success path, free of fire damage, must be available to meet the nuclear safety goals, objectives and performance criteria for a fire under any plant operational mode or configuration.

Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety. the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents.

193 J>ffil 99 O 1 (Re,*isioR e) J>fo*yemaer 2012 In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of I-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). Even though DAEC has adopted the alternate approach provided by NFPA-805 in lieu of the deterministic requirements of Appendix R, the 30-minutes to verify a single alarm as used in EAL HU4.2 is considered a reasonable amount of time to determine if an actual FIRE exists without presenting a challenge to the nuclear safety performance criteria.Basis Related Requirements from Appendix R Appendi>l R to 10 CFR 50 , states in part: Criterion 3 of Appendi>, A to this part specifies that "Structures , systems, and components important to safety shall be designed and located to minimize , consistent 1.vith other safety requirements , the probability and effect of fires and e>,plosions." '.¥hen considering the effects of fire, those systems associated 1.vith achieving and maintaining safe shutdovm conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil off. Because fire may affect safe shutdovm systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents.

In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> fire barriers for the enclosure of cable and equipment and associated non safety circuits of one redundant train (G.2.c). As used in EAL #2 , the 30 minutes to verify a single alarm is ,veil within this worst case 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> time period. Depending upon the plant mode at the time of the event, escalation of the emergency classification level wou ld be via IC CA6 or SA-9SA8. De...eloper Notes: The "site specific list of plant rooms or areas" should specify those rooms or areas that contain SAFETY SYSTEM equipment.

As noted in the EALs and Basis section , include the term ISFSI if the site has an ISFSI outside the plant Protected Area. EGL Assignment Attributes:

3 .1.1.A 194 ECL: Notification of Unusual Event NEJ 99 01 (RevisioA

6) ~fo*,<emeer 2012 HU-76 Initiating Condition:

Other conditions exist which in the judgment of the Emergency Director warrant declaration of a fN 0 1 UE. Operating Mode Applicability:

All l Exemple Emergency Action Levels: 1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.

No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS safety systems occurs. Definitions:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or p l acing it in the cold shutdown condition, i nc l uding the ECCS. These systems are classified as safety-related

.A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdovm condition, including the EGGS. Systems classified as safety re l ated. Basis: This IC addresses unanticipated conditions not addressed explicit l y elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fa ll under the emergency classificat i on level description for a NOUE. 195 NEJ 99 QI (RevisioR

6) No 1 t'0FR00F 2()12 HA1 ECL: Alert Initiating Condition:

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

All Example Emergency Action Levels.:.: (1 or 2) A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the (site specific security shift supervision)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 CONTROLL E D 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. Th i s event will require rapid response and assistance due to the possibi l ity 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 Sh i ft Supervision and the Control Room is essentia l for proper c l assification of a security-re l ated 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}.

196 NEI 99 0 I (RevisioA

  • i) NoYeFAeer 2012 As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures ( e.g., evacuation , dispersal or sheltering).

The Alert declaration will also heighten the awareness of Offsite Response Organizations, 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 HA I.I is applicable for any HOSTILE ACTION occurring , or that has occurred , in the OWNER CONTROLLED AREA. This includes any action directed against att-the ISFSI that 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 GRGoffsite response organization s are in a heightened state of readiness.

This EAL is met when the threat-related information has been validated in accordance with fAbnormal Operating Procedure (AOP) 914, Security Events site specific procedure).§.,_

The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft.

The status and size of the plane may be provided by NORAD through the NRC. In some cases , it may not be readily apparent if an aircraft impact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected , although not certain , that notification by an appropriate Federal agency to the site would clarify this point. In this case , the appropriate federal agency is intended to be NORAD , FBI , FAA or NRC. The emergency declaration , including one based on other ICs/EALs , should not be unduly delayed while awaiting notification by a Federal agency. Emergency plans and implementing procedures are public documents; therefore, EALs should 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 should be ~contained in non public documents such as the Security Plan. Escalation of the emergency classification level would be via IC HS 1. DevelopeF Notes: The (site specific security shift supervision) is the title of the on shift individual responsible for supervision of the on shift security force. Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location.

Security sensitive information should be contained in non public documents such as the Security Plan. 197

~ffil 99 0 I (Re*,isioA e) urth 8 . November 201? .r1ue cons1aeration given to the abo*c e ., -**~hoR>0 :'*'°'""***

te soloote0 ovoAls Bosoribo~

iElet~el';J'er

  • .***. EALs "'"'.I ***teio alpha e, imp ementmg proceaures.

Such references h la n e ec~nty Plan ans associates event. For e:irnm I e s ou not contam a re

  • bl . ( * . ----p*e , ao -AL fllay be wer0e8 as " S _eeg,,,za
  • Elesenptieo efthe site specific security shift supervision)." ecunty event #2 , #5 or #9 is reportea by the See the relates Developer Note . A
  • aevelopment of a scheme aefinition for1:h;*~~~~~~BCgefinitions , for guiaance on the ** }>HROLLBD A~ A eCL Assignment Attributes:

3.1.2.D '..:. n. 198 NEI 99 01 (RevisioR e) ?>lovember 2012 HA5 ECL: Alert Initiating Condition:

Gaseous release impeding access to equipment necessary for normal plant operations , cooldo'+'m , or shutdown. Operating Mode Applieal>ility:

All Example Emergeney Aetion Levels: Note: If the equipment in the listed room or area \Vas already inoperable or out of service before the event occurred , then no emergency classification is warranted.

follo'+'i'ing plant rooms or areas: (site specific I ist of plant rooms or areas 1 Nith entry related mode applicability identified) ,t\..l\JD

b. Entry into the room or area is prohibited or impeded. This IC addresses an event involving a release of a hazardous gas that precludes or impedes access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An Alert declaration is warranted if entry into the affected room/area is, or may be , procedurally required during the plant operating mode in effect at the time of the gaseous release. The emergency classification is not contingent upon whether entry is actually necessary at the time of the release. Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the Emergency Director's judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly impede procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis , report of ill effects on personnel, advice from a subject matter ex.pert or operating e>tperience vrith the same or similar hazards. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment , such as SCBAs , that is not routinely employed).

An emergency declaration is not 1 Narranted if any of the following conditions apply.

  • The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous release).

For e>rnmple, the plant is in Mode 1 1.vhen the gaseous release occurs , and the procedures used for normal operation , cooldown and shutdown do not require entry into the affected room unti I Mode 4 . 199

h. h address the r , measures 'tV ie . ludes compensate. Y , tern testing). . :~:::;:::*;;;n;e~d;a~c~ftp~,'i~t)~

'~ th~a~t;m:;c;;Efir:es:u:p:p:re:s:s:10:n::S)~:s*:::re:c:o:r:dk:e:e:pm:g

  • TAe gas ,el ease IS a.:'~~ efa roam a, area (e.g;j is ef!lfl admiaistrot,ve e, erary iaaeeess, , , rea eat,y is req**ff> d "*euld Rel temp . fer wAieA ,eem,a fRe iRSj)eet10as).

f """'Y """'"'* aA "

  • The action I rounds or mu I 'ati, '0 or precau I E norma f consen
  • naturee.g., I easures are o a . The aeeess eeRtro* '." ede a required aeheA. d . te daagerous levels. aeUJally preveat er ,mp . tAe level ef eKygea iR !lie ~=s:.i eaviroameat. -!~" al,le ef redue1ag . I eiag aiflA aa eae * ., 4.ieA eaa lea e ' R aspAy,<iaat is a ga s *IIJ' ats werk l,y merely d1Sp ** al !~tel ef al'0uad 19 ,S , " n moni)', asphyxia , n below the nor Mest eem eRtmtiea ef e*) ge """" deatla. reduces the cone I
  • unconsciousness , or 200 201 NEI 99 O 1 (RevisieR e) },fovemaer 2012 ECL: Alert NEI 99 01 (RevisioR e) ~fo*refReer 2012 HA6HA5 Initiating Condition:

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

Operating Mode Applicability:

All Example Emergency Action Level s: H 65.1 An event has resulted in plant control being transferred from the Control Room to specific remote shutdo 1 Nn panels and local control stations)the Remote Shutdown Panel (1 C388). Definitions:

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 HS 62_. Developer Notes: The "site specific remote shutdovrn panels and local control stations" are the panels and control stations referenced in plant procedures used to cooldown and shutdown the plant from a location(s) outside the Control Room. EGL Assignment Attributes:

3.1.2.B 202 ECL: Alert NEI 99 01 (Re11ision t'i) ~lo*reff!eer 2012 HA7HA6 Initiating Condition:

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

All l E,ample Eme.-geeeyEmergency Action Leve ls: 1 Other conditions exist which , in the judgment of the Emergency Director , indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. Definitions:

HOSTILE ACTION: An act toward DAEC or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities

{i.e .. this may include violent acts between individuals in the owner controlled area). HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. PROJECTILE:

An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an Alert. 203 ECL: Site Area Emergency Initiating Condition:

HOSTILE ACTION within the PROTECTED AREA. Operating Mode Applicability:

All Example EmergeneyEmergency Action Levels: NEI 99 01 (RevisioR e) ~lovemeer 2012 HS1 S 1.1 A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the (site specific security shift supervision)DAEC Security Shift Supervision. Definitions:

HOSTIL E 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 device s 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-ba s ed EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). HOST AGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTIL E 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 acce s s 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 , Safeguard s Contingency Plan [and Independent Spent Fuel Storage Installation Security Program}.

204 NEI 99 01 (RevisioA

6) NoYemeer 2012 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 GRGoffsite 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 atrthe 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 f-73.71 or 10 CFR f-50.72. Emergency plans and implementing procedures are public documents; therefore, EALs should 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 should beis contained in non public documents such as the Security Plan. Escalation of the emergency classification level would be via IC HGl. Develeper Netes: The (site specific security shift supervision) is the title of the on shift individual responsible for supervision of the on shift security force. Emergency plans and implementing procedures are public documents; therefore , EALs should not incorporate Security sensitive information.

This inc l udes information that may be advantageous to a potential adversary , such as the particulars concerning a specific threat or threat location.

Security sensitive information should be contained in non public documents such as the Security Plan. 'Nith due consideration given to the above developer note , EALs may contain alpha or numbered references to selected ev e nts described in the Security Plan and associated implementing procedures.

Such references should not contain a recognizable description of the event. For e).ample, an EAL may be 1.vorded as " Security event #2 , #5 or #9 is reported by the (site specific security shift supervision)." See the related Developer Note in Appendix B , Definitions , for guidance on the development of a scheme definition for the PROTECTED AREA. EGL i\ssignment Attributes:

3.1.3.D 205 l'JEI 99 QI (Re*,isioR e) }'fo*,emaer 2Q 12 HS6HS5 ECL: Site Area Emergency Initiating Condition:

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

All Example Emergency Action Levels: Note: The Emergency Director should declare the Site Area Emergencyevent promptly upon determining that (site specific number the applicable timeof'.20 minutes) has been exceeded , or will likely be exceeded.

a. b. Definitions:

Basis: An event has resulted in plant control being transferred from the Control Room to (site specific remote shutdown panels and_control stations) the Remote Shutdown Panel (1 C388). AND Control of ANY of the following key safety functions is not reestablished within (site specific number of20 minutes).

  • Reactivity control
  • Core cooling [PWR] I RPV water level [BWR]
  • RCS heat removal 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 g ain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time. The determination of whether or not "control" is established at the remote safe shutdown location(s)Remote Shutdown Panel (1 C388-islli based on Emergency Director judgment.

The Emergency Director is expected to make a reasonable , informed judgment within (the site specific time for transfer)

~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 th e r e sult ofa fir e , transfer of control at panels 1 C388, 1 C389, 1 C390. JC391, JC392and JC392 i s required to be completed within 20 minutes." E scalation of the emergency classification level would be via IC FG I or CG 1. 206 Developer Notes: NEJ 99 01 (ReYisioR

6) No1rember 2012 The "site specific remote shutdown panels and local control stations" are the panels and control stations referenced in plant procedures used to eooldown and shutdown the plant from a location(s) outside the Control Room. The "site specific number of minutes" is the time in which plant control must be (or is expected to be) reestablished at an alternate location as described in the site specific fire response analyses.

Absent a basis in the site specific analyses, 15 minutes should be used. Another time period may be used with appropriate basis/justification.

EGL Assignment Attributes:

3.1.3.B 207 ECL: Site Area E m ergency }ffil 99 QI (Re\1 isioR *i) }loyemeer 2012 HS7HS6 Initiating Condition:

Other conditions ex i s t w hi c h in th e judgment of the E mer ge nc y Dir ec tor warrant declaration of a S it e Area Emerge nc y. Operating Mode Applicability:

Al l E .. mple Emergency Action Level s: 1 Other conditions ex i st w hich in the jud gment of the Emerge nc y Dir ector indicat e t ha t events are in progress or hav e occ urr e d which invol ve act ual or lik e l y major failures of plant f un ct ion s n eede d fo r protection of the public or HOSTILE ACTI O N that r es ult s in int ent i o nal d amage or m a lici o u s acts , (1) toward si t e p erso nnel or eq uipm e nt that could l ead to th e lik e l y fai lur e of or , (2) th at prevent effect i ve access to equipme nt needed fo r the protection of the p ubli c. Any releases are not expected to result in expos ur e l eve l s w hich exceed EPA Protective Action Guideline expos ur e l evels b eyo nd the s it e boundar y. 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. terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individual s 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: T hi s IC addresses unanti ci pat e d conditions not ad dr esse d explicitly e l sew h ere but that warrant d e claration of an emergency because conditions exist which are believed b y the E mergenc y Director to fall under the e mer ge nc y classification l eve l d esc ription for a Site Area E mer ge ncy. 208 NE! 99 01 (Re\'isioR 6) ~foYemaer 2012 HG1 ECL: General Emergency Initiating Condition:

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

Operating Mode Applicability:

All Example Emergency Action Level s: H 1.1 a. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the (site specific security shift supervision)DAEC Security Shift Supervision. AND b. EITHER of the following has occurred:

1. ANY of the following safety functions cannot be controlled or maintained.
  • Reactivity control
  • Core cooling [PWR] I RPV water level [BWR]
  • 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. terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). HOST AGE: 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. 209 NEI 99 QI (Re;cisioA a) November 2012 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. 210 Basis: 211 tffil 99 01 (Re1, 1 isioR fi) tfo1, 1 emeer 2012 NEI 99 C:l 1 (Re,*isioA 6) Novemeer 2C:ll2 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 HOSTIL E 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].

E mergency plans and implementing procedures are public documents; therefore, EALs should 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 should beis contained in non public documents such as the Security Plan. 212 Developer Notes: l>ffil 99 Q l (ReYisieA a) l>foi,*ember 2Ql 2 The (site specific security shift supervision) is the title of the on shift individual responsible for supervision of the on shift security force. Emergency plans and implementing procedures are public documents; therefore , EALs should not incorporate Security sensitive information.

This includes information that may be advantageous to a potential adversary , such as the particulars concerning a specific threat or threat location. Security sensitive information should be contained in non public documents such as the Security Plan. With due consideration given to the above developer note , EALs may contain alpha or numbered references to selected events described in the Security Plan and associated implementing procedures. Such references should not contain a recognizable description of the event. For e1rnmple , an EAL may be worded as " Security event #2, #5 or #9 is reported by the (site specific security shift supervision)." 8ee the related Developer Note in Appendix.

B , Definitions , for guidance on the development of a scheme definition for the PROTECTED AREA. EGL Assignment Attributes:

3 .1.4 .D 213 ECL: General Emergency l'>ffil 99 01 (ReYisioR

6) NoYember 2012 HG7HG6 Initiating Condition:

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

Operating Mode Applicability:

All 6 E,ae,ple Emergency Action Leve ls: 1 Other conditions exist which in the judgment of the Emerge ncy Director indicate that events are in progress or have occurred which involve actua l or IMMINENT substantial core degradation or melting with potential for l oss of containment integrity or HOSTILE ACTION that results in an actua l loss of physica l 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 person(s) held as leverage against the station to ensure that demands will be met by the station. IMMINENT:

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

An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a General Emergency.

214 11 SYSTEM MALFUNCTION ICS/EALS NEI 99 01 (ReYisieR e) }foyemaer 2012 Table S 1: Reeognition Categorv "S" Initiating Condition Matrix UNUSUAL EVENT SUl Loss of all offsite AC power capability to emergency buses for 15 minutes or longer. Op. },fades: 1. 2. 3. 4Pewer Operetien, Stertup , Het St611'lde:y, Hat 8/mtdewl'l SU2 illlPL",1'll>lED loss of Control Room indications for 15 minutes or longer. Op. ,\fades: Pewer Operetien , Stertup , Hat Stemie;*, Hat 8!1utdewnL...l,_

3. 4 SUJ Reactor coolant actiYity greater tkan Technical Specification allowable limits. Op. },fades: 1. 2. 3. 4Pewer Operntien, Sf6lrtup, Het Stendh;*, Het Shutdewn SU4 RCS leakage for 15 minutes or longer. Op. }.fades: 1. 2. 3, '/Pewer Operetien , Stertup, Het Stendby , Het Shutdem'l SUS Automatic or manual (trip [PWRJ / scram [BWR]) fails to shutdown the reactor. Op. }.lodes: Pewer OpaGltien}

ALERT SAl Loss of all but one AC power source to emergency buses for 15 minutes or longer. Op. },fades: L...1....J...

1:.Pewer Opeffllien , Stertup , Het Sf6lndby, Het 8!1utdewn SA2 ill~PLA1'Jl>ffiD loss of Control Room indications for 15 minutes or longer witk a significant transient in progress.

Op. Mades: 1. 2. 3. 4 Pewer Operetien , Stertup , Het Stendby , Het Shu1dew1q SAS Automatic or manual (trip [PWR] / scram [BWR]) fails to shutdown the reactor, and subsequent manua l actions taken at the reactor control consoles are not successful in shutting down the reactor. Op. }.{edes: Pewer Operetienl_

2 15 SITE A.REA EMERGENCY SS1 Loss of all offsite and all onsite AC pov,rer to emergency buses for 15 minutes or longer. Op. },fades: 1, 2, 3, 4Pewer Operetien, Stertup, Het Stendb;*, Het Shutdewn SSS Inability to shutdovm the reactor causing a challenge to (core cooling [PWR] I R.0 V water level [BWR]) or RC8 heat removal. Op. }.fades: Pewer Operetienl 1 GENERAL EMERGENCY SGl Prolonged loss of all offsite and all onsite AC power to emergency buses. Op. }.fades: 1, 2. 3, 4Pewer Operetien, Stertup, Het Stendh;*, Het 8!1utdewn

Table intended for use by 1 EAL developers.
Inclusion in licensee I 0 * . d , ocuments 1s not require . L------------------1 UNUSUAL EVENT SU6 Loss of all onsite or offsite eommunieations eapabilities.

Op. },fades: 1, 2, 3 , 4.Pewer Opereti e n , Stcwtup , Hat St a l'ldhy , Het Slw tde wn SU7 Failure to i s olate eontainment or loss of eontainment pressure eontrol. [PWR] Op. Afade s: 1 , 2 , 3, 1. Pewe,* Operatie,"i , Starh , tp, Het Stendhy , Hat Shi1tdewl'I A,LERT SITE AREA EMERGENCY NEI 99 01 (Re,,*isioR e) },foyemeer 2012 GENER .... L EMERGENCY SS8 Loss of all Vital DC SG8 Loss of all AC and SA9 Ha:mrdous event affeeting a SAFETY SYSTEM needed for the eurrent operating mode. Op. Med es: 1, 2 , 3, 4.Pewer Operatiel'I , Startitp , Hat S:emihy , Hat Shutdewl'1 po 1 Ner for 15 minutes or Op. },fades: 1 , 2, 3, 4Pewer Operetien , Startitp , H e t Standby*, Hat Shutdewl'I 216 Vital DC power souree s for 15 minutes or longer. Op. },1edes: 1, 2, 3, 4Pewer Operetien , St*lrt1,!fJ , Het Stendey , Hat Shutdewn ,-------------------, : Table intended for use b)' I EAL de>,<elopers.

lnelusion in lieensee I d * . d , oeuments ts not require . 1 L------------------J ECL: Notification of Unusual Event NE! 99 0 I (ReYisioR
6) NoYember 2012 SU1 Initiating Condition:

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

Power Operation , Startup , Hot Standby , Hot Shutdown.1..1.,_]_

Example Emergency Action Level s: Note: The E mergency Director should declare the Unusual E~vent promptly upon determining that the applicable time 15 minutes has been exceeded , or will likely be exceeded.

S 1.1 Loss of ALL offsite AC power capability to (site specific emergency buses)1A3 AND 1 A4 buses for 15 minutes or longer. Definitions:

Basis: This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC emergency essential buses-. .,_ This condition represents a potential reduction in the level of safety of the plant. The intent of this EAL is to declare an 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 emergency 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. E scalation of the emergency classification level would be via IC SAL De*,zeloper Notes: The " site specific emergency buses" are the buses fed by offsite or emergency AC power sources that supply power to the electrical distribution system that powers SAFETY SYSTEMS. There is typically 1 emergency bus per train of SAFETY SYSTEMS. At multi unit stations , the EALs may credit compensatory measures that are proceduralized and can be implemented 1 within 15 minutes. Consider capabilities such as power source cross ties, " s 1.ving" generators , other power sources described in abnormal or emergenC)' operating procedures , etc. Plants that have a proceduralized capability to supply offsite AC power to an 217 NEl 99 QI (ReyisioR

6) tlo*,<emeer 2Q 12 affeeted unit via a eross tie to a eompanion unit may eredit this power souree in the EAL provided that the planned eross tie strategy meets the requirements of 10 CFR 50.63. EGL Assignment Attributes:

3.1.1.A 218 ECL: Notification of Unusua l Event NEI 99 QI (Re'iisioA

6) No¥ember 2Q 12 SU2SU3 Initiating Condition:

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

Power Operation , Startup , Hot Standby , Hot Shutdown.L..LJ Examf)le Emergency Action Level s: Note: The Emergency Director shou ld declare the Unusual E~ve n t promptly up o n determining that the applicable time 15 minutes h as been exceeded , or w ill l ikely be exceeded.

S 3.1 ++-a.--An UNPLANNED event results in the inability to monitor one or more of the Definitions:

Reactor Power R.0 V '.Vater Level RPV Pressure Primary Containment Pressure Suppression Pool Le 1 rel

  • Suppression Pool Temperature Suppression Pool Temperature Table S-1 Safety System Parameters
  • Reactor power
  • RPV Water Level
  • 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. These systems are 219

}ffil 99 0 I (ReYisioA

6) N0Ye1'l'!eer 2012 classified as safety-related.A system required for safe plant operation, cooling dovm the plant and/or placing it in the cold shutdovm condition, including the EGGS. Systems classified as safety related. 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. Basis: This IC addresses the difficulty associated with monitoring normal p l ant 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 , dig i tal and recorder source within the Control Room. 220 NE! 99 0 I (Re,*isioH

6) 1-foyemeer 2012 An event involving a los s of plant indications , annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required.

The event would be reported if it significantly impaired the capability to perform emergency assessments.

In particular , emergency assessments necessary to implement abnormal operating procedures , emergency operating procedures , and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control , core cooling [PWR] I RPV level [BWR] and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition.

In addition, if all indication so urces for one or more of the listed parameters are lost , then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level [PWR] / RPV water level [BWR] cannot be determined from the indications and recorders on a main control board , the SPDS or the plant computer, the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Esca lation of the emergency classification level would be via IC SA+/-}. De*,releper Netes: In the PWR parameter list column , the "site specific number" should reflect the minimum number of steam generators necessary for plant cooldown and shutdovm. This criterion may also specify whether the level value should be *wide range , narrow range or both , depending upon the monitoring requirements in emergency operating procedures.

Developers may specify either pressurizer or reactor 1 1essel level in the PWR parameter column entry for RCS Le 1 1el. The number , type , location and layout of Control Room indications , and the range of possible failure modes, can challenge the ability of an operator to accurate!)'

determine , within the time period available for emergenC)' classification assessments , if a specific percentage of indications have been lost. The approach used in this EAL facilitates prompt and accurate emergency classification assessment s by focusing on the indications for a selected subset of parameters. By focusing on the availability of the specified parameter values , instead of the sources of those values , the EAL recognizes and accommodates the wide variety of indications in nuclear power plant Control Rooms. Indication types and sources may be analog or digital , safety related or not , primary or alternate , individual meter value or computer group display , etc. /1 , loss of plant annunciators will be evaluated for reportability in accordance

  • with 10 CFR 50.72 (and the associated guidance in 1'ruREG 1022), and reported if it significantly impairs the capability to perform emergency assessments.

Compensatory measures for a loss of annunciation can be readily implemented and may include increased monitoring of main control boards and more frequent plant rounds by non licensed operators.

Their alerting function notwithstanding , annunciators do not provide the parameter values or specific component status information used to operate the plant , or process through AOPs or EOPs. Based on these considerations , a loss of annunciation is considered to be adequately addressed by reportability criteria, and therefore not included in this JC and EAL. 221

~ffil 99 QI (RevisieA

6) ~fo>,*emeer 2012 With respect to establishing event severity , the response to a loss of radiation monitoring data (e.g., process or effluent monitor values) is considered to be adequately bounded by the requirements of 10 CPR 50.72 (and associated guidance in NUREG 1022). The reporting of this event 1,vill ensure adequate plant staff and NRG awareness , and drive the establishment of appropriate compensatory measures and corrective actions. In addition , a loss of radiation monitoring data , by itself , is not a precursor to a more significant event. Personnel at sites that have a Failure Modes and Effects Analysis (H,4EA) included within the design basis of a digital I&C system should consider the FMEA information when developing their site specific EALs. Due to changes in the configurations of SAFETY SYSTEMS , including associated instrumentation and indications , during the cold shutdown , refueling , and defueled modes , no analogous IC is included for these modes of operation. EGL Assignment Attributes:

3.1.1.A 222 2 NEI 99 QI (ReYisioR e) Noyemeer 2Q 12 SU3SU4 ECL: Notification of Unusual Event Initiating Condition:

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

Power Operation , Startup, Hot Standby , Hot Shutdov,mL 2, 3 Example Emergency Action Levels: (1 or 2) (Site specific radiation monitor) reading greater than (site specific value). Pretreatment Off gas System (RM-4104)

Hi-Hi Radiation Alarm. Sample analysis indicates that reactor coolant specific activity is greater than 2.0 µCi/gm dose equivalent 1-131 for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or longerSample analysis indicates that a reactor coolant activity value is greater than an allowable limit specified in Technical Specifications.,,.,.

Definitions:

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-m i nute delay and decay of 1 Ci/sec. F or EAL SU4.2, coolant samples exceeding the 2.0 µCi/gm dose equivalent l-131concentration require prompt action by DAEC Technical Specifications and are representative of minor fuel cladding degradation.

Escalation of the emergency classification level would be via ICs FAl or the Recognition Category A-R I Cs. DevelepeF Netes: For EAL #1 Enter the radiation monitor(s) that may be used to readily identify 1.vhen RCS activity levels e>rneed Technical Specification allovvable limits. This EAL may be developed using different method s and sites should use existing capabilities to address it (e.g., de;, elopment of new capabilities is not required).

E>rnmples of e>C.isting methods/capabilities include:

  • An installed radiation monitor on the letdown system or air ejector.
  • A hand held monitor or deployed detector reading with pre calculated conversion values or readily implementable conversion calculation capability.

223 l>ffil 99 O 1 (Re*,isioR e) November 2012 The monitor reading values should eorrespond to an RCS aetivity leve l approximate l y at , Teehnieal Specification allowable limits. If there is no e>dsting method/capability for determining this EAL , then it should not be included.

IC evaluation will be based on EAL #2. For EAL#2 Developers may reword the EAL to include the reactor coolant activity parameter(s) specified in Teehnieal Specifieations and the assoeiated allowable limit(s) (e.g., va l ues for dose equ i valent I 131 and gross activity, time dependent or transient va l ues, ete.). If this approach is selected, all RCS aetivity allowable limits should be ineluded. EGL Assignment Attributes:

3 .1.1.A and 3 .1.1.B 224 NEI 99 Ql (RevisioA

6) November 2012 SU4SU5 ECL: Notification of Unusual Event Initiating Condition:

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

Po*.ver Operation , Startup , Hot Standby , Hot Shutdownl..,_Ll Exem13le Emergency Action Levels: (1 or 2 or 3) Note: S 5.1 $2 ~3 The Emergency Director should declare the Unusual E~vent promptl y upon determining that the applicable time 15 minutes has been exceeded, or will likely be exceeded.

RCS unidentified or pressure boundary leakage greater than (site specific Yalue) 10 gpm for 15 minutes or longer. RCS identified leakage greater than (site specific Yalue)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~ steam generator tube leakage in a PWR) 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 l eakage. 225 tJEI 99 Ql (RevisieR C:i) tl0 1 1em0er 2Q 12 The release of mass from the RCS due to the as-designed

/expected operation of a relief valve does not warrant an emergency classification.

For PWRs , an emergency classification would be required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flo 1 t't' cannot be isolated). For BWRs , aA 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 E AL. 226 227 ~ffil 99 01 (RevisioR

6) ~fo*remeer 2012 NEI 99 Gl (Re,*isioA
6) l>fo*,*emeer 2G 12 The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage , if possible.

Escalation of the emergency classification level would be via ICs of Recognition Category A-R or F. Develof)eF Notes: Ei\L #1 For the site specific leak rate 1 1alue , enter the higher of l O gpm or the Yalue specified in the site's Technical Specifications for this type of leakage. EAL #2 For the site specific leak rate ,,atue, enter the higher of 25 gpm or the value specified in the site's Technical Specifications for this type of leakage. For sites that haYe Technical Specifications that do not specify a leakage type for steam generator tube leakage , developers should include an EAL for tube leakage greater than 25 gpm for 15 minutes or longer. EGL Assignment Attributes:

3.1. l .,", 228

l>JEI 99 0 I (RevisioA

6) l>fo\'emeer 2012 SU5SU6 ECL: Notification of Unusual Event Initiating Condition:

Automatic or manual (trip [PWR] / scram [BWR]) fails to shutdown the reactor. Operating Mode Applicability:

Power Operationl,2 Nate: A manual action is any operator action , or set of actions , \Yhich 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.

Exemf)le Emergency Action Levels: (1 or 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.

SU6.1 a. An automatic (trip [PWR] / scram [B'.VR]) did not shutdown the reactor. S 6.2 AND b. ANY of the following manual actions taken at 1C05 are successful in lowering reactor power below 5% power a.

  • Mode Switch to Shutdown
  • Alternate Rod Insertion (ARI)A subsequent manual action taken at the reactor control consoles (I C05) is successful in shutting down the reactor. A manual trip ([P'.1/R]

/ scram [B'.1/R])

did not shutdown the reactor. AND b. EITHER of the following: 1. -ANY of the following subsequent manual actions taken at l COS are successful in lowering reactor power below 5% power

  • Mode Switch to Shutdown
  • Alternate Rod Insertion (ARI)A subsequent manual action taken at the reactor control console (1 C05)s is successful in shutting down the reactor. ---__ OR 230 Definitions:

Basis: NEI 99 0 I (Re*,isioA

6) ~lovemeer 2012 2. -A subsequent automatic (trip [PWR] / scram [BWR]) is successful in shutting down the reactor. This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor~ [PWR] I scram [BWR]) that results in a reactor shutdown , and either a subsequent operator manual action taken at the reactor control consoles or an automatic (trip [PWR] I scram [BWR]) is successful in shutting down the reactor._ This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant. Following the failure on an automatic reactor (trip [PWR] I scram [BWR]), operators will promptly initiate manual actions at the reactor control console s to shutdown the reactor (e.g., initiate a manual reactor (trip [PWR] I scram [BWR])). If these manual actions are successful in shutting dov,rn the reactor , core heat generation willscram quickly fall to a level within the capabilities of the plant's decay heat removal systems. 231

. ------------------~JBI 99 g I (ReYisioA e) November 2012 If an initial manual reactor (trip [PWR] I scram [BWR]) is unsuccessful , operators will promptly take manual action at another location(s) on the reactor control console s to shutdown the reactor (e.g., initiate a manual reactor (trip [PWR] I scram [BWR])) using a different switch). Depending upon several factors, the initial or subsequent effort to manually (trip [PWR] I scram [BW~]) the reactor , or a concurrent plant condition , may lead to the generation of an automatic reactor ftrtp [PWR] I scram [BWR]) signal. If a subsequent manual or automatic (trip [PWR] I scram [BWR]) is successful in shutting down the reactor , core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. A manual action at the reactor control console s 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 ftrtp [PWR] I scram [BW1"])). This action does not include manually driving in control rods or implementation of boron injection strategies.

Actions taken at back-panels or other locations with in the Contro l Room , or any loc ation outs ide the Contro l Room , are not cons id ered to be " at the reactor control console s". Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. [BWR] The plant response to the failure of an automatic or manual reactor (trip [PWR] / scram [BWR]) will vary based upon several factors including the reactor power level prior to the event , availability of the condenser , performance of mitigation equipment and actions , other concurrent plant conditions , etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutt in g down the reactor , then the emerge n cy classification l evel wi ll escalate to an Alert via IC SA:)§. Depending upon the plant response , escalation is a l so possible via IC FAl. Absent the plant conditions needed to meet either IC SA:)§ or FAI , 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).A reactor shutdovm is determined in accordance

  • .vith applicable Emergency Operating Procedure criteria.

Should a reactor (trip [PWR] / scram [BWR]) signa l be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied. I

  • If the signal causes a plant transient that should have included an automatic reactor ftrtp [PWR] / scram [BWR]) and the RPS fai l s to automatically shutdown the reactor , then this IC and the EALs are applicab l e, and should be eva lu ated. If the signa l does not cause a plant trans i ent and the (trip [PWR] / scram [BWR]) failure is determined through other means (e.g., assessment oftest results), then this IC and the EALs are not applicab l e and no classification is warranted.

De*1el013er Netes: This IC is applicable in any Mode in which the actual reactor power level could eJrneed the power level at which the reactor is considered shutdown.

A PWR with a shutdown reactor power level that is less than or equal to the reactor power level which defines the lov,er bound of Pov,er Operation (Mode 1) will need to include Startup (Mode 2) in the Operating Mode Applicability.

For example , if the reactor is considered to be shutdov,rn at 3% and Power Operation starts at >5%, then the IC is also applicable in Startup Mode. Developers may include site specific EOP criteria indicative of a successful reactor shutdown in 232 aA EAL staterneAt , the Basis or aoth (e.g., a reactor pov,cer leYel). wm 99 Ql (ReYisioR a) NoYemeer 2Q12 The term " reactor coAtrol coAsoles" may ae replaced with the appropriate site specific term (e.g., rnaiA eoAtrol aoards). EGL AssigArneAt Attriautes:

3.1.1.A 233

1>,'EI 99 QI (Re1, 1 isioA e) l>io't'emeer 2Q 12 SU6SU7 ECL: Notification of Unusual Event Initiating Condition:

Loss of al-I-ALL onsite or offsite communications capabilities. Operating Mode Applicability:

Power Operation , Startup, Hot Standby , Hot Shutdown.L..LJ Example Emergency Action Levels: (1 or 2 or 3) S 7.1 .fill.L 2 S 7.3 Basis: Loss of ALL of the following onsite communication methods: * (site specific list of communications methods) Plant Operations Radio System

  • In-Plant Phone System
  • Plant Paging System (Gaitronics)

Loss of ALL of the following GRGoffsite response organization communications methods: _* _(site specific list of 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 Loss of ALL of the following NRC communications methods: _* _(site specific list of 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 GRGoffsite response organization s 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.). 235

l>IEI 99 0 I (Re*,isioA

6) NoYemeer 2012 EAL SU7.l addresses a total l oss 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 GRGoffsite response organization s of an emergency declaration.

The GRGoffsite response organization s referred to here are-the S t ate of Iowa, Linn County, and Benton County (see Developer Notes). ---EAL SU7.3 addresses a tota l loss of the communications methods used to notify the NRC of an emergency declaration.

DeYelopeF Notes: EAL #1 The " site specific list of communications methods" should include all communications methods used for routine plant communications (e.g., commercial or site telephones , page party systems , radios , etc.). This listing should include installed plant equipment and components , and not items owned and maintained by individuals. EAL #2 The " site specific list of communications methods" should include all communications methods used to perform initial emergency notifications to OROs as described in the site Emergency Plan. The listing should include installed plant equipment and components, and not items ovmed and maintained by individuals.

Ei>cample methods are ring dovm/dedicated telephone lines , commercial telephone lines , radios , satellite telephones and internet based communications technology. In the Basis section , insert the site specific listing of the OROs requiring notification of an emergency declaration from the Control Room in accordance 1 Nith the site Emergency Plan , and typically within 15 minutes. EAL #3 The "site specific list of communications methods" should include all communications methods used to perform initial emergency notifications to the NRG as described in the site E,mergency Plan. The listing should include installed plant equipment and components , and not items ovmed and maintained b)' individuals. These methods are typically the dedicated Emergency Notification System (m-18) telephone line and commercial telephone lines. EGL Assignment Attributes:

3 .1.1.C 237

1'JE I 99 QI (R ev i s i o R e) Novembe r 2Q 1 2 SU7 ECL: Notification of Unusual Event Initiating Condition:

Failure to isolate containment or loss of containment pressure control. [P\"JR] Operating Mode Applicability:

Po,.*,er Operation, Startup, Hot Standby, Hot Shutdov.*n Example Emergency Action Levels: (1 or 2) 1 a. Failure of containment to isolate 'A'hen required by an actuation signal. AND b. ALL required penetrations are not closed within 15 minutes of the actuation signal. 2 a. Containment pressure greater than (site specific pressure).

AND b. Less than one full train of (site specific system or equipment) is operating per design for 15 minutes or longer. Basis: Th is IC addresses a failure of one or more containment penetrations to automatically isolate (close) '.\.'hen required by an actuation signal. It also addresses an event that results in high containment pressure ,,,ith a 239 NE I 9 9 0 I (R ev i s i o R a) No,*embe r 20 1 2 concurrent failure of containment pressure control systems. Absent challenges to another fission product barrier, either condition represents potential degradation of the level of safety of the plant. For EAL 1, the containment isolation signal must be generated as the result on an off normal/accident condition (e.g., a safety injection or high containment pressure);

a failure resulting from testing or maintenance does not warrant classification.

The determination of containment and penetration status isolated or not isolated should be made in accordance v.*ith the appropriate criteria contained in the plant AOPs and EOPs. The 15 minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.

EAL 2 addresses a condition

'Nhere containment pressure is greater than the setpoint at '.*Jhich containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15 minute criterion is included to allo'I-' operators time to manually start equipment that may not have automatically started, if possible.

The inability to start the required equipment indicates that containment heat removal/depressurization systems (e.g., containment sprays or ice condenser fans) are either lost or performing in a degraded manner. This event ,,.,ould escalate to a Site Area Emergency in accordance

,,.,ith IC FS1 if there ,.,.,ere a concurrent loss or potential loss 240 1>IBI 99 QI (Re,*isioA e) "!>lo'iemeer 2Q 12 of either the Fuel Clad or RCS fission product barriers.

Developer Notes: Enter the "site specific pressure" value that actuates containment pressure control systems (e.g., containment spray). Also enter the site specific containment pressure control system/equipment that should be operating per design if the containment pressure actuation setpoint is reached. If desired, specific condition indications such as parameter values can also be entered (e.g., a containment spray flow rate less than a certain value). EAL #2 is not applicable to the U.S. Evolutionary Po'..ver Reactor (EPR) design. Attributes:

241 ECL Assignment 3.1.1.A 242 ~m, 99 Ql (Re,*isioR e) ~fovemaer 2()12 Ne! 99 0 I (ReYisioA

6) :Jlolo 1 ,em0er 2012 SA1 ECL: Alert Initiating Condition:

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

Po 1 Ner Operation , Startup , Hot Standby , Hot Slrntdown.1.1..,_J.

Example Emergency Action Leve ls: Note: The Emergency Director should declare the Alert-event promptly upon determining that the applicable time 15_ minutes has been exceeded, or will likely be exceeded. S 1.1 a. AC power capability to (site specific emergency buses) 1A3 and 1 A4 buses is reduced to a single power source for 15 minutes or longer. AND b. Any /\NYANY additional single power source failure will result in a loss of alt ALL AC power to SAFETY SYSTEMS. Definitions:

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

including the EGGS. Systems classified as safety related. Basis: This IC describes a significant degradation of off site and onsite AC power so urces such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the so l e AC power source may be powering one, or more than one , train of safetyrelated equipment.

This IC provides an esca l ation path from IC SUL An "AC power source" is a source recognized in AOPs and EOPs , and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.

  • A loss of all offsite power with a concurrent failure of all but one emergency power source ( e.g., an onsite diesel generator).
  • A loss of all offsite power and loss of all emergency pO'Ner sources (e.g., onsite diesel generators) 1 Nith a single train of emergency buses being back fed from the unit main generator.
  • A loss of emergency power sources (e.g., onsite diesel generators) with a sing l e train of essentialemergency buses being -eaek-fed from an offsite power source. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. Escalation of the emergency classification level would be via IC SSl. De*;eleper Notes: For a po 1 Ner source that has multiple generators , the EAL and/or Basis section should reflect the minimum number of operating generators necessary for that source to provide required power to 243 Nel 99 QI (Re\1 isioA a) l>lo*,*eme er 2Q 12 an AC emergency bus. For e>rnmple , if a backup pov,er source is comprised of two generators (i.e., two 50% capacity generators sized to feed 1 AC emergency bus), the EAL and Basis section must specify that both generators for that source are operating.

The " site specific emergency buses" are the buses fed by offsite or emergency

/\C po 1 Ner sources that supply power to the electrical distribution system that powers SAFETY 8¥8TEM8. There is typically 1 emergency bus per train of SAFETY 8¥8TEM8. Developers should modify the bulleted e>rnmples provided in the basis section , above , as needed to reflect their site specific plant designs and capabilities.

The EALs and Basis should reflect that each independent offsite pov,er circuit constitutes a single po:wer souroe. For e>rnmple, three independent 345kV offsite power circuits (i.e., incoming power lines) comprise three separate power sources. Independence may be determined from a revie\\' of the site specific UF8AR , 8BO analysis or related loss of electrical power studies. The EAL and/or Basis section may specify use of a non safety related power source provided that operation of this sou roe is recognized in AOPs and EOPs , or beyond design basis accident response guidelines (e.g., FLEX support guidelines).

8uch power sources should generally meet the " Alternate ac souroe" definition provided in 10 CFR 50.2. At multi unit stations , the E ALs may credit compensatory measures that are proceduralized and can be implemented within 15 minutes. Consider capabilities such as power source cross ties, " sv,*ing" generators , other power sources described in abnormal or emergency operating procedures , etc. Plants that have a proceduralized capability to supply offsite AC power to an affected unit Yia a cross tie to a companion unit may credit this power source in the EAL provided that the planned cross tie strategy meets the requirements of 10 CFR 50.63. EGL Assignment Attributes:

3.1.2.B 244 NEI 99 01 (Re\1 isioA e) ~fovemeer 2012 SA2SA3 ECL: Alert Initiating Condition:

UNPLANNED loss of Contro l Room indications for 15 minute s or lon ger wit h a s i gnificant transient in progress.

Operating Mode Applicability:

Power Operation , Startup , Hot Standby , Hot Shutdovml.,_1,_l Example Emergency Action Level s: Note: The Emergency Director should declare the A-left-event promptly upon determining that the applicable time 15_ minutes has been exceeded , or w ill likely be exceeded.

S 3.1 a. An UNPLANNED event results in the inability to monitor one or more Table S-1 parameters from within the Control Room for 15 minutes or longerAn UNPLANl'H3D e'rent results in the inability to monitor one or more of the follo 1 Ning parameters from within the Control Room for 15 minutes or longer. ......-Table S-1 Safety System Parameters

  • Reactor power
  • RPV Water Level
  • Suppression Pool Level
  • Suppression Pool Temperature PeweF RPV Water Level R1)V Pressure Primary Containment Pressure Suppression Pool Level Suppression Pool Temperature Suppression Pool Temperature AND -Reactor ___ b __ . __ ANY of the Table S-2 transient events are in progress.

245

-Table S-2 Significant Transients

  • Automatic or manual runback greater than 25% thermal reactor power
  • Electrical l oad rejection greater than 25% full electrical load
  • Thermal power oscillations greater than 10% transient eyents in progress.

}>JEI 99 QI (Re\*isioA

6) }>Jo\*emeer 2Q 12 of the following Automatic or manual runback greater than 25% thermal reactor power Electrical load rejection greater than 25% full electrical load Reactor scram [BWR] / trip [PWR] EGGS (81) actuation Thermal power oscillations greater than 10% [BWR] 246 Definitions:

~JEJ 99 0 I (Re\*isieA e) ~Je\1 eme er 2012 SAFETY SYSTEM: A system required for safe plant operation, coo l ing down the p l ant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.A system required for safe plant operation, cooling dov-m the plant and/or placing it in the cold shutdov,rn condition, including the EGGS. 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. 247 NEI 99 0 I (RevisioA

6) ])ofo¥emeer 2012 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 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 p l ant parameters associated with the key safety functions of reactivity control , core cooling [PWR] I RPV level [BWR] and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition.

In addition , if all indication sources for one or more of the listed parameters are lost, then the ability to determ i ne the values of other SAFETY SYSTEM parameters may be impacted as well. For example , if the value for reactor vessel level [PW~] I RPV water level [BWR] cannot be determined from the indications and recorders on a main control board , the SPDS or the p l ant 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 I Cs FS 1 or IC AS+RS 1. Developer Notes: In the PWR parameter list column , the " site specific number" should reflect the minimum number of steam generators necessary for plant cooldown and shutdown. This criterion may also specif)' \\'Aether the level value should be vride range, narrow range or both , depending upon the monitoring requirements in emergency operating procedures. Developers may specify either pressurizer or reactor vessel level in the P'.VR parameter column entry for RCS Level. Developers should consider if the "transient events" list needs to be modified to better reflect site specific plant operating characteristics and eJ(pected responses.

The number, type , location and layout of Control Room indications , and the range of possible failure modes , can challenge the ability of an operator to accurately detennine , within the time period available for emergency classification assessments , if a specific percentage of indications have been lost. The approach used in this EAL facilitates prompt and accurate emergency classification assessments by focusing on the indications for a selected subset of parameters.

By focusing on the availability of the specified parameter values , instead of the sources of those values , the EAL recognizes and accommodates the 1.,*ide variety of indications in nuclear power plant Control Rooms. Indication types and sources may be analog or digital, safety related or not , primary or alternate , individual meter value or computer group display , etc. 248 Ne! 99 QI (RevisioR e) Jlolovemeer 2Q 12 A loss of plant annunciators will be evaluated for reportability in accordance with 10 CPR 50.72 (and the associated guidance in NUREG 1022), and reported if it significantly impairs the capability to perform emergency assessments.

Compensatory measures for a loss of annunciation can be readily implemented and may include increased monitoring of main control boards and more frequent plant rounds by non licensed operators.

Their alerting function notwithstanding , annunciators do not provide the parameter values or specific component status information used to operate the plant , or process through AOPs or EOPs. Based on these considerations , a loss of annunciation is considered to be adequately addressed by reportability criteria , and therefore not included in this IC and EAL. With respect to establishing event severity , the response to a loss of radiation monitoring data (e.g., process or effluent monitor values) is considered to be adequately bounded by the requirements of 10 CPR 50.72 (and associated guidance in NUREG 1022). The reporting of this event will ensure adequate plant staff and NRG awareness , and drive the establishment of appropriate compensatory measures and corrective actions. In addition , a loss of radiation monitoring data , by itself , is not a precursor to a more significant event. Personnel at sites that have a Failure Modes and Effects Analysis (FMEA) included within the design basis of a digital J&C s;'stem should consider the FMEA information when developing their site specific EALs. Due to changes in the configurations of 8AFETY 8¥8TEM8 , including associated instrumentation and indications , during the cold shutdown , refue l ing , and defueled modes , no analogous IC is included for these modes of operation.

EGL Assignment Attributes:

3.1.2.B 249

ECL: Alert }ffil 99 0 I (ReYisioA

6) }lo\1 emeer 2012 SA5SA6 Initiating Condition:

Automatic or manual (trip [P'.l/R] / scram [BWR]) fails to shutdown the reactor , and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor. Operating Mode Applicability:

Pov,er Operation.1.....1 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. l E,ample EmeFgeeeyEmergency Action Level s: 1 a. An automatic or manual (trip [PWR] / scram [BWR]) did not shutdown the reactor. AND b. ALL of the following manual actions taken at 1 COS are not successful in lowering reactor power below 5% power

  • Mode Switch to Shutdown
  • Alternate Rod Insertion (ARI)Manual actions taken at the reactor control consoles (1 COS) are not successful in shutting down the reactor. Definitions:

Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor ftrtJ3 [PWR] I s cram [BWR]) that results in a reactor s hutdown , 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 emer g ency declaration is required even if the reactor is s ubsequently 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 ftrtJ3 [PWR] I scram [BWR])). This action does not include manually driving in control rods or implementation of boron injection strategies.

If this action(s) is unsuccessful , operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers).

_Actions taken at back-panels or other locations within the Control Room , or any location outside the Control Room , are not considered to be " at the reactor control consoles.,_-:-251 NEI 99 01 (RevisioA a) l>fovemeer 2012 Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. [BWR] 252

NEI 99 0 I (RevisioA

6) ]I,[ 0\'emeer 2012 The plant response to the failure of an automatic or manual reactor (trip [PWR] I scram [BWR]) will vary based upon several factors including the reactor power level prior to the event , availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions , etc. If the failure to shutdown the reactor is prolonged enough to cause a challenge to the core cooling [PWR] / RPV water level [BWR] or RCS heat removal safety functions , the emergency classification level will escalate to a Site Area E mergency via IC SS~§. Depending upon plant responses and symptoms, escalation is also possible via IC FSl. Absent the plant conditions needed to meet either IC SS~§ or FSl, an Alert declaration is appropriate for this event. It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however , this IC and EAL are included to ensure a timely emergency declaration. The reactor should be considered shutdown when it is producing less heat than the maximum decay heat load for which the SAFETY SYSTEMS are designed (typically 3 to 5% power).A reactor shutdovm is determined in accordance
  • .vith applicable Emergency Operating Procedure criteria.

Develeper Netes: This IC is applicable in any Mode in which the actual reactor power level could exceed the pov,rer level at which t he reactor is considered shutdown.

A PWR *.vith a shutdown reactor power level that is less than or equal to the reactor power level which defines the lower bound of Pmver Operation (Mode l) will need to include Startup (Mode 2) in the Operating Mode Applicability. For example , if the reactor is considered to b e shutdovm at 3% and Power Operation starts at >5%, then the IC is also applicable in Startup Mode. Developers may include site specific EOP criteria indicative of a successful reactor shutdown in an EAL statement , the Basis or both (e.g., a reactor power level). The term " reactor control consoles" may be replaced v r ith the appropriate site specific term (e.g., main control boards). EGL Assignment Attributes

3.1.2.B 254

Nm 99 QI (ReYisioR

6) ~loYemller 2Q 12 SA9SA8 E CL: A l ert Initiating Condition:

Hazar d o u s eve nt affec tin g a SAFETY SYSTEM n ee d e d for th e c u rre nt o p e ratin g m o d e. Operatin g Mode Applicability:

Power Operation , Startup , Hot Standby , Hot Shutdovm.1..1.,_]_

Example Emergenc y A ction Le v el s: Notes: S 8.1

  • If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardot1s event occurred, then this emergency classification is not warranted. -If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of the SAFETY SYSTEM, then this emergency classification is not warranted .
  • a. AND b. The occurrence of ANY of the Table S-3 hazardous events: ----l Table S-3 Hazardous Events -* Seismic event (earthquake)
  • Internal or external flooding event
  • FIRE
  • EXPLOSION
  • Othe r events w i th s i milar hazard c h aracteristics as determined by the Sh i ft Manager or Emergency Director Director 1. 2. Event damage has caused indications of degraded performance in one train of a SAFETY SYSTEM needed for the current operating mode. AND EITHER of the following:

256

~IEJ 99 Q I (Re,,isioA

6) *Novemeer 2Q 1 2
  • Event damage has caused indications of degraded performance to a second train of t h e SAFETY SYSTEM needed for the current operating mode,
  • The event has resulted in VfSTBLE DAMAGE to the second train of a SAFETY SYSTEM needed for the current operating mode.;: E 1 ,rent damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode, or The eyent has resu l ted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode. Loss of the safety function of a single train SAFETY SYSTEM. 257
  • ------------------------Definitions:

NEI 99 0 I (ReYisioA e) 1-Jo,,ember 20 12 EXPLOSION:

A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization.

A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion.

Such events may require a post-event inspection to determine if the attributes of an explosion are present. FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, inc l uding the ECCS. These systems are classified as safety-related.A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdov,rn condition, including the BCCS. Systems classified as safety related. VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis.

The visua l 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. EITHER of the following:

1. event damage has caused indications of degraded performance in at least one train of a SAFBTY 8Y8TBM needed for the current operating mode. OR 2. The e1,*ent has caused V18IBLB DAMAGE to a SAFBTY 8¥8TBM component or structure needed for the current operating mode. 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 SA98.1.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. An event affecting a single-train SAFETY SYSTEM (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the one train) would not be classified under SA8 because the two-train impact criteria that underlie the EALs and Bases would not be met. If 2 58 Ne! 99 01 (Re;1 isioR e) ~J o*temeer 2012 a n event affects a single-train SAFETY SYST E M, then the emergency cl ass ification should be made based on plant parameters

/symptoms meeting the E ALs for another TC. Depending upon the circ u mstances, classificatio n may also occur based o n Shift Manager/Emergency Director judgement.

Ind i cations of degraded performance addresses damage to a SAFETY SYSTEM tra i n that is i n serv i ce/operation s i nce indications for it wi ll be readily available.

The indications of degraded performance should be significa n t enough to cause concern regarding the operabi l ity or reliability of the SAF E TY SYSTEM train. 259

~JEI 99 QI (ReYisioA e) No,*ember 2Q 12 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. 261 Tl-IE! 99 0 I (Re,,*isioA

6) November 2012 This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier , and therefore represents an actual or potential substantial degradation of the level of safety of the EAL l.b.l addresses damage to a SAFETY SYSTEM train that is in serYice/operation since indications for it 1,vill be readily available.

The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. EAL 1.b.2 addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components.

Operators will make this determination based on the totality of available eYent and damage report information.

This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. Escalation of the emergency classification level would be via IC FS 1 or Ml-RS 1. Develeper Netes: 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 t>.¥0 or more separate and redundant trains of equipment in accordance with site specific design criteria.

EGL Assignment ,r\ttributes:

3.1.2.B 262 ECL: Site Area Emergency NEI 99 0 I (Re*,isioR

6) No*,emeer 2012 551 Initiating Condition:

Loss of ALLa-lt offsite and al-I-ALL onsite AC power to emergency essential-buses for 15 minutes or longer. Operating Mode Applicability:

Pov,*er Operation , Startup , Hot Standby , Hot Shutdownl....1.J Example Emergency Action Level s: Note: The Emergency Director should declare the Site Area Emergencyevent promptly upon determining that the applicable time 15 minutes has been exceeded , or will likely be exceeded.

S 1.1 Loss of ALL offsite and ALL onsite AC power to (site specific emergenC)' buses) 1 A3 and 1A4 buses for 15 minutes or longer. Definitions:

SAFETY SYSTEM: A s ystem 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.A system required for safe plant operation, cooling do 1.vn the plant and/or placing it in the cold shutdovm condition, including the EGGS. Systems cla s sified a s safety related. Basis: This IC addresses a total loss of AC power that compromises the performance of a ll 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. Esca l ation of the emergency c l assification l eve l wou ld be via I Cs AG+RG 1 , FG I or SG I. De*;eloper Notes: For a power source that has multiple generators , the EAL and/or Basis section should reflect the minimum number of operating generators necessary for that source to provide adequate power to an AC emergency bus. For e>rnmple , if a backup power source is comprised of two generators (i.e., tv,o 50% capacity generators sized to feed 1 AC emergency bus), the E AL and Basis section must specify that both generators for that source are operating.

The " site specific emergency buses" are the buses fed by offsite or emergency AC power sources that supply power to the electrical distribution system that powers SAFETY SYSTEMS. There is typically l emergenC)' bus per train of SAFETY SYSTEMS. 263 _J NBT 99 0 I (RevisioA

6) November 2012 The EAL and/or Basis section rnay specify use of a non safety related power source pro 1 1ided that operation of this source is controlled in accordance with abnorrnal or ernergency operating procedures , or beyond design basis accident response guidelines (e.g., FLEX support guidelines).

Such power sources should generally rneet the "Alternate ac source" definition provided in 10 GFR 50.2. At rnulti unit stations , the EA.Ls rnay credit compensatory rneasures that are proceduralized and can be irnplernented within 15 minutes. Consider capabilities such as power source cross ties, " swing" generators , other po 1.Yer sources described in abnormal or ernergency operating procedures , etc. Plants that have a proceduralized capability to supply offsite AG power to an affected unit via a cross tie to a cornpanion unit may credit this po 1.ver source in the EAL provided that the planned cross tie strategy meets the requirernents of 10 GFR 50.63. EGL Assignrnent Attributes:

3.1.3.B 264 ECL: Site Area Emergency

~JEI 99 0 I (RevisioA

6) November 2012 SS8SS2 Initiating Condition:

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

l, 2, 3 Emergency Action Levels: Note: The Emergency Director should declare the Site Area Emergenc)'event promptly upon determining that the applicable time 15 minutes has been exceeded, or will likely be exceeded.

~S+=-2~. l __ Indicated voltage is less than (site specific bus voltage value) 105 VDC on ALL(site specific Vital DC busses) BOTH Div 1 and Div 2 125 VDC buses for 15 minutes or longer. Definitions:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.A system required for safe plant operation, cooling do,Yn the plant and/or placing it in the cold shutdown condition, including the EGGS. S)'stems classified as safet)' related. Basis: This IC addresses a loss of Vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown , this condition involves a major failure of plant functions needed for the protection of the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via I Cs AG-1-RG 1 , FG 1 or SG2. 265

~ml 99 Ql (RtwisioA a) No*,emaer 2Q12 SS5SS6 ECL: Site Area Emergency Initiating Condition:

Inability to shutdown the reactor causing a challenge to (core cooling [PWR] I RPV water level [BWR]) or RCS heat removal. Operating Mode Applicability:

Power OperationLl Examf)le Emergency Action Levels: S 6.1 a. b. An automatic or manual (trip [PWR] / scram [BV/R]) did not shutdown the reactor. AND ALL of the following manual actions taken at 1 COS are not successful in lowering reactor power be l ow 5% power:

  • Mode Switch to Shutdown
  • Alternate Rod Insertion (ARI)AII manual actions to shutdown the reactor have been unsuccessful.

AND c. EITHER of the following conditions exist: Definitions:

Basis: _* _(Site specific indication of an inability to adequately remove heat from the core) Reactor vessel 'NaterRPV level cannot be restored and maintained above -25 inches. * (Site specific indication of an inability to adequately remove heat from the -RGSjHCL (Graph 4 of EOP 2) exceeded.

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor ftFtp [PWR] I scram [BWR]) that results in a reactor shutdown , all subsequent operator actions to manually shutdown the reactor are unsuccessful , and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.

In some instances , the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition 266 NE! 99 QI (ReYisieR

6) "t-Je¥ember 2Q 12 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 shou l d 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). A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria. Esca lati on of the emergency c l assification le ve l would be via IC AG+-RG 1 or FG 1. De*,relof)er Notes: This IC is applicable in any Mode in which the actual reactor povt'er level could e,rneed the power level at vt'hich the reactor is considered shutdown.

A PWR with a shutdown reactor power level that is less than or equal to the reactor power level which defines the lower bound of Power Operation (Mode 1) will need to include Startup (Mode 2) in the Operating Mode Applicability.

For e,rnmple , if the reactor is considered to be shutdown at 3% and Power Operation starts at >5%, then the IC is also applicable in Startup Mode. Developers may include site specific EOP criteria indicative of a successful reactor shutdown in an EAL statement, the Basis or both (e.g., a reactor power level). Site specific indication of an inability to adequately remove h e at from the core: [BWR] Reactor vessel water level cannot be restored and maintained above Minimum Steam Cooling RPV Water Level (as described in the EOP bases). [PWR] Insert site specific values for an incore/core e,cit thermocouple temperature and/or reactor vessel water level that drives entry into a core cooling restoration procedure (or otherwise requires implementation of prompt restoration actions).

Alternately , a site may use incore/core e J lit thermocouple temperatures greater than l , 200 6 F and/or a reactor vessel water level that corresponds to apprmcimatel)

' the middle of active fuel. Plants vt'ith reactor vessel level instrumentation that cannot measure down to approximately the middle of active fuel should use the lowest on scale reading that is not above the top of active fuel. If the lo'Nest on scale reading is above the top of active fuel , then a reactor vessel level value should not be included. For plants that have implemented Westinghouse Owners Group Emergency Response Guidelines , enter the parameters used in the Core Cooling Red Path. Site specific indication of an inability to adequately remove heat from the RCS: [BWR] Use the Heat Capacit)' Temperature Limit. This addresses the inability to remove heat via the main condenser and the suppression pool due to high pool water temperature.

[PWR] Insert site specific parameters associated with inadequate RCS heat removal via the steam generators. These parameters should be identical to those used for the Inadequate Heat Removal threshold Fuel Clad Barrier Potential Loss 2.B and threshold RCS Barrier Potential Loss 2.A in the PWR EAL Fission Product Barrier Table. EGL Assignment Attributes:

3.1.3.B 267 SS8 ECL: Site Area EmergeRcy Initiating Condition:

Loss of all Vital DC power for 15 miRutes or loRger. --------------~ 1%1 99 GI (ReYisioA

6) November 2G 12 Of)erating Mode Af)f)lieability:

Power OperatioR , Startup , Hot 8taRdb;*, Hot 8hutdovm1, 2, 3, 4 Examf)le Emergeney Aetion Levels: Note: The EmergeRcy Director should declare the Site Area EmergeRcy promptly upoR determiRiRg that 15 miRutes has beeR e)weeded, or will likely be e)weeded.

l lAdicated voltage is less thaR (site specific bus voltage value)l 15 VOC OR ALL (site specific Vital DC busses) 1(2) D O l, D 02, D 03, aRd D 04 for 15 miRutes or loRger. Basis+ SAFETY 8Y8TEM: /*, system required for safe plaRt operatioR, cooliRg dowR the plaRt aRd/or placiRg it iR the cold shutdovm coRditioR, iRcludiRg the EGGS. Systems classified as safety related. This IC addresses a loss of Vital DC power which compromises the ability to moRitor aRd coRtrol SAFETY 8Y8TEM8. IA modes above Cold 8hutdowR , this coRditioR iRvolves a major failure of plaRt fuRctioRs Reeded for the protectioR of the public. FifteeR miRutes was s~lected as a threshold to e,wlude traRsieRt or momentary power losses. EscalatioR of the emergeRcy classificatioR level 1 Nould be via ICs AGlB.Ql , FGl or 808. DeYelof)er Notes: The "site specific bus Yoltage value" should be based OR the miRimum bus voltage Recessary for adequate operatioR of SAFETY 8Y8TEM equipmeRt.

This voltage value should iRcorporate a margiR of at least 15 miRutes of operatioR before the oRset of iRability to operate those loads. This voltage is usually Rear the miRimum voltage selected wheR battery siziRg is performed. The typical value for aR eRtire battery set is appFO>(imately 105 VDC. For a 60 cell striRg of batteries , the cell voltage is apprmdmately 1.75 Volts per cell. For a 58 striRg battery set , the miRimum voltage is approximately 1.81 Volts per cell. The " site specific Vital DC busses" are the DC busses that provide moRitoriRg aRd coRtrol capabilities for SAFETY 8Y8TEM8. EGL AssigRmeRt Attributes:

3.1.3.B 268 1'IBI 99 O I (ReYisioR e) NoYemeer 2012 SG1 ECL: General Emergency Initiating Condition:

Prolonged loss of al-l-ALL offsite and ALLalt onsite AC power to emergency essential buses. Operating Mode Applicability

Power Operation , Startup , Hot Standby , Hot ShutdovmL..1.,__J Example Emergency Action Level s: Note: The Emergency Director should declare the General Emergencyevent promptly upon determining that (site specific hours) the applicable time 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> has been exceeded, or will likely be exceeded. a. Loss of ALL offsite and ALL onsite AC power to 1A3 and 1 A4 specific emergency buses). AND b. EITHER of the following:

Definitions:

_* _Restoration of at least one AC emergency essential bus in less than specific hours)4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely. OR * (Site specific indication of an inability to adequately remo*,e heat from the serejRPV level cannot be restored and maintained above -25 inches. 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.A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the EGGS. Systems classified as safety related. Basis: Th i s IC addresses a prolonged loss of all power sources to AC emergency 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/press u re 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 capabi l ities may be degraded under these cond i tions. 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 offsite protective actions. Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency essential bus by the end of the analyzed 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> station blackout coping period. Beyond this time , plant responses and event trajectory are subject to greater uncertainty, and there is an increased l i kelihood of challenges to multiple fission product barriers.

269

~JEI 99 QI (ReYisioA e) ~Je*remeer 2Q 12 The estimate for restoring at least one essentialemergency 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. 270

NEI 99 0 I (RevisioA

6) No¥ember 2012 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. De*,reloper Notes: Although this JG and EAL may be vie*,*,ced as redundant to the Fission Product Barrier IGs , it is included to provide for a more timely escalation of the emergency classification level. The " site specific emergency buses" are the buses fed by offsite or emergency AG pov,*er sources that supply power to the electrical distribution system that powers SAFETY SYSTEMS. There is typical I)' 1 emergency bus per train of SAFETY SYSTEMS. The "site specific hours" to restore AG power to an emergency bus should be based on the station blackout coping analysis performed in accordance with 10 GFR § 50.63 and Regulatory Guide 1.155 , Stetien Bkwkeut. Site specific indication of an inability to adequately remove heat from the core: [BWR] Reactor vessel ;vater level cannot be restored and maintained above Minimum Steam Cooling RPV \1/ater Level (as described in the EOP bases). [PWR] Insert site specific values for an incore/core e>dt thermocouple temperature and/or reactor vessel water le*,rel that drive entry into a core cooling restoration procedure (or othervrise requires implementation of prompt restoration actions).

Alternately , a site may use incore/core e>(it thermocouple temperatures greater than l , 200°F and/or a reactor vessel water level that corresponds to appro>(imately the middle of active fuel. Plants with reactor vessel level instrumentation that cannot measure down to approximately the middle of active fuel should use the lowest on scale reading that is not above the top of active fuel. If the lowest on scale reading is above the top of active fuel , then a reactor vessel level value should not be included.

For plants that have implemented Westinghouse Ovmers Group Emergency Response Guidelines , enter the parameters used in the Gore Cooling Red Path. EGL Assignment Attributes:

3.1.4.B 272 NEI 99 0 I (ReYisioR

6) r>Jovemeer 2012 SG8SG2 ECL: General Emergency Initiating Condition:

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

Power Operation , Startup , Hot Standby , Hot Shutdown.l.,_l,_l Example Emergency Action Level s: Note: The Emergency Director should declare the General Emergencyevent promptly upon determining that the applicable time 15 minutes has been exceeded , or will likely be exceeded.

a. b. Definitions:

Loss of ALL offsite and ALL onsite AC power to (site specific emergency

---b-u ... s-e-s)+-1A3 and 1A4 buses for 15_-minutes or longer. AND Indicated voltage is less than (site specific bus voltage value) 105 VDC on Abb (site speeifie Vital DC busses) BOTH Div 1 and Div 2 125 VDC buses for 15 minutes or longer. 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.A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the EGGS. Systems elassified as safety related. Basis: This IC addresses a concurrent and prolonged loss of both AC and Vital DC power. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling , containment heat removal/pressure control , spent fuel heat removal and the ultimate heat sink. A loss of Vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challenges to fission product barriers.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met. De,*eleper Netes: The "site specific emergency buses" are the buses fed by offsite or emergency AC pov;er sources that supply power to the electrical distribution system that powers SAFETY SYSTEMS. There is typically l emergency bus per train of SAFETY SYSTEMS. The " site specific bus voltage value" should be based on the minimum bus voltage necessar)' for adequate operation of SAFETY SYSTEM equipment.

This Yoltage value should incorporate a margin of at least 15 minutes of operation before the onset of inability to operate those loads. This voltage is usually near the minimum voltage selected when battery sizing is performed.

273 NE[ 99 0 I (Re*,*isieA a) Nevemeer 20 I 2 The typical value for an entire battery set is apprmlimately l 05 VDC. For a 60 cell string of batteries , the cell voltag e is approximately 1.75 Volts per cell. For a 58 string battery set , the minimum voltage is apprmlimately 1.81 Volts per cell. The " site specific Vital DC busses" are the DC busses that pro11ide monitoring and control capabilities for SAFETY SYSTEMS. This IC and EAL 1.vere added to Revision 6 to address operating experience from the March , 2011 accident at Fukushima Daiichi. EGL Assignment Attributes:

3.1.4.B 2 7 4

}ffil 99 0 I (Re\*isioA e) }fo\1 eme er 20 1 2 APPENDIX A -ACRONYMS AND ABBREVIATIONS AC ......................................................

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

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

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

Abnormal Operating Procedure A .......................................................................................................................................... r, PFJi.4 ........................................................

............................................ /\verage Po 1.ver Range Meter ATWS ........................

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

Anticipated Transient Without Scram .......................................................................................................................................... 8 &¥.' ..........................................................

......................................................... Babcock and \l/ilcox ............................................

.............................................................................................. B HT ....................................................................

................... Boron Injection Initiation Temperature BWR ...............................................................

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

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

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

Committed Dose Eq uivalent CFR .........................

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

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

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

...... Code of Fe deral Re g ul at ions CT~4T/CNMT .................................................................

.............................................. Containment

.......................................................................................................................................... C 8F ....................................................

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

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

................ Critica l Safety Fu n ction .......................................................................................................................................... C 8F8T ...........................................................

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

...... Critical Safety Function Status Tree ..............................................................................

............................................................ D BA .....................................................................

............................................ Design Basis i\ccident DC ...................................................

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

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

....................... Direct Curre nt EAL .......................

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

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

....... E mergency Action Level ECCS ..................

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

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

.........

E mer ge nc y Core Cooling System ECL ....................................

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

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

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

E mer ge nc y Class i fication Level E OF ...................

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

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

............... E mergenc y Operations Fac ilit y E OP .....................

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

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

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

Eme r gency Operating Procedure EPA ......................................................................................

....... Env iron1nental Protection Agency E PG ...........................................................

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

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

E mer ge nc y Procedure Guideline

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

............................................................ E PIP .......................................................

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

.. gmergency Plan Implementing Procedure

.......................................................................................................................................... E PR ...............................

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

................... g yolutionary Po\ver Reactor .......................................................................................................................................... E PRJ ....................................................

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

.......... Electric Power Research Institute

.......................................................................................................................................... g RG ...................................................................................

............... Emergency Response Guide l ine .......................................................................................................................................... F EMA .................

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

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

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

Federal Emergency Management Agency f8AR .......................................................

............................................ fin al Safety Ana l ysis Report GE ...........

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

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

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

...................... General E mergency HC+L ...................................................................

....................... Heat Capacity Temperature Lim i t HPCI .............

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

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

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

High Pressure Coo l ant Injection

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

........................................................................................... H 81 .......................................................................

..........

............................... Human System Interface IC .....................................................................................

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

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

lnitiating Condition Nel 99 01 (Re\'i s ioA e) T>lo¥efl'leer 2012 ...................................................

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

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

................ I D ...............

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

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

Inside Diameter IP E EE. ............................ Individual Plant E>rnmination of External Events (Generic Letter 88 20) ISFSI ...........................................................................

Independent Spent F u el Storage Installation Keff ....................................................................................

Effective Neutron M ultiplic ation Factor LCO ...............................

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

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

......... Limiting Conditio n of Operation

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

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

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

L OCA .........................................................................................

................. Loss of Coo l ant Accident ****************************************************************************************************************************************** ~4 CR ................................................

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

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

........... ~4ain Control Room ..................................

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

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

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

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

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

.. ~4 SIV .............

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

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

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

....................... ~4ain Steam Isolation Valve ~<ISL ...............

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

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

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

....................... ~<fain Steam Line mR , mRem , mrem , mREM ....................

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

........ milli-Roentgen E quivalent Man MW .......................

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

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

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

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

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

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

Nuclear E ner gy Inst itut e ...............

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

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

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

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

l>+ PP ...........

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

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

..........

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

.... l'+uclear Po\ver Plant .......................................................................................................................................... N RC ................................

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

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

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

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

N ucl ear Regulatory Commission NSSS ...............................................................

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

Nuclear Steam Supply System NORAD ................................

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

....... North American Aerospace Defense Com mand fN O)UE ..........................................................

...........

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

fN ot ifi cation O ff Unusual Event NUMARC 1 *************************

          • N ucl ear Management and Resources Co un ci l OBE ....................................................

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

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

..... Operating Basis Eart hquake OCA .......................

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

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

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

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

............. Owner Co ntr olled Area .................................

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

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

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

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

............................... 0 f)GML ODAM .........................................................

Offsite Dose Calculation (Assessment) Man u al ORO .........................

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

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

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

Off site Re s ponse Organiz:ation PA ...........

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

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

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

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

....... Protected Area ...................

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

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

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

................................................. p ACS .........................

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

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

Priority Actuation and Centro I System PAG ........................

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

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

...........

.......... Protective Action Guide lin e *************************

                                        • p JCS ................................................

................................... Proces s Information and Control System PRA/PSA ....................................

Probabilistic Risk Assess m e nt/ Probabilistic Safety Assessment PWR ........................................................................................................

Pressurized Water Reactor ******************

                                                                                      • p S ..........................................................

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

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

.............................. Protection System PSIG ........................................................................................

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

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

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

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

.......................................... Roentgen ........................................................................

.................................................................. R CC ...........................................................

...........

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

..........

.......... Reactor Control Console RCIC .....................

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

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

Reactor Core Isolation Coo lin g 1 NUMARC was a predecess or o r ga nization of the Nuc l ear Energy In st itut e (NE I). A-2 NE! 99 0 I (Re¥isioA

6) ~io¥ember 2012 RCS .........................................................................

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

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

................. Roentgen Equivalent Man .........................

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

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

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

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

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

... R ETS ......................

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

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

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

...... Radiological Effluent Technical Specifications RPS ................................

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

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

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

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

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

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

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

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

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

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

R VLIS ...................

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

...........

.... Reactor Vessel Level Instrumentation System RWCU ......................................................................................

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

................ s AR ....................................

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

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

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

........... Safet)' ,<\nalysis Report ***********************************************************

s AS .......................

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

...........

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

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

........... Safety ,<\utomation System .......................

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

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

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

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

s BO .................................................

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

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

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

..... Station Blackout SCBA ...............

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

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

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

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

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

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

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

..................... s G .........................

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

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

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

......................... Steam Generator

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

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

...........

s I ..........................................

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

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

........................................ Safety Injection

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

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

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

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

.. s JCS ....................................................................

................. Safety Information and Control System ******************************

s PDS ...................

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

................. Safety Parameter Display System SRO .................

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

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

Senior Reactor Operator TEDE .................................

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

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

Tota l Effective Dose Equivalent T G AF ............................................................

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

........................ Top of Active Fue l TSC ..................

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

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

..................................... Technical Support Center ........................

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

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

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

.......................... U FSAR ............................

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

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

Updated Final Safety Analysis Report \VOG ...................

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

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

\Vestinghouse 0 1.vners Group NEI 99 QI (ReYisioA a) November 2Ql2 APPENDIX B -DEFINITIONS t>i:El 99 0 I (Re.,.isioA

6) }fo.,.ember 2012 The following definitions are taken from T itl e 10 , Code of Federal Regulations, and related regu lat ory guidance documents.

Alert: Events are in progress or have occ urr ed 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 actua l lo ss of physical contro l of the facility. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Notification of Unusual Event (NOUE)\ Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.

No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systemsSAFETY SYSTEMS occurs. Site Area Emergency:

Events are in progress or have occurred which involve actua l or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that resu lt s in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to , equipment needed for the protection of the public. _Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the site boundary.

The following are key terms necessary for overall understanding the NEI 99 01DAEC emergency classification scheme. Emergency Actio n Level (EAL): A pre-determined , site-specific, observable threshold for a n Initiating Condition that, when met or exceeded , places the plant in a g i ven 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 conditio n s according to (1) potential or actual effects or consequences , and (2) resulting onsite and offsite response actions. The emergency classification leve l s , in ascending order of severity , are: Notification of Unusual Event (NOUE) Alert Site Area Emergency (SAE) General Emergency (GE) + This term is sometimes shorteAed to UAusual E.,.eAt (VE) or other similar site speeifie termiAology.

B-1

~lEI 99 01 (Re\*isioR

6) November 2012 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: (Insert a site speeific definition for this term.) Developer Note -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.

CONTAINMilff CLOSURE: (Insert a site speeific definition for this term.) DeYeloper Note Site speeific proeedurally defined aetions taken to seeure eontainment and its assoeiated struetures, systems, and eomponents as a funetional barrier to fission product release under e>(isting plant eonditions.

For DAECs, this is considered to be Secondary Containment as required by Teehnieal Specifieations.The proeedurally defined eonditions or aetions taken to secure containment (primary or secondary for BWR) and its assoeiated struetures , systems , and eomponents as a functional barrier to fission product release under shutdovm eonditions.

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. F/\ULTED:

The term applied to a steam generator that has a steam leak on the seeondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to beeome eompletely depressurized.

Developer Note This term is applicable to PV/Rs only. B-2

Ne! 99 0 I (RevisioR

6) ~/0 1 ,emaer 2012 FIRE: Combustion characterized by heat and li ght. Sources of smoke such as slipping drive belts or overheated electrica l eq uipm ent do not constitute FIRES. Observation of flame is preferred but is NOT required if l arge quantities of smoke and heat a r e observed. HOSTAGE: A person(s) held as l everage aga in st the stat i on to ensure that demands will be met by the station. HOSTILE ACTION: An act toward a NP-llnuclear power plant or its personnel that includes the use of violent force to destroy equipment , take HOSTAGES , and/or intimidate the lic ensee to ac hi eve an end._ This includes attack by air , land, or water using guns , explos i ves , PROJECTILEs , veh i cles , or other devices used to de li ver destructive force. _Other acts that sat isf y the overa ll int ent may be inc luded. HOSTILE ACTION shou ld not be construed to include acts of civil disobedience or fe l on iou s acts that are not part of a concerted attack on the NPJ!nuclear power plant._ Non-terrorism-b ased 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 assau l t , overtly or by stea lth and deception , equ ipp ed with su it ab l e weapo n s capable of killing , maiming , or causing destruction.

IMMINENT:

The trajectory of events or conditions is s u ch that an EAL will be met within a re lativel y 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. NOR~4AL LEVELS: As applied to radiologieal IG/EALs , the highest reading in the past twenty four hours exeluding the eurrent peak value. OPERA TING BASIS EARTHQUAKE (OBE): An OBE is vibratory ground motion for which those features of a nuclear power plant necessary for continued operat i on without undue risk to the health and safety of the public will remain functional.

OWNER CONTROLLED AREA: (Insert a site speeifie definition for this term.) Developer Note This term is typically taken to mean the site property owned by , or otherw i se und er the co ntrol of , the licensee.

Tn some cases , it may be appropriate for a lieensee to define a smaller area with a perimeter eloser to the plant Proteeted Area perimeter (e.g., a site with a large OGA where some portions of the boundary may be a signifieant distanee from the Proteeted Area). In these eases , developers should eonsider using the boundary defined by the Restricted or Seeured Ovmer Controlled Area (ROGfJSOGA). The area and boundar)'

seleeted for seheme use must be eonsistent 1 ,vith the deseription of the same area and boundary eontained in the Seeurity Plan. PROJECTILE:

An object directed toward a NP-llnuclear power plant that could cause concern for its continued operability , rel i abi lit y, or personnel safety. B-3 NE! 99 QI (RevisieA e) ~Jevemaer 2Q 12 PROTECTED AREA:_ (Insert a site specific definition for this term.) De~relaper Nate This term is typically taken to mean ti he area under continuous access monitoring and control , and armed protection as described in the site Security Plan. REFUELING PATHWAY:_ (Insert a site specific definition for this term.) Develaper Nate This description should include all the cavities, tubes, canals and pools through which irradiated fuel may be moved , but not including the reactor vessel. Includes all the cavities, tubes, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel. RtJPTURE(D):

The condition of a steam generator in 1 , 1 ,rhich primar)' to secondary leakage is of sufficient magnitude to require a safety injection. Develaper Nate This term is applicable to PWRs only. SAFETY SYSTEM: A system required for safe plant operation , cooling down the plant and/or placing it in the cold shutdown condition , including the ECCS. These are typically systems are classified as safety-related.

Develaper Nate 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 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-4 B-5 NEI 99 g I (Re'lisieA e) NeYember 2012 NEI 99 0 I (Re,*isioA

6) ~tOY6fl'l06F 20 J 2 APPENDIX C PERMANENTLY DEFUElED STATION ICs/EAls Recognition Category PD provides a stand alone set of ICs/EALs for a Permanently Defueled nuclear power plant to consider for use in developing a site specific emergency classification scheme. For development , it was assumed that the plant had operated under a 10 CFR § 50 license and that the operating company has permanently ceased plant operations.

Further , the company intends to store the spent fuel within the plant for some period of time. When in a permanently defueled condition , the plant licensee typically receives approval from the NRG for eJ(emption from specific emergency planning requirements.

These eJ(emptions reflect the lowered radiological source term and risks associated 1.vith spent fuel pool storage relative to reactor at power operation.

Source terms and accident analyses associated

\.Vith plausible accidents are documented in the station's Final Safety Anal)1 sis Report (FSAR), as updated. As a result, each licensee will need to develop a site specific emergency classification scheme using the :NRG approved exemptions , revised source terms , and revised accident analyses as documented in the station's FSAR. Recognition Category PD uses the same ECLs as operating reactors; however , the source term and accident analyses typically limit the ECLs to an Unusual Event and Alert. The Unusual Event ICs provide for an increased awareness of abnormal conditions 1.vhile the Alert ICs are specific to actual or potential impacts to spent fuel. The source terms and release motive forces associated with a permanently defueled plant would not be sufficient to require declaration of a Site Area Emergency or General Emergency.

A permanently defueled station is essentially a spent fuel storage facility 1.vith the spent fuel is stored in a pool of water that serves as both a cooling medium (i.e., removal of decay heat) and shield from direct radiation. These primary functions of the spent fuel storage pee I are th e focus of the Recognition Category PD ICs and EALs. Radiological effluent IC and EALs were included to provide a basis for classifying eyents that cannot be readily classified based on an obserYable e\.1 ents or plant conditions alone. Appropriate ICs and EALs from Recognition Categories A, C , F , H , and S 1 ,\.1 ere modified and included in Recognition Category PD to address a spectrum of the events that may affect a spent fuel pool. The Recognition CategOF)'

PD ICs and EALs reflect the relevant guidance in Section 3 of this document (e.g., the importance of avoiding both over classification and under classification).

Nonetheless , each licensee will need to develop their emergency classification scheme using the NRG approved eJ(emptions , and the source terms and accident analyses specific to the licensee.

Security related events will also need to be considered.

C-1 l>lel 99 0 I (Re1t*isioR

6) l>Jo,*emeer 2012 Table PD 1: Reeognition Category "PD" Initiating Condition MatFix UNUSUAL EVENT PD ,A .. Ul Release of gaseous or liquid radioactivit)*

greater than 2 tirnes the (site specific effluent release controlling docurnent) limits for 60 rninutes or longer. Op. },lodes: llfetApplicehlc PD AU2 UNPLA1'J1'ffiD rise in plant radiation levels. Op. A fede s: Ne tA ppli c ehle PD SUl UNPLA1'J1'ffiD spent fuel pool temperature rise. Op. },lodes: l*let Applicehlc PD HUl Confirrned SECURITY C01'JDITI0N or threat. Op. },lodes: ,VetAppliceblc PD HU2 Hazardous event affecting SAFETY SYSTEM equipment necessary for spent fuel cooling. Op. Modes: ,"hlet Appliceblc PD HUJ Other conditions e>dst which in the judgment of the Emergency Director *.varrant declaration of a (NO)UE. Op. },fedes: ,"hlet Applicehlc ALERT PD AAl Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrern TEDE or 50 rnrern thyroid CDE. Op. },fedes: ,"hletAppliceble PD AA2 UNPLA1'J1'mD rise in plant rad i at i on le;*els that irnpedes plant access required to maintain spent fuel integrity.

Op. },fedes: , Vet Applirnble PDHAl HOSTIL E ACTION v,ithin the OW1'J:ER CONTROLLED AREA or airborne attack threat within 30 minutes. Op. },fedes: A'etApplicehle PD HAJ Other conditions e>dst which in the judgrnent of the Ernergency Director warrant declaration of an Alert. Op. },{edes: ,VetAppliceble

,-------------------, : Table intenaea for use by 1 1 BAL aevelopers.

Inclu s ion in licensee C-2 I ..I , , ..I , uocurnents 1s not requ1reu. 1 L------------------J

---~------------ECL: Notification of Unusual Event ~/EI 99 Ql (Revision

6) ~lovomeor 2Ql 2 PD AU1 Initiating Cenditien:

Release of gaseous or liquid radioactivit)

' greater than 2 times the (site specific effluent release controlling document) limits for 60 minutes or longer. Netes: Operating Mede Applieability:

Not Applicable Example Emergeney Aetien Levels: (l or 2)

  • The Emergency Director should declare the Unusual Event promptl)' 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 unlmown , assume that the release duration has e>weeded 60 minutes.
  • If the effluent flow past an effluent monitor is knovm to have stopped due to actions to isolate the release path , then the effluent monitor reading is no longer valid for classification purposes. (l) (2) Reading on ,A_._..""JY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactiYity 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 (site specific effluent release controlling document) limits for 60 minutes or longer. Basis: This IC addresses a potential decrease in the level of safety of the plant as indicated by a low level radiological release that e>weeds regulatory commitments for an extended period of time (e.g., an uncontrolled release).

It includes any gaseous or liquid radiological release , monitored or un monitored , including those for which a radioactivity discharge permit is normally prepared.

Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment.

Further , there are administrative controls established to prevent unintentional releases , and to control and monitor intentional releases.

The occurrence of an e>(tended , uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent eALs more fully addresses the spectrum of possible accident events and conditions. C-1 NEI 99 o I (Re*,*isieR e) Jl,levemeer 2012 ti:! te the mes that a release pa ed *ter readiAgs assu . . l'Ae"'A te haYe stepp 8 ffiHOR! !HORI ffi t lflORl!eFIS

' j" dfeF ClassifieatieR

_8*s: ~~:e efflHeRt flew past aR * "":,er rea8iRg is "" leRge, va , ,, AmeAt is estabhs!:ie . at!:i tl=ieA the efflueAt me eA~ ire . . late the release p ' 8HO !0 80!l0RS le 199 8' g 4 times *c: f A purpeses.

I a release exoee IA elassiNea-,e-

-8 Fe, °'""'P**, Aet be prerated er averag; . Releases sheHlf AHies Sees Rel !fleet the llnL. a<liatieR ffi0Riter release liffiits ie, 39"' e* aetivity releases that eause h~:*;::lflit.

+his eAL B' L #1 +his on . . lal,lishe8 By a ro ,0 ' tiRH0HS release pa-" c A :b addresses ra 10 d. aot 11*1ty d1so ar t!:i.,*ays n 2

  • s the l1m1t es ffem nen readiAgs te e>,oeed t1~e~ .. :ti:! planAed bato!:i releases "
  • iealiy Be as,eeiate n I 8 B " w1ll l}jl ,.. te gas). . that are Seteete ) (e.g., ra8waste, ..as eR!rnlle8 gaseeus er liq~,e ~'::~::,.ys (e.g., spills ef . g, L addresses""" Rlfl0R1te,e ) B* L #2 +his~ I r>'eys paFtieularly 0R * . . *er>**atersystelfls , ete .. , r
  • eAta su-. ' I I e IA rn sample aAalyses er eAv1reAm draiAs heat e>wl=iaAgerea rng ra<lieaetive liquids iRte stem, ' . . ... 18 Be via IC PD AA!. esoalatien ef the emer geAoy olassifioatieA le,,el weu N t s* n. h Ra<lielegieal Effl**Rt D""eleper e;,
  • eeRtrnlliRg 8ee~ffieRt 19 t
  • C:eRerie Leiter 89 91 , +he "site Sfl**ifie effiueRt ,el~aseplaRts that have 11F1pie1HeRte8 le1HeRt "'gulatieRs related . Ii ati eRs (Rll+S) **, <<>* MJ +hese 8eeulfleRts

'"'P. l ' *l'JJrnp,iate, the +eehRieal Si**: ~aleulatieR MaRual (O~C;;.9 ~FR Part 59 , AppeR81K Lt;,:shtles ier this IC. the Offs1te es E 10 CFR Part 2 aA t blisl=iiAg the memter 11 1 eeRtrnls e.g., 18 B *see Jer es* M te ef HOR he8elegy sheu -e -. 1Uff8 er 00C~. RB+8 er ODCM !flat *10,s 8eseriBe8 IR the

  • e the offiueRt ffieRI . I . RiteFS sh0Hl8 iRelu e . te8 with ether pete*ti* Listed lfl0 . . t lied lfl0RiteFS ass*~** e e B' L values ier i8er i**iu81Rg IRS a {;M"H If IRelu * , fl e . Develepers

"'"Y al:::~::eriBe8 i* tRe llll+8 er~~.;;,~

Sesokelease lilflit:,v:*::::*

tR>>*ays tRat are

  • the 1H0st "PP 1 , Be Belew """ efflueR! pa--.. R 18 Be 8eterlfliAe8 usmg *-I lated BAL value "'a) : R r t Alse , selfle these Fll0Rll0FS s *:i It is reeegaiaee tRat a ea **.1 ROOS 10 Be iRelue.ee IR t * .::.d ,elated the llll'.f8 e, ,?~Ci~ ;Rat ease, the lfl0~te, a~":;e:i!ieatieRs e, ethe~ ~**.:s:.::/eR elearly i8eRtify lfl0R1te, eaR ' .. med By +ee ***
  • 1 e BAL aR '" meRiteFS "'"Y *et B: ge,i: is i1Hp0FtaRt that the assee,~~FS . ts* t!:iere43re , bT . f tliese ffi0R
  • 1 reqH1re1HeR--, --h seer avaihH ,ty e . . I sos with sepa,a-e a*y li1Hitati0*s
    • t *
  • address gaseeus aR8 hqu,e re** 8eme sites m8) . , fiAd it advaAtageeus te u ,..,,,...,...,,.,, ., ~7 {on *'*he 'Feehn,eo.~

{; ' l*o/ P,og,*""'

' I** so,Ha*(*)

wh"* , __ ,. '"""""""'

-~*1 ."'; ',,;;:,,.1

..

  • 10 /he P,*~-*"",; *** ;. 1h, ,;,. .... ,.. ** , ' ; ', Ojfeile Do,e c.,"'"'°":'" flh effi,,at memlo,s '" .. *a O By l>JPO ,.1a1, l,re ........ 8 0 h o ***** , ,,.., O Thi, iAol,Oa, ";~**~FR >o.47(1,)(8)

,.; (9). I ofl O CFR SO .S4 (q) .,; H ~:**ilo,s.

the reE!HiremeRtS e J iR miRe the FeE!HIF~lfle~

S h aeeitieR efether efffoeR H De**elepers sheule ceep . Rt , .. heR eeRs1eermg t e . ro * ..,, . .,..., ,e ' . 'S eifieoi,on, ** ' . h e; ss . , $ 'fiec1tief'ls il'I .'he . , . Bl Ejfh:e:?t Teehme~, ~e;t;:, edt!rnl Detc1i!s &j RET. le ti C,ntrelsfer R86He,egie_ . dtfie Releec1ltel'I 8.rr. ee

  • EfHipme geRey respeRse e te emer C-2 NEI 99 0 I (Re*,isioA e) Jl,lo*,emeer 2012 Radiati_on mo.niter readings should reflect va l ues that correspond to a radiolo ical release exceedmg 2_times a releas_e ~ontrol limit. The controlling document typical I)~ describes

~eetho:olo~ies fo~ dete~mmmg effluent radiat i on monitor setpoints; these methodologies should .~se to etermme En~ values. In cases where a methodology is not adequate! , defined :***1';1'"" skeelB Beterm1Ae valees eeAsisteAt with effleeAt eeAtrel ,egelatieAs

(/g lQ cf~ art and IO CFR Part 50 Appendix T) and related guidance.

  • ., . ~o: EA_L #1 Values in this EAL should be 2 times the setpoint established b , the ::::**elivity 8 15 ei,arge P*Fffi it le wam ef * ,el ease tkat is Aot iA eompl iaAee with tke ;peeifieB ts: Inde)ong the *~alue m this manner ensures consistency betv~*een the EAL and th t
  • established by a speeific discharge permit. e se13omt Developers should researeh radiati~n monitor design documents or other information s~uree~ to ensure th~t I) the EAL value bemg considered is within the usable response and displ_a) rang~ of ~he 1~strument , and 2) there are no automatic features that may render the momtor readmg mvahd (e.g., an auto purge feature triggered at a particular indication level). ffl It is recognized that the condition described by this IC may result in a radio l ogical e uen~ :*~ue ~eyond the operating or display range of the installed effluent monitor In those case~~ 1' ;.a u~s sho~ld be determined with a margin sufficient to ensure that an ac.curate mom-or rea-mg 1s a"a1lable Fe I c AL *
  • f .* ..r ex.amp e , an en momtor readmg might be set at 90% to 95% o~~he h1g7est accurate m~mtor re~dmg. This provision notwithstanding , if the est1matedrcalculated monitor read mg is greater than approx i mately 110% of the h" h t *t d. h ' 1g es accurate ~oni:~r rea mg , t en developers may choose not to include the monitor as an indication and 1 ent1 'an alternate EAL threshold. , _Indications from a real _time do~~ projection system are not included in the eneric EALs t~aA) heeR;ees Bo AO! ka*,e th!S eapah,hty.

eor those that Bo, tke eapal,ility may A:t ~e ***itkiA * --e scope o-the p l t T h

  • I 8 *fi
  • 1-~ 1 f d~n :ec mcapec1 cations. A licensee may request to inelude an EAL using rea ,me ose prOJectJon s ystem results; approval ',¥ill be considered on a case by case ;asis. C f; 4 /ndications from a perimeter monitoring system are not included in the generic EA Ls
  • any 1censees d~ no~ have this capability.

for those that do , these monitors may not

  • T:::o~le~
nd ;amt~med to the ~~me I eve~ as plant equipment , or within the scope of the plant mca 6 ~eCI cations. In add1t10n , readmgs may be influenced by environmental or other actor~. 1 1 1:.~1censee m_ay request to include an EAL using a perimeter monitoring system; a ppro1-a n ii I be considered on a case by case basis. EGL Assignment Attributes:

3.1.l.B C-3 I-8 "cVV Od: JO l \IV 0d :)l l'l!A eq pf ROA\ f9A9f UO!ll'lO!J!SSBfO

,(ouei5J9UJ9 94lj0 UO!ll'JfBOSff "Sf 13!J9ll3W 9lSl3A\ 9A!l0l'lO!Pl'lJ JO weweAOUJ pua S90JROS 0!4de~O!PBJ JO esn se 4ons se9!A!lOl'l peuue1d WOJJ l[RSeJ ll'l4l seseeJOU!

teAet UO!ltJ!PBJ sepn1oxe c# 1\/!I *doJp 1eAet J9ll3M G!IN:N:V1d:Nfl ua Ol enp S! i5U!pl'leJ pelBAete e4l eJe4"A sestJo LI! ,(1uo e1qeo!Jdde S! t# 1\/'3 ll34l 9lON *peJep!suoo eq p1no4s suO!lRfOAe peuutJjdJO Sl99JJ9 e4+/- "SUO!ll'JOOJ eS04l U! SJOl!UOW ,(q pel09l9P eq UB9 lB4l Sl'JeJe lUeoefpl'lJO Sf9A9f UO!ll'l!Pl'lJ e41 U! 9Sl'J9J0U!

ue esneo OSfl'l Al'JUJ f9A9f J9ll3A\ e41 LI! doJp lUl39!J!Ui5!S y *Ee1qef!BABJ!)

suO!ll'JAJesqo tJJeweo oep!A JO f9UU0SJed lUl3fd UJOJj sµodeJ epRfOU! , (l'JUJ SUO!ll'lO!PU!

f0A9f JO S09JROS J94l0 "UO!ll'JlU0WR.llSU!

f9A9f e1qef !l'lAl'l UJOJJ suO!lBO!PU!

,(q peu!UJJelep

, (f!JBW!Jd eq lf!A*, estJeJoep f9A9f J9ll'J,'A y "lUl3Jd 04lj0 Al9jl3S JO J9A0J 04l U! UO!lBPBJi50p J13!lU9lOd 13 S! UO!l!PUOO J94l!3 " SJ13!J9ll3UJ 9A!l0130!Pl'lJ JO lUl3Jd e41 U!4l!N. SJ9A9f UO!ll3!Pl3J fOJlUOO Ol ,(l!f!ql'l e41 U! sso1 JOU!lli 13JO 9A!ll'l0!PU!

eJe s1eAet UO!lB!P13J peseeJOU!

e4+/- *slueAe G'3N:N:V1d:Nfl Je410 JO 1enJ E1ueds) petB!Ptl"!

eAoqtJ 1eAe1 JeltJNr LI! estJeJoep tJ ,(q pesntJo SfeAef UO!ll'l!Pl'lJ lUBfd peltJAete sesseJppl'l

)I S!4+/- "81!IA311Vt"q1f0N JQAO J4ffllli §cJO es!J G3NNV1d:Nfl ue sell'lO!PU!

lfRSeJ , (e,uns JO i5u!peeJ JOl!UOUJ UO!lll!Pl'lJ tJeJv "SJOl!UOW UO!ll'l!PtlJ i3U!MOjf Oj e4uo _;\~TV' , (q pell'JO!PU!

Sl'J SJ9A9f UO!ll'l!Pl'lJ l'J9Jl'J U! 9S!J G3NNV1dN:fl

i5U!A\OIIOJ e4uo :A.NV ,(q pell'lO!PU!

stJ 100d 1eRJ 1ueds e4l LI! doJp 1eAet Je1e;A G3:N:N:V1d:Nfl "t) Ee) Et) lU9Aff jl3RSRUf1 JO UO!lBO!J!lON_

'}.)';I znv ad z l oz JaqwaAON:

(~ IW!S!Aa'd)

LO 66 13:!"1:

Develof)eF Notes: NEI 99 Ql (Re\'isioR e) "!>Jove me er 2Q 12 For EAL #1 Site specific indications may include instrumentation values such as water level and area radiation monitor readings , and personnel reports. If available , video cameras may allow for remote observation.

Depending on available in strumentation , the declaration may also be based on indications of water makeup rate and/or decreases in the lev el of a water storage taflb For EAL #2 The specified value of 25 mR/hr may be set to another value for a specific application with appropriate justification.

EGL Assignment Attributes:

3.1.1.B B-2

~/El 99 QI (ReYisioR

6) November 2012 ECL: Notification of Unusua:I;, E:*,e=n~t

~PRO SU 1 Initiating Cenditien:

UNPLA1'H-ffi9 spent fuel pool temperature rise OpeFating Mede A I' . .

  • af)f) iealllh*"*

Not A 1* hi ~J

  • 1tpp-1cat1-e Example EmeFgeney Aetien Levels: (1) UNPLANNE9 spent fuel I pee temperature rise to greater than (s1*te s *.c: o peClt1C Fr-This IC addresses a condition that. potential degradation in the level o~ s:f:te~ursor to a more serious event and represents
a. occur , and result in a loss of pool level a:d ~fthe plant. T~u.ncorrected , soiling in the pool will . mcreased rad1at1on levels. Escalation of the emer e . g nc~* classification level *would ee via IC pg A A 1 >> .. eloper Notes: ' ** PD AA2. Th
  • e site specific temperature should e EGL A
  • n.ss1gnment Attrieutes:

3 .1.1.A B-3 ECL: Notification of Unusual Event Initiating Cenditien:

Confirmed SECURITY CONDITION or threat. Operating Mede ,\pplieability:

Not Applicable Example Emergeney Aetien Le~1 els: (I or 2 or 3) NEI 99 QI (ReYisieA

6) l>le¥ember 2Ql 2 PD HU1 (1) A SECURITY CONDITiffi,1 that does not involi,re a HOSTILE ACTION as reported by the (site specific security shift superi,rision).

(2) (3) }ofotification of a credible security threat directed at the site. A validated notification from the NRG providing information of an aircraft threat. Thi_s I~ addre~ses events that pose a threat to plant personnel or the equipment necessary to mamtam cool_mg of spent ~el , and thus represent a potential degradation in the level of plant safet_y. Security events v,rh1ch 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 ACTiffi,lS are classifiable under IC PD HAl. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security related event. Classification of these events will initiate appropriate threat related notifications to plant personnel and OROs. Security plans and terminology are based on the guidance provided by NEI 03 12 , Temptatcfor the Security Ptan , Treining e,9d QHelificetion Plen , Se.fcgMerds Contingency Plen {end Independent Spent-Fuel St-orege Instell8tio1~

Security Progrmn}.

EAL #1 references (site specific security shift supervision) because these are the individuals trained to confirm that a security event is occurring or has occurred.

Training on security event confirmation and classification is controlled due to the nature of Safeguards and 1 O CFR § 2.39 information. EAL #2 addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with (site specific procedure).

EAL#~ addresses the threat from the impact of an aircraft on the plant. The NRG 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 NRG. Validation of the threat is performed in accordance with (site specific procedure). Emergency plans and implementing procedures are public documents; therefore , EALs should not incorporate gecurity sensitive information.

This includes information that may be

  • B-4 NEI 99 QI (ReYisieA
6) ~Jeyemeer 2Q 12 advantageous to a potential adversary , such as the particulars concerning a specific threat or threat location.

Security sensitive information should be contained in non public documents such as the Security Plan. Escalation of the emergency classification le*,el would be via JC PD HAI. DevelepeF Netes: The (site specific security shift supervision) is the title of the on shift individual responsible for supervision of the on shift security force. The (site specific procedure) is the procedure(s) used by Control Room and/or Security personnel to determine if a security threat is credible , and to validate receipt of aircraft threat information.

Emergency plans and implementing procedures are public documents; therefore , EALs should not incorporate Security sensitive information.

This includes information that may be advantageous to a potential adversary , such as the particulars concerning a specific threat or threat location.

Security sensitive information should be contained in non public documents such as the Security Plan. With due consideration given to the above developer note , EALs may contain alpha or numbered references to selected events described in the Security Plan and associated implementing procedures.

Such references should not contain a recognizable description of the event. For e>rnmple , an EAL may be 't't'Orded as " Security event #2 , #5 or #9 is reported by the (site specific security shift superYision)." EGL Assignment Attributes:

3. I. I .A B-5 Nel 99 o I (ReYisioH G) No,*ember 2012 PD HU2 ECL: Notification of Unusual Event Initiating Cenditien:

Hazardous event affecting SAFHTY SYSTEM equipment necessary for spent fuel cooling. Operating Mede Applieability:

Not Appl i cable Example Emergeney A,etien Le,;els: (]) a. b. C. The occurrence of i+ ...... ~Y of the following hazardous events:

  • Internal or external flooding event
  • FIRE
  • EXPLOSION
  • (site specific hazards)
  • Other events vrith similar hazard characteristics as determined by the Shift Manager The event has damaged at least one train of a SAFETY SYSTEM needed for spent fuel cooling. AND The damaged SAFETY SYSTEM train(s) cannot , or potentially cannot , perform its design function based on EITHER:
  • Indications of degraded performance
  • VISIBLE; DAM.AGE This IC addresses a hazardous event that causes damage to at least one tra i n of a SAFHTY SYSTEM needed for spent fuel cooling. The damage must be of sufficient magnitude that the system(s) train cannot , or potentially cannot , perform its design function. This cond i tion reduces the margin to a loss or potential loss of the fuel clad barrier , and therefore represents a potentia l degradation of the level of safety of the plant. For HAL l.c , the first bullet addresses damage to a SAFHTY SYSTHM train that is in service/operation since indications for it will be readily available.

For EAL l.c , the second bullet addresses damage to a SAFETY SYSTHM train that is not in service/operation or readily apparent through indications a l one. Operators will make this B-6 NE[ 99 0 I (Re*,*isieA

6) Jl,fo*,remeer 2012 determination based on the total it)' of available event and damage report information.

This is intended to be a brief assessment not requiring lengthy anal)'Sis or quantification of the damage. Escalation of the emergency classification level could , depending upon the event , be based on any ofthe Alert ICs; PD AAl , PD AA2 , PD HAl orPD HA3. Devel01Jer Netes: 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 8Y8TEM8 are comprised oftvro or more separate and redundant trains of equipment in accordance with site specific design criteria.

EGL Assignment Attributes:

3 .1.1.A and 3 .1.1 C B-7 NEI 99 0 I (RevisioA

6) November 2012 PD HU3 ECL: Notification of Unusual Event Initiating Condition:

Other conditions e>(ist which in the judgment of the Emergency Director warrant declaration of a (NO)UE. Operating Mode Applieability:

Not Applicable Example Emergeney Aetion Levels: (1) Other conditions exist v,hich 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 e>(pected unless further degradation of safety systems occurs. This IC addresses unanticipated conditions not addressed e>(plicitly 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. B-8 ECL: Alert Nel 99 GI (RevisioR

6) J>Jo,*ember 2G 12 PD AA1 Initiating Condition:

Release of gaseous or liquid radioactivit)*

resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE. Operating Mode Applieability:

Not Applicable Example Emergeney Aetion Le*;els: (1 or 2 or 3 or 4) Notes:

  • The ?merg~ncy Director should declare the Alert promptly upon determining that the applicable time has been e>(ceeded , or *Nill likely be exceeded.
  • If an ongoin~ release is detected and the release start time is unknown , assume that the release duration has exceeded 15 minutes. * ~f the effluent flovr past an effluent monitor is known to haYe stopped due to actions to isolate the release path , then the effluent monitor reading is no longer *ralid for classification purposes.
  • The pre calculated effluent monitor Yalues presented in EAL #l should be used for emergency classific~ion assessments until the results from a dose assessment us i ng actual meteorology are aYa1Jable.

(1) (3) (4) Read~ng on ANY of the following radiation monitors greater than the reading shown for 15 mmutes or longer: (site specific monitor list and threshold values) (2) Dose assessment

~sing actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond (site specific dose receptor point). " I . f 1*

  • n:na y~1s o a1qu1d effluent sample indicates a concentration or release rate that would resul_t m doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond (site specific dose receptor po int) for one hour of e>(posure.

Field surve~ results indicate EITHER of the follovring at or beyond (site specific dose receptor point):

  • Closed windov, dose rates greater than 10 mR/hr e>(pected to continue for 60 minutes or longer.
  • Analyse_s of fie.Id survey samples indicate thyroid CDE greater than 50 mrem for one hour of mhalat1on.

Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual B-9 NEI 99 QI (Re¥isioR e) November 2Q 12 ef fsite deses greater thaR er equal te 1% efthe BPA Preteetive Aet~eR Gui~esd(PAGs).

1! eludes beth meRitered aRd uR meRitered releases. Releases ef this magmtu _e r~preseR , aR iR ae tual er peteRtial substaRtial degradatieR efthe level ef s~fi:ty efthe pl~Rt_as md1::!eedR~?o1~ed dielegieal release that sigRifieaRtly e , weeds regulatery l1m1ts (e.g., a s1gmfieaRt ra re lease). R adiele ieal efflueRt EALs are alse iReluded te previde a basis for ela_ssifyiRg eveRts _a~d aRElitia~s that eaRRat be "'&<lily ar app_ropriate~y

  • ~*s:ifiatpressed iR terms ef the sum efthe effe~tive des~ equivaleRt (EDE) eemmitted effeetive dese equivaleRt (CEDE), er as the thyre1d eemm1tted

~ese ::i~a~eRt (CDH). Far the perpase afthese ICIBALs , the Elase ~~aRtity ta~~~:~~*~'k

.... ". equivaleRt (TEDE), as defiRed iR 10 CFR § 20 , 1s used m lieu ef ... sum e The EPA PAG guidaRee prevides fer the use adul_t thyreid ~ese eew~ersieR faeters; he"'e"er seme states haYe deeided te base preteetive aetwRs eR eh_Ild thyre1d CD~. Nuelear , ., v 1' t IC LE, A Ls Reed te be eeRsisteRt with the preteetive aet1eR methedeleg1es emple) ed pewer p-aR s , n

  • h IC d EAL sheuld be b , the States withiR their EPZs. The thyreid COE des~ used i_R _t e a~, s. a~justed as Reeessary te aligR *Nith State preteetive aetwR dee1swR makmg entena. The " site speeifie meRiter list aRd thresheld values" sheuld be determiRed with eeRsideratieR ef the follewiRg:
  • SeleetieR ef the apprepriate iRstalled gaseeus aRd liquid effl~et~ meRitc;~DE er 50 mrem
  • The efflueRt meRiter r e adiRgs sheuld eerrespeRd te a dese mr~m . thyreid COE at the " site speeifie dese reeepter peiRt" (eeRs1steRt with the ealeulat1eR methedelegy empleyed) fer eRe heur ef expesure. .
  • MeRiter readiRgs will be ealeulated usiRg a set ef assumed meteereleg1eal data er ~tmespherie dispersieR faeters; the data er faeters seleeted for use sheuld be the same as these empleyed te ealeulate the meRiter readiRgs fer TC PD AUl
  • B-10 Nm 99 01 (Re\*isioA e) Noyember 2012
  • The calculation of monitor readings will also require use of an assumed release isotopic mi>E:; the selected mi>E: should be the same as that employed to calculate monitor readings for TC PD AUi.
  • Depending upon the methodology used to calculate the EAL values , there may be overlap of some values between different ICs. Developers will need to address this overlap by adjusting these values in a manner that ensures a logical escalation in the EGL. The " site specific dose receptor point" is the distance(s) and/or locations used by the licensee to distinguish bet>.veen on site and offsite doses. The selected distance(s) and/or locations should reflect the content of the emergency plan , and the procedural methodology used to determine offsite doses and Protecfr, , e Action Recommendations. The variation in selected dose receptor points means there may be some differences in the distance from the release point to the calculated dose point from site to site. Developers should research radiation monitor design documents or other information sources to ensure that 1) the EAL value being considered is within the usable response and display range of the instrument , and 2) there are no automatic features that may render the monitor reading invalid (e.g., an auto purge feature triggered at a particular indication level). It is recognized that the condition described by this IC may result in a radiological effluent value beyond the operating or display range of the installed effluent monitor. In those cases , EAL values should be determined 1,vith a margin sufficient to ensure that an accurate monitor reading is available.

For e>rnmple , an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision nohvithstanding, if the estimated/calculated monitor reading is greater than approximately 110% of the highest accurate monitor reading , then developers may choose not to include the monitor as an indication and identify an alternate EAL threshold.

Although the IC references TEDE , field survey results are generally available only as a " whole body" dose rate. For this reason , the field sur*,ey EAL specifies a " closed window" survey reading. Indications from a real time dose projection system are not included in the generic EALs. Many licensees do not have this capability.

For those that do , the capability may not be within the scope of the plant Technical Specifications.

A licensee may request to include an EAL using real time dose projection system results; approval will be considered on a case by case basis. Indications from a perimeter monitoring system are not included in the generic EA.Ls. Many licensees do not have this capability.

For those that do, these monitors may not be controlled and maintained to the same level as plant equipment , or 1 ,vithin the scope of the plant Technical Specifications.

In addition , readings may be influenced by environmental or other factors. A licensee may request to include an EAL using a perimeter monitoring S)'Stem; approval 1 Nill be considered on a case by case basis. EGL Assignment Attributes:

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ECL: Alert Ne! 99 0 I (RevisioA e) ~lovemeer 2012 PD HA1 Initiating Condition:

HOSTILE i\CTION v,'ithin the OWNER CONTROLLED AREA or airborne attack threat 1.vithin 30 minutes. Operating Mode Applieability:

Not Applicable Example Emergeney Aetion Levels: (1 or 2) (1) (2) A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the (site specific security shift supervision).

/*, validated notification from NRG of an aircraft attack threat 1 Nithin 30 minutes of the This IC addresses the occurrence of a HOSTILE ACTION within the O\l/NER CONTROLLED AREA or notification of an aircraft attack threat. This event v,ill 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 bet\Neen 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 NET 03 12 , Templ8tc fer the Security Plan , Trainil'lg end Qualificetion Plan , Sejegbu1rds Co1'1tingenc;*

Plen {end Independent Spent Fuel Storege Instelletion Security Progreni}. As time and conditions allow , these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation , dispersal or sheltering).

The Alert declaration will also heighten the awareness of Offsite Response Organizations , 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. E>rnmples include the crash of a small airoraft , shots from hunters , physical disputes between employees , etc. Reporting of these types of events is adequately addressed by other EA.Ls , or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. EAL #1 is applicable for any HOSTILE ACTION occurring , or that has occurred , in the O\V1'mR CONTROLLED AREA. This includes any action directed against an TSfST that is located within the OW1'mR CONTROLLED AREA. EAL #2 addresses the threat from the impact of an airoraft on the plant , and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat related B-13 NEI 99 0 I (RevisioA

6) November 2012 notifications.

are made in a timely manner so that plant personnel and OROs are in a heightened state of read mess. This EAL is met when the threat related information has been validated in accordance with (site specific procedure). !he NRG He~dquarters Operations Officer (HOO) 'Nill communicate to the licensee if the threat =*es an aircraft.

The status and size of the plane may be provided by NOR/\.D through the In some cases , it may not be readily apparent if an aircraft impact 1.vithin the O\l/NER CONTR?LLED A~A ~vas intentional (i._e., a HOSTILE ACTION). It is e),pected , although not c_ertam , that not1fica~1on by an appropnate Federal agency to the site would clarify this point. In this case , the appropnate federal agency is intended to be NORAD , FBI , FAA or }iRC. The em~rgenc):

?eclar~ion , _including one based on other ICs/EALs , should not be unduly delayed while awaitmg not1ficat10n by a Federal agency. Em:rgency plans and implementing procedures are public documents; therefore , EALs should no: , mcorporate Security ~ensitive information.

This includes information that may be ad" antageo~s to a pot:nt1al a~v.ers~ry , such as the particulars concerning a specific threat or threat location. Secunty sens1t1ve mformation should be contained in non public documents such as the Security Plan. Denio per Notes: The (si_t~ specific securi~y shift supervision) is the title of the on shift individual responsible for superv1s1on of the on shift security force. Em:rgenC)'

plans and implementing procedures are public documents; therefore , EALs sh~, uld not mcorporate S:curity sensitive information. This includes information that may be ad rantageo~s to a pot:ntial a~v.ers~ry , such_as the particulars concerning a specific threat or threat location.

Security sens1t1ve mformat1on should be contained in non public documents such as the Securit)' Plan. urth d *d * * ,r1ue cons1 erat1on given to the above developer note , EALs may contain alpha or ~umbered ~eferences to selected events described in the Security Plan and associated

-1mplementmg procedures.

Such references should not contain a recognizable description of the e','.ent.

Fo_r exampl~, an EAL may be worded as " Security event #2 , #5 or #9 is reported by the *(site specific secunty shift supervision)." See the related Developer Note in Appendi>, B , Definitions , for guidance on the development of ,a scheme definition for the OWNER CONTROLLED AR~A. EGL Assignment Attributes:

3.1 .2.D B-14 NE! 99 Ql (ReYisioA 1:i) NO'>'emeer 2Q 12 PD HA3 ECL: Alert Initiating Canditian:

Other conditions e>(ist 1tvhich in the judgment of the Emergency Director warrant declaration of an Alert. Operating Made Applieability:

Not Applicable Example Emergeney Aetian L~1 els: (1) Other conditions e>(ist v,hich in the judgment of the Emergenc;'

Director indicate that events are in progress or have occurred *which invohe 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. An;' releases are expected to be limited to small fractions of the EPA ProtectiYe Action Guideline exposure leYels. This IC addresses unanticipated conditions not addressed explicitly else'tvhere but that 1 Narrant declaration of an emergency because conditions e>(ist which are belie't 1 ed by the Emergency Director to fall under the emergency classification level description for an Alert. B-15

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