NEI 99-01 [Revision 6]

Development of

Emergency Action Levels

for Non-Passive Reactors

November 2012
[THIS PAGE IS LEFT BLANK INTENTIONALLY]
Nuclear Energy Institute, 1776 I Street N. W., Suite 400, Washington D.C. (202.739.8000)

NEI 99-01 [Revision 6]

Nuclear Energy Institute

Development of

Emergency Action Levels

for Non-Passive Reactors

November 2012
ACKNOWLEDGMENTS

This document was prepared by the Nuclear Energy Institute (NEI) Emergency Action Level 

(EAL) Task Force. 

NEI Chairperson: David Young 

Preparation Team

Larry Baker – Exelon Nuclear/Corporate 

Craig Banner – PSEG Nuclear/Salem and Hope Creek Nuclear Generating Stations/USA

John Egdorf – Dominion Generation/Kewaunee Power Station

Jack Lewis – Entergy Nuclear/Corporate

C. Kelly Walker – Operations Support Services, Inc. 

 

Review Team

Chris Boone – Southern Nuclear/Corporate

John Callahan – Xcel Energy/Corporate/USA

Bill Chausse – Enercon Services, Inc.

Kent Crocker – Progress Energy/Brunswick Nuclear Plant 

Don Crowl – Duke Energy/Corporate 

Roger Freeman – Constellation Energy Nuclear Group/Corporate

Walt Lee – TVA Nuclear/Corporate

Ken Meade – FENOC/Corporate 

Don Mothena – NextEra Energy/Corporate

David Stobaugh – EP Consulting, LLC 

Nick Turner – Callaway Plant/STARS

Maureen Zawalick – Diablo Canyon Power Plant/STARS

NOTICE

Neither NEI, nor any of its employees, members, supporting organizations, contractors, or 

consultants make any warranty, expressed or implied, or assume any legal responsibility for the 

accuracy or completeness of, or assume any liability for damages resulting from any use of, any 

information apparatus, methods, or process disclosed in this report or that such may not 

infringe privately owned rights. 
NEI 99-01 (Revision 6)

November 2012

i 

EXECUTIVE SUMMARY

Federal regulations require that a nuclear power plant operator develop a scheme for the 

classification of emergency events and conditions. This scheme is a fundamental component of 

an emergency plan in that it provides the defined thresholds that will allow site personnel to 

rapidly implement a range of pre-planned emergency response measures. An emergency 

classification scheme also facilitates timely decision-making by an Offsite Response 

Organization (ORO) concerning the implementation of precautionary or protective actions for 

the public. 

The purpose of Nuclear Energy Institute (NEI) 99-01 is to provide guidance to nuclear power 

plant operators for the development of a site-specific emergency classification scheme. The 

methodology described in this document is consistent with Federal regulations, and related US 

Nuclear Regulatory Commission (NRC) requirements and guidance. In particular, this 

methodology has been endorsed by the NRC as an acceptable approach to meeting the 

requirements of 10 CFR § 50.47(b)(4), related sections of 10 CFR § 50, Appendix E, and the 

associated planning standard evaluation elements of NUREG-0654/ FEMA-REP-1, Rev. 1, 

Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and 

Preparedness in Support of Nuclear Power Plants, November 1980. 

NEI 99-01 contains a set of generic Initiating Conditions (ICs), Emergency Action Levels 

(EALs) and fission product barrier status thresholds. It also includes supporting technical basis 

information, developer notes and recommended classification instructions for users. Users 

should implement ICs, EALs and thresholds that are as close as possible to the generic material 

presented in this document with allowance for changes necessary to address site-specific 

considerations such as plant design, location, terminology, etc.

Properly implemented, the guidance in NEI 99-01 will yield a site-specific emergency 

classification scheme with clearly defined and readily observable EALs and thresholds. Other 

benefits include the development of a sound basis document, the adoption of industry-standard 

instructions for emergency classification (e.g., transient events, classification of multiple events, 

upgrading, downgrading, etc.), and incorporation of features to improve human performance. 

An emergency classification using this scheme will be appropriate to the risk posed to plant 

workers and the public, and should be the same as that made by another NEI 99-01 user plant in 

response to a similar event.

The individuals responsible for developing an emergency classification scheme are strongly 

encouraged to review all applicable NRC requirements and guidance prior to beginning their 

efforts. Questions concerning this document may be directed to the NEI Emergency 

Preparedness staff, NEI EAL task force members or submitted to the Emergency Preparedness 

Frequently Asked Questions process. 

Finally, unique State and local requirements associated with an emergency classification scheme 

are not reflected in this guidance. Incorporation of these requirements may be performed on a 

case-by-case basis in conjunction with the appropriate ORO agency. Any such changes will 

require a review under the applicable sections of 10 CFR 50. 
NEI 99-01 (Revision 6)

November 2012

ii

[THIS PAGE IS LEFT BLANK INTENTIONALLY]
NEI 99-01 (Revision 6)

November 2012

iii

TABLE OF CONTENTS

EXECUTIVE SUMMARY ........................................................................................................ i 

1 REGULATORY BACKGROUND .....................................................................................1

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

1.2 PERMANENTLY DEFUELED STATION .............................................................................1

1.3 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI).....................................2

1.4 NRC ORDER EA-12-051 ................................................................................................2

1.5 APPLICABILITY TO ADVANCED AND SMALL MODULAR REACTOR DESIGNS ...............4

2 KEY TERMINOLOGY USED IN NEI 99-01 ....................................................................5

2.1 EMERGENCY CLASSIFICATION LEVEL (ECL) ...............................................................5

2.2 INITIATING CONDITION (IC) ..........................................................................................7

2.3 EMERGENCY ACTION LEVEL (EAL) .............................................................................7

2.4 FISSION PRODUCT BARRIER THRESHOLD .....................................................................7

3 DESIGN OF THE NEI 99-01 EMERGENCY CLASSIFICATION SCHEME.......................9

3.1 ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS (ECLS) ...............................9

3.2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS....................12

3.3 NSSS DESIGN DIFFERENCES ........................................................................................12

3.4 ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION .............................13

3.5 IC AND EAL MODE APPLICABILITY ............................................................................14

4 SITE-SPECIFIC SCHEME DEVELOPMENT GUIDANCE ..............................................16

4.1 GENERAL IMPLEMENTATION GUIDANCE ....................................................................16

4.2 CRITICAL CHARACTERISTICS ......................................................................................17

4.3 INSTRUMENTATION USED FOR EALS ..........................................................................18

4.4 PRESENTATION OF SCHEME INFORMATION TO USERS ...............................................18

4.5 INTEGRATION OF ICS/EALS WITH PLANT PROCEDURES ...........................................19

4.6 BASIS DOCUMENT .........................................................................................................20

4.7 DEVELOPER AND USER FEEDBACK ..............................................................................21

5 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS ......................................22

5.1 GENERAL CONSIDERATIONS ........................................................................................22

5.2 CLASSIFICATION METHODOLOGY ...............................................................................23

5.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS ........................................23

5.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION ..............................23
NEI 99-01 (Revision 6)

November 2012

iv

5.5 CLASSIFICATION OF IMMINENT CONDITIONS .............................................................24

5.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING .................24

5.7 CLASSIFICATION OF SHORT-LIVED EVENTS ...............................................................25

5.8 CLASSIFICATION OF TRANSIENT CONDITIONS ............................................................25

5.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION ..............26

5.10 RETRACTION OF AN EMERGENCY DECLARATION .......................................................26

6 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT ICS/EALS ........................27

7 COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS ....................49

8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS..............76

9 FISSION PRODUCT BARRIER ICS/EALS ..................................................................78

10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS.......123

11 SYSTEM MALFUNCTION ICS/EALS.........................................................................148

APPENDIX A – ACRONYMS AND ABBREVIATIONS........................................................ A-1

APPENDIX B – DEFINITIONS ........................................................................................... B-1

APPENDIX C – PERMANENTLY DEFUELED STATION ICs/EALs .................................... C-1
NEI 99-01 (Revision 6)

November 2012

1 

DEVELOMENT OF EMERGENCY ACTION LEVELS

FOR NON-PASSIVE REACTORS

1 REGULATORY BACKGROUND 

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)(1)(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

The above list is not all-inclusive and it is strongly recommended that scheme developers 

consult with licensing/regulatory compliance personnel to identify and understand all 

applicable requirements and guidance. Questions may also be directed to the NEI 

Emergency Preparedness staff.

1.2 PERMANENTLY DEFUELED STATION

NEI 99-01 provides guidance for an emergency classification scheme applicable to a 

permanently defueled station. This is a station that generated spent fuel under a 10 CFR 

§ 50 license, has permanently ceased operations and will store the spent fuel onsite for an 

extended period of time. The emergency classification levels applicable to this type of 

station are consistent with the requirements of 10 CFR § 50 and the guidance in NUREG 

0654/FEMA-REP-1. 

In order to relax the emergency plan requirements applicable to an operating station, the 

owner of a permanently defueled station must demonstrate that no credible event can 

result in a significant radiological release beyond the site boundary. It is expected that 
NEI 99-01 (Revision 6)

November 2012

2 

this verification will confirm that the source term and motive force available in the 

permanently defueled condition are insufficient to warrant classifications of a Site Area 

Emergency or General Emergency. Therefore, the generic Initiating Conditions (ICs) 

and Emergency Action Levels (EALs) applicable to a permanently defueled station may 

result in either a Notification of Unusual Event (NOUE) or an Alert classification. 

The generic ICs and EALs are presented in Appendix C, Permanently Defueled Station 

ICs/EALs. 

1.3 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-HU1 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 HU1 and 

IC HA1 should also be included to address a HOSTILE ACTION directed against an 

ISFSI. 

The analysis of potential onsite and offsite 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 if requested). Also, the licensee’s Emergency Response Organization (ERO) 

required for 10 CFR § 72.32 emergency plan is different than that prescribed for a 10 

CFR § 50.47 emergency plan (e.g., no emergency technical support function). 

1.4 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 
NEI 99-01 (Revision 6)

November 2012

3 

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 existing IC AA2, and new ICs AS2 and AG2. Associated EAL 

notes, bases and developer notes are also provided. 

It is recommended that these EALs be implemented when the enhanced spent fuel pool 

level instrumentation is available for use.

The regulatory process that licensees follow to make changes to their emergency plan, 

including non-scheme changes to EALs, is 10 CFR 50.54(q). In accordance with this 

regulation, licensees are responsible for evaluating a proposed change and determining 

whether or not it results in a reduction in the effectiveness of the plan. As a result of the 

licensee's determination, the licensee will either make the change or submit it to the NRC 

for prior review and approval in accordance with 10 CFR 50.90. 
NEI 99-01 (Revision 6)

November 2012

4 

1.5 APPLICABILITY TO ADVANCED AND SMALL MODULAR REACTOR DESIGNS

The guidance in this document primarily addresses commercial nuclear power reactors in 

the United States, operating or permanently defueled, as of 2012 (so called 1st and 2nd

generation plant designs); however, it may be adapted to advanced non-passive designs 

(often referred to as 3rd generation plant designs) as well. Developers of an emergency 

classification scheme for an advanced non-passive reactor plant may need to propose 

deviations from the generic guidance to account for the differences in design parameters 

and criteria, and operating characteristics and capabilities, between 2nd and 3rd generation 

plants. 

There are significant design and operating differences between large commercial nuclear 

power plants (of any generation) and Small Modular Reactors (SMRs) (e.g., differences 

in source term). For this reason, this document is not applicable to SMRs. 
NEI 99-01 (Revision 6)

November 2012

5 

2 KEY TERMINOLOGY USED IN NEI 99-01 

There are several key terms that appear throughout the NEI 99-01 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 Alert SAE GE

   

Initiating Condition Initiating Condition Initiating Condition Initiating Condition

   

Emergency Action 

Level (1)

• Operating Mode 

Applicability

• Notes

• Basis

Emergency Action 

Level (1)

• Operating Mode 

Applicability

• Notes

• Basis

Emergency Action 

Level (1)

• Operating Mode 

Applicability

• Notes

• Basis

Emergency Action 

Level (1)

• Operating Mode 

Applicability

• Notes

• 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)1

Events are in progress or have occurred which indicate a potential degradation of the 

level of safety of the plant or indicate a security threat to facility protection has been 

initiated. No releases of radioactive material requiring offsite response or monitoring are 

expected unless further degradation of safety systems occurs.

 

1 This term is sometimes shortened to Unusual Event (UE) or other similar site-specific terminology. The terms 

Notification of Unusual Event, NOUE and Unusual Event are used interchangeably throughout this document
NEI 99-01 (Revision 6)

November 2012

6 

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. 

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 offsite 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 PAG exposure levels offsite for more than the 

immediate site area.

Purpose: The purpose of the General Emergency declaration is to initiate predetermined 

protective actions for the public, to provide continuous assessment of information from 

the licensee and offsite organizational measurements, to initiate additional measures as 

indicated by actual or potential releases, to provide consultation with offsite authorities, 

and to provide updates for the public through government authorities. 
NEI 99-01 (Revision 6)

November 2012

7 

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. 

Considerations for the assignment of a particular Initiating Condition to an emergency 

classification level are discussed in Section 3. 

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 
NEI 99-01 (Revision 6)

November 2012

8 

Levels/ Radiological Effluent (A) 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.). 
NEI 99-01 (Revision 6)

November 2012

9 

3 DESIGN OF THE NEI 99-01 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 NEI 99-01 emergency classification scheme attempts 

to strike an appropriate balance between reasonably anticipated event or condition 

consequences, potential accident trajectories, and risk avoidance or minimization. 

There are a range of “non-emergency events” reported to the US Nuclear Regulatory 

Commission (NRC) staff in accordance with the requirements of 10 CFR § 50.72. 

Guidance concerning these reporting requirements, and example events, are provided in 

NUREG-1022. Certain events reportable under the provisions of 10 CFR § 50.72 may

also require the declaration of an emergency. 

In order to align each Initiating Conditions (IC) with the appropriate ECL, it was 

necessary to determine the attributes of each ECL. The goal of this process is to answer 

the question, “What events or conditions should be placed under each ECL?” The 

following sources provided information and context for the development of ECL 

attributes. 

 Assessments of the effects and consequences of different types of events and

conditions 

 Typical abnormal and emergency operating procedure setpoints and transition criteria

 Typical Technical Specification limits and controls 

 Radiological Effluent Technical Specifications (RETS)/Offsite Dose Calculation 

Manual (ODCM) 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 industry subject matter experts and NRC staff members

The following ECL attributes were created by the Revision 6 Preparation Team to aid in

the development of ICs and Emergency Action Levels (EALs). The team decided to 

include the attributes in this revision since they may be useful in briefing and training 

settings (e.g., helping an Emergency Director understand why a particular condition is 

classified as an Alert). It should be stressed that developers not attempt to redefine these 

attributes or apply them in any fashion that would change the generic guidance contained 

in this document2

. 

 

2 The use of ECL attributes is at the discretion of a licensee and is not a requirement of the NRC. If a licensee 

chooses in incorporate the ECL attributes into their scheme basis document, it must be very clear that the NRC staff 

has not endorsed their acceptability or application for any purpose. In particular, the staff does not consider the 

attribute statements to supersede the established ECL definitions. As a result, the use of the attributes as a basis for 

justifying EAL changes is unacceptable. 
NEI 99-01 (Revision 6)

November 2012

10 

The attributes of each ECL are presented below.

3.1.1 Notification of Unusual Event (NOUE)

A Notification of Unusual Event, as defined in section 2.1.1, includes but is not limited to 

an event or condition that involves: 

(A)A precursor to a more significant event or condition. 

(B) A minor loss of control of radioactive materials or the ability to control radiation 

levels within the plant.

(C) A consequence otherwise significant enough to warrant notification to local, State and 

Federal authorities. 

3.1.2 Alert

An Alert, as defined in section 2.1.2, includes but is not limited to an event or condition 

that involves: 

(A)A loss or potential loss of either the fuel clad or Reactor Coolant System (RCS) 

fission product barrier. 

(B) An event or condition that significantly reduces the margin to a loss or potential loss 

of the fuel clad or RCS fission product barrier. 

(C) A significant loss of control of radioactive materials resulting in an inability to 

control radiation levels within the plant, or a release of radioactive materials to the 

environment that could result in doses greater than 1% of an EPA 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

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 PAG at or beyond the site boundary. 
NEI 99-01 (Revision 6)

November 2012

11 

(D)A HOSTILE ACTION occurring within the plant PROTECTED AREA. 

3.1.4 General Emergency

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 PAG 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 of risk-based 

assessment results. To that end, the development and assignment of certain ICs and 

EALs also considered insights from several site-specific probabilistic safety assessments 

(PSA - also known as probabilistic risk assessment, PRA). 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 Pressurized Water Reactors (PWRs) 

and 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.
NEI 99-01 (Revision 6)

November 2012

12 

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 site-specific 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.

3.2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS

The NEI 99-01 methodology makes use of symptom-based, barrier-based and eventbased 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 offnormal, 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 barrierbased 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 ICs and EALs define a variety of specific occurrences that have potential or 

actual safety significance. These include the failure of an automatic reactor scram/trip to 

shut down the reactor, natural phenomena (e.g., an earthquake), or man-made hazards 

such as a toxic gas release.

3.3 NSSS DESIGN DIFFERENCES

The NEI 99-01 emergency classification scheme accounts for the design differences 

between PWRs and BWRs by specifying EALs unique to each type of Nuclear Steam 

Supply System (NSSS). There are also significant design differences among PWR 

NSSSs; therefore, guidance is provided to aid in the development of EALs appropriate to 

different PWR NSSS types. Where necessary, development guidance also addresses 

unique considerations for advanced non-passive reactor designs such as the Advanced 

Boiling Water Reactor (ABWR), the Advanced Pressurized Water Reactor (APWR) and 

the Evolutionary Power Reactor (EPR).

Developers will need to consider the relevant aspects of their plant’s design and operating 

characteristics when converting the generic guidance of this document into a site-specific 

classification scheme. The goal is to maintain as much fidelity as possible to the intent of 

generic ICs and EALs within the constraints imposed by the plant design and operating 

characteristics. To this end, developers of a scheme for an advanced non-passive reactor 

may need to add, modify or delete some information contained in this document; these 

changes will be reviewed for acceptability by the NRC as part of the scheme approval 

process. 
NEI 99-01 (Revision 6)

November 2012

13 

The guidance in NEI 99-01 is not applicable to advanced passive light water reactor

designs. An Emergency Classification Scheme for this type of plant should be developed 

in accordance with NEI 07-01, Methodology for Development of Emergency Action 

Levels, Advanced Passive Light Water Reactors. 

3.4 ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION

The scheme’s generic information is organized by Recognition Category in the following 

order.

 A - Abnormal Radiation Levels / Radiological Effluent – Section 6

 C - Cold Shutdown / Refueling System Malfunction – Section 7

 E - Independent Spent Fuel Storage Installation (ISFSI) – Section 8 

 F - Fission Product Barrier – Section 9

 H - Hazards and Other Conditions Affecting Plant Safety – Section 10 

 S - System Malfunction – Section 11 

 PD - Permanently Defueled Station – Appendix C 

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). 

 Example Emergency Action Level(s) – Provides examples of reports and 

indications that are considered to meet the intent of the IC. Developers should 

address each example EAL. If the generic approach to the development of an 

example EAL cannot be used (e.g., an assumed instrumentation range is not available 

at the plant), the developer should attempt to specify an alternate means for 

identifying entry into the IC. 

For Recognition Category F, the fission product barrier thresholds are presented in 

tables applicable to BWRs and PWRs, 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.

 Developer Notes - Information that supports the development of the site-specific ICs 

and EALs. This may include clarifications, references, examples, instructions for 

calculations, etc. Developer notes should not be included in the site’s emergency 
NEI 99-01 (Revision 6)

November 2012

14 

classification scheme basis document. Developers may elect to include information 

resulting from a developer note action in a basis section.

 ECL Assignment Attributes – Located within the Developer Notes section, 

specifies the attribute used for assigning the IC to a given ECL. 

3.5 IC AND EAL MODE APPLICABILITY

The NEI 99-01 emergency classification scheme was developed recognizing that the 

applicability of ICs and EALs will vary with plant mode. For example, some symptombased 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 A C E F H PD S 

Power Operations X X X X X 

Startup X X X X X 

Hot Standby 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 

Permanently 

Defueled X X 
NEI 99-01 (Revision 6)

November 2012

15 

Typical BWR Operating Modes 

Power Operations (1): Mode Switch in Run 

Startup (2): Mode Switch in Startup/Hot Standby or Refuel 

(with all vessel head bolts fully tensioned) 

Hot Shutdown (3): Mode Switch in Shutdown, Average Reactor 

Coolant Temperature >200 °F 

Cold Shutdown (4): Mode Switch in Shutdown, Average Reactor 

Coolant Temperature ≤ 200 °F 

Refueling (5): Mode Switch in Shutdown or Refuel, and one or 

more vessel head bolts less than fully tensioned. 

Typical PWR Operating Modes 

Power Operations (1): Reactor Power > 5%, Keff ≥ 0.99 

Startup (2): Reactor Power ≤ 5%, Keff ≥ 0.99 

Hot Standby (3): RCS ≥ 350 °F, Keff < 0.99 

Hot Shutdown (4): 200 °F < RCS < 350 °F, Keff < 0.99

Cold Shutdown (5): RCS < 200 °F, Keff < 0.99 

Refueling (6): One or more vessel head closure bolts less than 

fully tensioned 

Developers will need to incorporate the mode criteria from unit-specific Technical 

Specifications into their emergency classification scheme. In addition, the scheme must 

also include the following mode designation specific to NEI 99-01: 

Defueled (None): All fuel removed from the reactor vessel (i.e., full 

core offload during refueling or extended outage).
NEI 99-01 (Revision 6)

November 2012

16 

4 SITE-SPECIFIC SCHEME DEVELOPMENT GUIDANCE 

This section provides detailed guidance for developing a site-specific emergency classification 

scheme. Conceptually, the approach discussed here mirrors the approach used to prepare 

emergency operating procedures – generic material prepared by reactor vendor owners groups is

converted by each nuclear power plant into site-specific emergency operating procedures. 

Likewise, the emergency classification scheme developer will use the generic guidance in NEI 

99-01 to prepare a site-specific emergency classification scheme and the associated basis 

document. 

It is important that the NEI 99-01 emergency classification scheme be implemented as an 

integrated package. Selected use of portions of this guidance is strongly discouraged as it will 

lead to an inconsistent or incomplete emergency classification scheme that will likely not receive 

the necessary regulatory approval. 

4.1 GENERAL IMPLEMENTATION GUIDANCE

The guidance in NEI 99-01 is not intended to be applied to plants “as-is”; however, 

developers should attempt to keep their site-specific schemes as close to the generic 

guidance as possible. The goal is to meet the intent of the generic Initiating Conditions 

(ICs) and Emergency Action Levels (EALs) within the context of site-specific 

characteristics – locale, plant design, operating features, terminology, etc. Meeting this 

goal will result in a shorter and less cumbersome NRC review and approval process, 

closer alignment with the schemes of other nuclear power plant sites and better 

positioning to adopt future industry-wide scheme enhancements.

When properly developed, the ICs and EALs should be unambiguous and readily 

assessable. 

As discussed in Section 3, the generic guidance includes ICs and example EALs. It is the 

intent of this guidance that both be included in site-specific documents as each serves a 

specific purpose. 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. If some 

feature of the plant location or design is not compatible with a generic IC or EAL, efforts 

should be made to identify an alternate IC or EAL. 

If an IC or EAL includes an explicit reference to a mode dependent technical 

specification limit that is not applicable to the plant, then that IC and/or EAL need not be 

included in the site-specific scheme. In these cases, developers must provide adequate 

documentation to justify why the IC and/or EAL were not incorporated (i.e., sufficient 

detail to allow a third party to understand the decision not to incorporate the generic 

guidance). 

Useful acronyms and abbreviations associated with the NEI 99-01 emergency 

classification scheme are presented in Appendix A, Acronyms and Abbreviations. Sitespecific entries may be added if necessary.

Many words or terms used in the NEI 99-01 emergency classification scheme have 
NEI 99-01 (Revision 6)

November 2012

17 

scheme-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.

Below are examples of acceptable modifications to the generic guidance. These may be 

incorporated depending upon site developer and user preferences. 

 The ICs within a Recognition Category may be placed in reverse order for 

presentation purposes (e.g., start with a General Emergency at the left/top of a user 

aid, followed by Site Area Emergency, Alert and NOUE). 

 The Initiating Condition numbering may be changed. 

 The first letter of a Recognition Category designation may be changed, as follows, 

provided the change is carried through for all of the associated IC identifiers.

• R may be used in lieu of A 

• M may be used in lieu of S 

For example, the Abnormal Radiation Levels / Radiological Effluent category designator 

“A” (for Abnormal) may be changed to “R” (for Radiation). This means that the 

associated ICs would be changed to RU1, RU2, RA1, etc. 

 The ICs and EALs from Recognition Categories S and C may be incorporated into a 

common presentation method (e.g., one table) provided that all related notes and 

mode applicability requirements are maintained.

 The ICs and EALs for Emergency Director judgment and security-related events may 

be placed under separate Recognition Categories. 

 The terms EAL and threshold may be used interchangeably.

The material in the Developer Notes section is included to assist developers with crafting 

correct IC and EAL statements. This material is not required to be in the final emergency 

classification scheme basis document.

4.2 CRITICAL CHARACTERISTICS

As discussed above, developers are encouraged to keep their site-specific schemes as 

close to the generic guidance as possible. When crafting the scheme, developers should 

satisfy themselves 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, a sitespecific scheme must include some type of user-aid to facilitate timely and accurate 

classification of fission product barrier losses and/or potential losses. The user-aid 

logic must be consistent with the classification logic presented in Section 9. 

 The ICs, EALs, Operating Mode Applicability criteria, Notes and Basis information

are technically complete and accurate (i.e., they contain the information necessary to 

make a correct classification).

 EAL statements use objective criteria and observable values.
NEI 99-01 (Revision 6)

November 2012

18 

 ICs, EALs, Operating Mode Applicability and Note statements and formatting 

consider human factors and are user-friendly. 

 The scheme facilitates upgrading and downgrading of the emergency classification

where necessary.

 The scheme facilitates classification of multiple concurrent events or conditions. 

4.3 INSTRUMENTATION USED FOR EALS

Instrumentation referenced in EAL statements should include that described in the 

emergency plan section which addresses 10 CFR 50.47(b)(8) and (9) and/or Chapter 7 of 

the FSAR. Instrumentation used for EALs need not be safety-related, addressed by a 

Technical Specification or ODCM/RETS control requirement, nor powered from an 

emergency power source; however, EAL developers should strive to incorporate 

instrumentation that is reliable and routinely maintained in accordance with site programs 

and procedures. Alarms referenced in EAL statements should be those that are the most 

operationally significant for the described event or condition. 

Scheme developers should ensure that specified values used as 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 should 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. Findings 

and violations related to EAL instrumentation issues may be located on the NRC website.

4.4 PRESENTATION OF SCHEME INFORMATION TO USERS

The US Nuclear Regulatory Commission (NRC) expects licensees to establish and 

maintain the capability to assess, classify and declare an emergency condition promptly 

within 15 minutes after the availability of indications to plant operators that an 

emergency action level has been, or may be, exceeded. When writing an emergency 

classification procedure and creating related user aids, the developer must determine the 

presentation method(s) that best supports the end users by facilitating accurate and timely 

emergency classification. To this end, developers should consider the following points. 

 The first users of an emergency classification procedure are the operators in the 

Control Room. During the allowable classification time period, they may have 

responsibility to perform other critical tasks, and will likely have minimal assistance 

in making a classification assessment. 

 As an emergency situation evolves, members of the Control Room staff are likely to 

be the first personnel to notice a change in plant conditions. They can assess the 

changed conditions and, when warranted, recommend a different emergency 

classification level to the Technical Support Center (TSC) and/or Emergency 

Operations Facility (EOF). 

 Emergency Directors in the TSC and/or EOF will have more opportunity to focus on 

making an emergency classification, and will probably have advisors from Operations 

available to help them. 

Emergency classification scheme information for end users should be presented in a 

manner with which licensed operators are most comfortable. Developers will need to
NEI 99-01 (Revision 6)

November 2012

19 

work closely with representatives from the Operations and Operations Training 

Departments to develop readily usable and easily understood classification tools (e.g., a 

procedure and related user aids). If necessary, an alternate method for presenting 

emergency classification scheme information may be developed for use by Emergency 

Directors and/or Offsite Response Organization personnel. 

A wallboard is an acceptable presentation method provided that it contains all the 

information necessary to make a correct emergency classification. This information 

includes the ICs, Operating Mode Applicability criteria, EALs and Notes. Notes may be 

kept with each applicable EAL or moved to a common area and referenced; a reference to 

a Note is acceptable as long as the information is adequately captured on the wallboard 

and pointed to by each applicable EAL3

. Basis information need not be included on a 

wallboard but it should be readily available to emergency classification decision-makers.

In some cases, it may be advantageous to develop two wallboards - one for use during 

power operations, startup and hot conditions, and another for cold shutdown and 

refueling conditions.

Alternative presentation methods for the Recognition Category F ICs and fission product 

barrier thresholds are acceptable and include flow charts, block diagrams, and checklisttype tables. Developers must ensure that the site-specific method addresses all possible 

threshold combinations and classification outcomes shown in the BWR or PWR EAL 

fission product barrier tables. The NRC staff considers the presentation method of the 

Recognition Category F information to be an important user aid and may request a 

change to a particular proposed method if, among other reasons, the change is necessary 

to promote consistency across the industry. 

4.5 INTEGRATION OF ICS/EALS WITH PLANT PROCEDURES

A rigorous integration of IC and EAL references into plant operating procedures is not 

recommended. This approach would greatly increase the administrative controls and 

workload for maintaining procedures. On the other hand, performance challenges may 

occur if recognition of meeting an IC or EAL is based solely on the memory of a licensed 

operator or an Emergency Director, especially during periods of high stress. 

Developers should consider placing appropriate visual cues (e.g., a step, note, caution, 

etc.) in plant procedures alerting the reader/user to consult the site emergency 

classification procedure. Visual cues could be placed in emergency operating 

procedures, abnormal operating procedures, alarm response procedures, and normal 

operating procedures that apply to cold shutdown and refueling modes. As an example, a 

step, note or caution could be placed at the beginning of an RCS leak abnormal operating 

procedure that reminds the reader that an emergency classification assessment should be 

performed. 

 

3 Where appropriate, the Notes shown in the generic guidance typically include the event/condition ECL and the 

duration time specified in the EAL. If developers prefer to have several ICs reference a common NOTE on a 

wallboard display, it is acceptable to remove the ECL and time criterion and use a generic statement. For example, a 

common NOTE could read “The Emergency Director should declare the emergency promptly upon determining that 

the applicable EAL time has been exceeded, or will likely be exceeded.” 
NEI 99-01 (Revision 6)

November 2012

20 

4.6 BASIS DOCUMENT

A basis document is an integral part of an emergency classification scheme. The material 

in this document supports proper emergency classification decision-making by providing 

informing background and development information in a readily accessible format. It 

can be referred to in training situations and when making an actual emergency 

classification, if necessary. The document is also useful for establishing configuration 

management controls for EP-related equipment and explaining an emergency 

classification to offsite authorities. The content of the basis document should include, at 

a minimum, the following: 

 A site-specific Mode Applicability Matrix and description of operating modes, 

similar to that presented in section 3.5. 

 A discussion of the emergency classification and declaration process reflecting the 

material presented in Section 5. This material may be edited as needed to align with 

site-specific emergency plan and implementing procedure requirements.

 Each Initiating Condition along with the associated EALs or fission product barrier 

thresholds, Operating Mode Applicability, Notes and Basis information. 

 A listing of acronyms and defined terms, similar to that presented in Appendices A 

and B, respectively. This material may be edited as needed to align with site-specific

characteristics. 

 Any site-specific background or technical appendices that the developers believe 

would be useful in explaining or using elements of the emergency classification 

scheme. 

A Basis section should not contain information that could modify the meaning or intent 

of the associated IC or EAL. Such information should be incorporated within the IC or 

EAL statements, or as an EAL Note. Information in the Basis should only clarify and 

inform decision-making for an emergency classification.

Basis information should be readily available to be referenced, if necessary, by the 

Emergency Director. For example, a copy of the basis document could be maintained in 

the appropriate emergency response facilities.

Because the information in a basis document can affect emergency classification

decision-making (e.g., the Emergency Director refers to it during an event), the NRC 

staff expects that changes to the basis document will be evaluated in accordance with the 

provisions of 10 CFR 50.54(q). 

4.7 EAL/THRESHOLD REFERENCES TO AOP AND EOP SETPOINTS/CRITERIA 

As reflected in the generic guidance, the criteria/values used in several EALs and fission 

product barrier thresholds may be drawn from a plant’s AOPs and EOPs. This approach 

is intended to maintain good alignment between operational diagnoses and emergency 

classification assessments. Developers should verify that 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. 
NEI 99-01 (Revision 6)

November 2012

21 

4.8 DEVELOPER AND USER FEEDBACK

Questions or comments concerning the material in this document may be directed to the 

NEI Emergency Preparedness staff, NEI EAL task force members or submitted to the 

Emergency Preparedness Frequently Asked Questions process. 
NEI 99-01 (Revision 6)

November 2012

22 

5 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS

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 for 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 

ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, 

chemistry sampling, RCS leak rate calculation, etc.); the EAL and/or the associated basis 

discussion will identify the necessary analysis. In these cases, the 15-minute declaration 

period starts with the availability of the analysis results that show the threshold to be 
NEI 99-01 (Revision 6)

November 2012

23 

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. The NEI 99-

01 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/DPRISG-01. 

5.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS

When multiple emergency events or conditions are present, the user will identify all met 

or exceeded EALs. The highest applicable ECL identified during this review is declared. 

For example:

 If an Alert EAL and a Site Area Emergency EAL are met, whether at one unit or at 

two different units, a Site Area Emergency should be declared. 

There is no “additive” effect from multiple EALs meeting the same ECL. For example:

 If two Alert EALs are met, whether at one unit or at two different units, an Alert

should be declared. 

Related guidance concerning classification of rapidly escalating events or conditions is 

provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance 

for Emergency Notifications During Quickly Changing Events. 

5.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION

The mode in effect at the time that an event or condition occurred, and prior to any plant 
NEI 99-01 (Revision 6)

November 2012

24 

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.

The following approach to downgrading or terminating an ECL is recommended. 

ECL Action When Condition No Longer Exists 

Unusual Event Terminate the emergency in accordance with plant 

procedures. 

Alert Downgrade or terminate the emergency in 

accordance with plant procedures.

Site Area Emergency with no 

long-term plant damage

Downgrade or terminate the emergency in 

accordance with plant procedures.

Site Area Emergency with 

long-term plant damage

Terminate the emergency and enter recovery in 

accordance with plant procedures.

General Emergency Terminate the emergency and enter recovery in 

accordance with plant procedures. 
NEI 99-01 (Revision 6)

November 2012

25 

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 ICs and EALs define a variety of specific 

occurrences that have potential or actual safety significance. By their nature, some of 

these events may be short-lived and, thus, over before the emergency classification 

assessment can be completed. If an event occurs that meets or exceeds an EAL, the 

associated ECL must be declared regardless of its continued presence at the time of 

declaration. Examples of such events include a failure of the reactor protection system to

automatically scram/trip the reactor followed by a successful manual scram/trip or an 

earthquake. 

5.8 CLASSIFICATION OF TRANSIENT CONDITIONS

Many of the ICs and/or EALs contained in this document employ time-based criteria. 

These criteria will require that the IC/EAL conditions be present for a defined period of 

time before an emergency declaration is warranted. In cases where no time-based 

criterion is specified, it is recognized that some transient conditions may cause an EAL to 

be met for a brief period of time (e.g., a few seconds to a few minutes). The following 

guidance should be applied to the classification of these conditions. 

EAL momentarily met during expected plant response - In instances where an EAL is 

briefly met during an expected (normal) plant response, an emergency declaration is not 

warranted provided that associated systems and components are operating as expected, 

and operator actions are performed in accordance with procedures. 

EAL momentarily met but the condition is corrected prior to an emergency 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. 
NEI 99-01 (Revision 6)

November 2012

26 

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. 
NEI 99-01 (Revision 6)

December 2010

27

6 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT ICS/EALS

Table A-1: Recognition Category “A” Initiating Condition Matrix 

UNUSUAL EVENT ALERT SITE AREA 

EMERGENCY

GENERAL 

EMERGENCY

AU1 Release of 

gaseous or liquid 

radioactivity greater 

than 2 times the (sitespecific effluent 

release controlling 

document) limits for 

60 minutes or longer. 

Op. Modes: All 

AA1 Release of 

gaseous or liquid 

radioactivity resulting in 

offsite dose greater than 

10 mrem TEDE or 50 

mrem thyroid CDE. 

Op. Modes: All 

AS1 Release of 

gaseous radioactivity 

resulting in offsite dose 

greater than 100 mrem 

TEDE or 500 mrem 

thyroid CDE. 

Op. Modes: All 

AG1 Release of 

gaseous radioactivity 

resulting in offsite 

dose greater than 

1,000 mrem TEDE 

or 5,000 mrem 

thyroid CDE. 

Op. Modes: All

AU2 UNPLANNED 

loss of water level 

above irradiated fuel. 

Op. Modes: All

AA2 Significant 

lowering of water level 

above, or damage to, 

irradiated fuel. 

Op. Modes: All 

AS2 Spent fuel pool 

level at (site-specific 

Level 3 description). 

Op. Modes: All

AG2 Spent fuel 

pool level cannot be 

restored to at least 

(site-specific Level 3 

description) for 60 

minutes or longer. 

Op. Modes: All

AA3 Radiation levels 

that impede access to 

equipment necessary for 

normal plant operations, 

cooldown or shutdown. 

Op. Modes: All

Table intended for use by 

EAL developers. 

Inclusion in licensee 

documents is not required. 
NEI 99-01 (Revision 6)

November 2012

28 

AU1

ECL: Notification of Unusual Event

Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the (sitespecific effluent release controlling document) limits for 60 minutes or longer. 

Operating Mode Applicability: All

Example Emergency Action Levels: (1 or 2 or 3) 

Notes:

 The Emergency Director should declare the Unusual Event promptly upon determining that 

60 minutes has been exceeded, or will likely be exceeded.

 If an ongoing release is detected and the release start time is unknown, assume that the 

release duration has exceeded 60 minutes. 

 If the effluent flow past an effluent monitor is known to have stopped due to actions to 

isolate the release path, then the effluent monitor reading is no longer valid for classification 

purposes. 

(1) Reading on ANY effluent radiation monitor greater than 2 times the (site-specific effluent 

release controlling document) limits for 60 minutes or longer: 

(site-specific monitor list and threshold values corresponding to 2 times the controlling 

document limits)

(2) Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint 

established by a current radioactivity discharge permit for 60 minutes or longer. 

(3) Sample analysis for a gaseous or liquid release indicates a concentration or release rate 

greater than 2 times the (site-specific effluent release controlling document) limits for 60 

minutes or longer. 

Basis:

This IC addresses a potential decrease in the level of safety of the plant as indicated by a lowlevel radiological release that exceeds regulatory commitments for an extended period of time 

(e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored 

or un-monitored, including those for which a radioactivity discharge permit is normally prepared.

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

effluents to the environment. Further, there are administrative controls established to prevent 

unintentional releases, and to control and monitor intentional releases. The occurrence of an 

extended, uncontrolled radioactive release to the environment is indicative of degradation in 

these features and/or controls. 
NEI 99-01 (Revision 6)

November 2012

29

Radiological effluent EALs are also included to provide a basis for classifying events and 

conditions that cannot be readily or appropriately classified on the basis of plant conditions 

alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses 

the spectrum of possible accident events and conditions. 

Classification based on effluent monitor readings assumes that a release path to the environment 

is established. If the effluent flow past an effluent monitor is known to have stopped due to 

actions to isolate the release path, then the effluent monitor reading is no longer valid for 

classification purposes. 

Releases should not be prorated or averaged. For example, a release exceeding 4 times release 

limits for 30 minutes does not meet the EAL.

EAL #1 - This EAL addresses normally occurring continuous radioactivity releases from 

monitored gaseous or liquid effluent pathways. 

EAL #2 - This EAL addresses radioactivity releases that cause effluent radiation monitor 

readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL 

will typically be associated with planned batch releases from non-continuous release pathways 

(e.g., radwaste, waste gas).

EAL #3 - This EAL addresses uncontrolled gaseous or liquid releases that are detected by 

sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of 

radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.).

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

Developer Notes:

The “site-specific effluent release controlling document” is the Radiological Effluent Technical 

Specifications (RETS) or, for plants that have implemented Generic Letter 89-014

, the Offsite 

Dose Calculation Manual (ODCM). These documents implement regulations related to effluent 

controls (e.g., 10 CFR Part 20 and 10 CFR Part 50, Appendix I). As appropriate, the RETS or 

ODCM methodology should be used for establishing the monitor thresholds for this IC. 

Listed monitors should include the effluent monitors described in the RETS or ODCM. 

Developers may also consider including installed monitors associated with other potential 

effluent pathways that are not described in the RETS or ODCM56

. If included, EAL values for 

these monitors should be determined using the most applicable dose/release limits presented in 

the RETS or ODCM. It is recognized that a calculated EAL value may be below what the 

monitor can read; in that case, the monitor does not need to be included in the list. Also, some 

monitors may not be governed by Technical Specifications or other license-related related 

 

4 Implementation of Programmatic Controls for Radiological Effluent Technical Specifications in the Administrative 

Controls Section of the Technical Specifications and the Relocation of Procedural Details of RETS to the Offsite 

Dose Calculation Manual or to the Process Control Program 5 This includes consideration of the effluent monitors described in the site emergency plan section(s) which address 

the requirements of 10 CFR 50.47(b)(8) and (9).

6 Developers should keep in mind the requirements of 10 CFR 50.54(q) and the guidance provided by INPO related 

to emergency response equipment when considering the addition of other effluent monitors. 
NEI 99-01 (Revision 6)

November 2012

30 

requirements; therefore, it is important that the associated EAL and basis section clearly identify

any limitations on the use or availability of these monitors. 

Some sites may find it advantageous to address gaseous and liquid releases with separate EALs. 

Radiation monitor readings should reflect values that correspond to a radiological release 

exceeding 2 times a release control limit. The controlling document typically describes 

methodologies for determining effluent radiation monitor setpoints; these methodologies should 

be used to determine EAL values. In cases where a methodology is not adequately defined, 

developers should determine values consistent with effluent control regulations (e.g., 10 CFR 

Part 20 and 10 CFR Part 50 Appendix I) and related guidance. 

For EAL #2 - Values in this EAL should be 2 times the setpoint established by the radioactivity 

discharge permit to warn of a release that is not in compliance with the specified limits. 

Indexing the value in this manner ensures consistency between the EAL and the setpoint 

established by a specific discharge permit.

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 with a margin sufficient to ensure that an accurate monitor reading 

is available. For example, an EAL monitor reading might be set at 90% to 95% of the highest 

accurate monitor reading. This provision notwithstanding, 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. 

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 EALs. 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 within 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 system; approval will be 

considered on a case-by-case basis. 

ECL Assignment Attributes: 3.1.1.B 
NEI 99-01 (Revision 6)

November 2012

31

AU2

ECL: Notification of Unusual Event

Initiating Condition: UNPLANNED loss of water level above irradiated fuel. 

Operating Mode Applicability: All

Example Emergency Action Levels: 

(1) a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by 

ANY of the following:

 (site-specific level indications). 

 AND

 b. UNPLANNED rise in area radiation levels as indicated by ANY of the following 

radiation monitors. 

 (site-specific list of area radiation monitors) 

Basis:

This IC addresses a decrease in water level above irradiated fuel sufficient to cause elevated 

radiation levels. This condition could be a precursor to a more serious event and is also 

indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore 

a potential degradation in the level of safety of the plant. 

A water level decrease will be primarily determined by indications from available level 

instrumentation. Other sources of level indications may include reports from plant personnel 

(e.g., from a refueling crew) or video camera observations (if available). A significant drop in 

the water level may also cause an increase in the radiation levels of adjacent areas that can be 

detected by monitors in those locations. 

The effects of planned evolutions should be considered. For example, a refueling bridge area 

radiation monitor reading may increase due to planned evolutions such as lifting of the reactor 

vessel head or movement of a fuel assembly. Note that this EAL is applicable only in cases 

where the elevated reading is due to an UNPLANNED loss of water level. 

A drop in water level above irradiated fuel within the reactor vessel may be classified in 

accordance Recognition Category C during the Cold Shutdown and Refueling modes. 

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

Developer Notes:

The “site-specific level indications” are those indications that may be used to monitor water level 

in the various portions of the REFUELING PATHWAY. Specify the mode applicability of a 

particular indication if it is not available in all modes. 
NEI 99-01 (Revision 6)

November 2012

32 

The “site-specific list of area radiation monitors” should contain those area radiation monitors 

that would be expected to have increased readings following a decrease in water level in the sitespecific REFUELING PATHWAY. In cases where a radiation monitor(s) is not available or 

would not provide a useful indication, consideration should be given to including alternate 

indications such as UNPLANNED changes in tank and/or sump levels. 

Development of the EALs should consider the availability and limitations of mode-dependent, or 

other controlled but temporary, radiation monitors. Specify the mode applicability of a particular 

monitor if it is not available in all modes.

ECL Assignment Attributes: 3.1.1.A and 3.1.1.B 
NEI 99-01 (Revision 6)

November 2012

33

AA1

ECL: Alert

Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater 

than 10 mrem TEDE or 50 mrem thyroid CDE. 

Operating Mode Applicability: All

Example Emergency Action Levels: (1 or 2 or 3 or 4) 

Notes:

 The Emergency Director should declare the Alert promptly upon determining that the 

applicable time has been exceeded, or will likely be exceeded. 

 If an ongoing release is detected and the release start time is unknown, assume that the 

release duration has exceeded 15 minutes. 

 If the effluent flow past an effluent monitor is known to have stopped due to actions to 

isolate the release path, then the effluent monitor reading is no longer valid for classification 

purposes. 

 The pre-calculated effluent monitor values presented in EAL #1 should be used for 

emergency classification assessments until the results from a dose assessment using actual 

meteorology are available.

(1) Reading on ANY of the following radiation monitors greater than the reading shown for 

15 minutes or longer: 

(site-specific monitor list and threshold values) 

(2) Dose assessment using actual meteorology indicates doses greater than 10 mrem TEDE 

or 50 mrem thyroid CDE at or beyond (site-specific dose receptor point). 

(3) Analysis of a liquid effluent sample indicates a concentration or release rate that would 

result in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond (sitespecific dose receptor point) for one hour of exposure. 

(4) Field survey results indicate EITHER of the following at or beyond (site-specific dose 

receptor point): 

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

or longer. 

 Analyses of field survey samples indicate thyroid CDE greater than 50 mrem for one 

hour of inhalation. 

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 
NEI 99-01 (Revision 6)

November 2012

34 

actual or potential substantial degradation of the level of safety of the plant as indicated by a 

radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled 

release). 

Radiological effluent EALs are also included to provide a basis for classifying events and 

conditions that cannot be readily or appropriately classified on the basis of plant conditions 

alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses 

the spectrum of possible accident events and conditions. 

The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE 

was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. 

Classification based on effluent monitor readings assumes that a release path to the environment 

is established. If the effluent flow past an effluent monitor is known to have stopped due to 

actions to isolate the release path, then the effluent monitor reading is no longer valid for 

classification purposes. 

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

Developer Notes:

While this IC may not be met absent challenges to one or more fission product barriers, it 

provides classification diversity and may be used to classify events that would not reach the 

same ECL based on plant status or the fission product matrix alone. For many of the DBAs 

analyzed in the Updated Final Safety Analysis Report, the discriminator will not be the number 

of fission product barriers challenged, but rather the amount of radioactivity released to the 

environment. 

The EPA PAGs are expressed in terms of the sum of the effective dose equivalent (EDE) and the 

committed effective dose equivalent (CEDE), or as the thyroid committed dose equivalent 

(CDE). For the purpose of these IC/EALs, the dose quantity total effective dose equivalent 

(TEDE), as defined in 10 CFR § 20, is used in lieu of “...sum of EDE and CEDE....”. 

The EPA PAG guidance provides for the use of adult thyroid dose conversion factors; however, 

some states have decided to base protective actions on child thyroid CDE. Nuclear power plant 

ICs/EALs need to be consistent with the protective action methodologies employed by the States 

within their EPZs. The thyroid CDE dose used in the IC and EALs should be adjusted as 

necessary to align with State protective action decision-making criteria. 

The “site-specific monitor list and threshold values” should be determined with consideration of 

the following: 

 Selection of the appropriate installed gaseous and liquid effluent monitors. 

 The effluent monitor readings should correspond to a dose of 10 mrem TEDE or 50 mrem 

thyroid CDE at the “site-specific dose receptor point” (consistent with the calculation 

methodology employed) for one hour of exposure. 

 Monitor readings will be calculated using a set of assumed meteorological data or 

atmospheric dispersion factors; the data or factors selected for use should be the same as 

those employed to calculate the monitor readings for ICs AS1 and AG1. Acceptable sources 
NEI 99-01 (Revision 6)

November 2012

35

of this information include, but are not limited to, the RETS/ODCM and values used in the 

site’s emergency dose assessment methodology. 

 The calculation of monitor readings will also require use of an assumed release isotopic mix; 

the selected mix should be the same as that employed to calculate monitor readings for ICs 

AS1 and AG1. Acceptable sources of this information include, but are not limited to, the 

RETS/ODCM and values used in the site’s emergency dose assessment methodology. 

 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 ECL.

The “site-specific dose receptor point” is the distance(s) and/or locations used by the licensee to 

distinguish between on-site and offsite doses. The selected distance(s) and/or locations should 

reflect the content of the emergency plan, and the procedural methodology used to determine 

offsite doses and Protective Action Recommendations. 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 with a margin sufficient to ensure that an accurate monitor reading 

is available. For example, an EAL monitor reading might be set at 90% to 95% of the highest 

accurate monitor reading. This provision notwithstanding, 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 survey 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 EALs. 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 within 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 system; approval will be 

considered on a case-by-case basis. 

ECL Assignment Attributes: 3.1.2.C 
NEI 99-01 (Revision 6)

November 2012

36 

AA2

ECL: Alert

Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel. 

Operating Mode Applicability: All

Example Emergency Action Levels: (1 or 2 or 3) 

(1) Uncovery of irradiated fuel in the REFUELING PATHWAY. 

(2) Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated 

by ANY of the following radiation monitors: 

(site-specific listing of radiation monitors, and the associated readings, setpoints and/or 

alarms) 

(3) Lowering of spent fuel pool level to (site-specific Level 2 value). [See Developer Notes] 

Basis:

This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel 

assembly, or a significant lowering of water level within the spent fuel pool (see Developer 

Notes). These events present radiological safety challenges to plant personnel and are precursors 

to a release of radioactivity to the environment. As such, they represent an actual or potential 

substantial degradation of the level of safety of the plant. 

This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded 

storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the 

CONFINEMENT BOUNDARY is classified in accordance with IC E-HU1. 

Escalation of the emergency would be based on either Recognition Category A or C ICs.

EAL #1

This EAL escalates from AU2 in that the loss of level, in the affected portion of the 

REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated 

fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation 

(e.g., reports from personnel or camera images), as well as significant changes in water and 

radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off 

curve). Classification of an event using this EAL should be based on the totality of available 

indications, reports and observations. 

While an area radiation monitor could detect an increase in a dose rate due to a lowering of water 

level in some portion of the REFUELING PATHWAY, the reading may not be a reliable 

indication of whether or not the fuel is actually uncovered. To the degree possible, readings

should be considered in combination with other available indications of inventory loss. 

A drop in water level above irradiated fuel within the reactor vessel may be classified in 

accordance Recognition Category C during the Cold Shutdown and Refueling modes. 
NEI 99-01 (Revision 6)

November 2012

37

EAL #2

This EAL addresses a release of radioactive material caused by mechanical damage to irradiated 

fuel. Damaging events may include the dropping, bumping or binding of an assembly, or 

dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be 

considered in conjunction with in-plant reports or observations of a potential fuel damaging 

event (e.g., a fuel handling accident). 

EAL #3

Spent fuel pool water level at this value is within the lower end of the level range necessary to 

prevent significant dose consequences from direct gamma radiation to personnel performing 

operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent 

fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool 

the irradiated fuel assembles stored in the pool.

Escalation of the emergency classification level would be via ICs AS1 or AS2 (see AS2 

Developer Notes).

Developer Notes:

For EAL #1 

Depending upon the availability and range of instrumentation, this EAL may include specific 

readings indicative of fuel uncovery; consider water and radiation level readings. Specify the 

mode applicability of a particular indication if it is not available in all modes. 

For EAL #2 

The “site-specific listing of radiation monitors, and the associated readings, setpoints and/or 

alarms” should contain those radiation monitors that could be used to identify damage to an 

irradiated fuel assembly (e.g., confirmatory of a release of fission product gases from irradiated 

fuel).

For EALs #1 and #2 

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 radiation value beyond the 

operating or display range of the installed radiation monitor. In those cases, EAL values should 

be determined with a margin sufficient to ensure that an accurate monitor reading is available. 

For example, an EAL monitor reading might be set at 90% to 95% of the highest accurate 

monitor reading. This provision notwithstanding, 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. 
NEI 99-01 (Revision 6)

November 2012

38 

To further promote accurate classification, developers should consider if some combination of 

monitors could be specified in the EAL to build-in an appropriate level of corroboration between 

monitor readings into the classification assessment.

Development of the EALs should also consider the availability and limitations of modedependent, or other controlled but temporary, radiation monitors. Specify the mode applicability 

of a particular monitor if it is not available in all modes.

For EAL #3 

In accordance with the discussion in Section 1.4, NRC Order EA-12-051, it is recommended that 

this EAL be implemented when the enhanced spent fuel pool level instrumentation is available 

for use. The “site-specific Level 2 value” is usually the spent fuel pool level that is adequate to 

provide substantial radiation shielding for a person standing on the spent fuel pool operating 

deck. This site-specific level is determined in accordance with NRC Order EA-12-051 and NEI 

12-02, and applicable owner’s group guidance. 

Developers should modify the EAL and/or Basis section to reflect any site-specific constraints or 

limitations associated with the design or operation of instrumentation used to determine the 

Level 2 value.

ECL Assignment Attributes: 3.1.2.B and 3.1.2.C 
NEI 99-01 (Revision 6)

November 2012

39

AA3

ECL: Alert

Initiating Condition: Radiation levels that impede access to equipment necessary for normal

plant operations, cooldown or shutdown. 

Operating Mode Applicability: All

Example Emergency Action Levels: (1 or 2)

Note: If the equipment in the listed room or area was already inoperable or out-of-service

before the event occurred, then no emergency classification is warranted. 

(1) Dose rate greater than 15 mR/hr in ANY of the following areas: 

 Control Room 

 Central Alarm Station

 (other site-specific areas/rooms) 

(2) An UNPLANNED event results in radiation levels that prohibit or impede access to any 

of the following plant rooms or areas: 

(site-specific list of plant rooms or areas with entry-related mode applicability identified) 

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. 

For EAL #2, an Alert declaration is warranted if entry into the affected room/area is, or may be, 

procedurally required during the plant operating mode in effect at the time of the elevated 

radiation levels. The emergency classification is not contingent upon whether entry is actually 

necessary at the time of the increased radiation levels. Access should be considered as impeded 

if extraordinary measures are necessary to facilitate entry of personnel into the affected 

room/area (e.g., installing temporary shielding, requiring use of non-routine protective 

equipment, requesting an extension in dose limits beyond normal administrative limits).

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

 The plant is in an operating mode different than the mode specified for the affected 

room/area (i.e., entry is not required during the operating mode in effect at the time of the 

elevated radiation levels). For example, the plant is in Mode 1 when the radiation increase

occurs, and the procedures used for normal operation, cooldown and shutdown do not require 

entry into the affected room until Mode 4. 
NEI 99-01 (Revision 6)

November 2012

40 

 The increased radiation levels are a result of a planned activity that includes compensatory 

measures which address the temporary inaccessibility of a room or area (e.g., radiography, 

spent filter or resin transfer, etc.). 

 The action for which room/area entry is required is of an administrative or record keeping 

nature (e.g., normal rounds or routine inspections). 

 The access control measures are of a conservative or precautionary nature, and would not 

actually prevent or impede a required action. 

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

ICs. 

Developer Notes:

EAL #1

The value of 15mR/hr is derived from the GDC 19 value of 5 rem in 30 days with adjustment for 

expected occupancy times. 

The “other site-specific areas/rooms” should include any areas or rooms requiring continuous 

occupancy to maintain normal plant operation, or to perform a normal cooldown and shutdown. 

EAL #2 

The “site-specific list of plant rooms or areas with entry-related mode applicability identified” 

should specify those rooms or areas that contain equipment which require a manual/local action 

as specified in operating procedures used for normal plant operation, cooldown and shutdown. 

Do not include rooms or areas in which actions of a contingent or emergency nature would be 

performed. (e.g., an action to address an off-normal or emergency condition such as emergency 

repairs, corrective measures or emergency operations). In addition, the list should specify the 

plant mode(s) during which entry would be required for each room or area. 

The list should not include rooms or areas for which entry is required solely to perform actions 

of an administrative or record keeping nature (e.g., normal rounds or routine inspections). 

If the equipment in the listed room or area was already inoperable, or out-of-service, before the 

event occurred, then no emergency should be declared since the event will have no adverse 

impact beyond that already allowed by Technical Specifications at the time of the event.

Rooms and areas listed in EAL #1 do not need to be included in EAL #2, including the Control 

Room. 

ECL Assignment Attributes: 3.1.2.C 
NEI 99-01 (Revision 6)

November 2012

41

AS1

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

Example Emergency Action Levels: (1 or 2 or 3) 

Notes:

 The Emergency Director should declare the Site Area Emergency promptly upon 

determining that the applicable time has been exceeded, or will likely be exceeded. 

 If an ongoing release is detected and the release start time is unknown, assume that the 

release duration has exceeded 15 minutes. 

 If the effluent flow past an effluent monitor is known to have stopped due to actions to 

isolate the release path, then the effluent monitor reading is no longer valid for classification 

purposes. 

 The pre-calculated effluent monitor values presented in EAL #1 should be used for 

emergency classification assessments until the results from a dose assessment using actual 

meteorology are available. 

(1) Reading on ANY of the following radiation monitors greater than the reading shown for 

15 minutes or longer: 

(site-specific monitor list and threshold values) 

(2) Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE 

or 500 mrem thyroid CDE at or beyond (site-specific dose receptor point). 

(3) Field survey results indicate EITHER of the following at or beyond (site-specific dose 

receptor point): 

 Closed window dose rates greater than 100 mR/hr expected to continue for 60 

minutes or longer. 

 Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one 

hour of inhalation. 

Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite 

doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes 

both monitored and un-monitored releases. Releases of this magnitude are associated with the 

failure of plant systems needed for the protection of the public. 

Radiological effluent EALs are also included to provide a basis for classifying events and 

conditions that cannot be readily or appropriately classified on the basis of plant conditions 

alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses 
NEI 99-01 (Revision 6)

November 2012

42 

the spectrum of possible accident events and conditions. 

The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE 

was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. 

Classification based on effluent monitor readings assumes that a release path to the environment 

is established. If the effluent flow past an effluent monitor is known to have stopped due to 

actions to isolate the release path, then the effluent monitor reading is no longer valid for 

classification purposes. 

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

Developer Notes:

While this IC may not be met absent challenges to multiple fission product barriers, it provides 

classification diversity and may be used to classify events that would not reach the same ECL 

based on plant status or the fission product matrix alone. For many of the DBAs analyzed in the 

Updated Final Safety Analysis Report, the discriminator will not be the number of fission 

product barriers challenged, but rather the amount of radioactivity released to the environment. 

The EPA PAGs are expressed in terms of the sum of the effective dose equivalent (EDE) and the 

committed effective dose equivalent (CEDE), or as the thyroid committed dose equivalent 

(CDE). For the purpose of these IC/EALs, the dose quantity total effective dose equivalent 

(TEDE), as defined in 10 CFR § 20, is used in lieu of “...sum of EDE and CEDE....”. 

The EPA PAG guidance provides for the use of adult thyroid dose conversion factors; however, 

some states have decided to base protective actions on child thyroid CDE. Nuclear power plant 

ICs/EALs need to be consistent with the protective action methodologies employed by the States 

within their EPZs. The thyroid CDE dose used in the IC and EALs should be adjusted as 

necessary to align with State protective action decision-making criteria. 

The “site-specific monitor list and threshold values” should be determined with consideration of 

the following: 

 Selection of the appropriate installed gaseous effluent monitors. 

 The effluent monitor readings should correspond to a dose of 100 mrem TEDE or 500 mrem 

thyroid CDE at the “site-specific dose receptor point” (consistent with the calculation 

methodology employed) for one hour of exposure. 

 Monitor readings will be calculated using a set of assumed meteorological data or 

atmospheric dispersion factors; the data or factors selected for use should be the same as 

those employed to calculate the monitor readings for ICs AA1 and AG1. Acceptable sources 

of this information include, but are not limited to, the RETS/ODCM and values used in the 

site’s emergency dose assessment methodology. 

 The calculation of monitor readings will also require use of an assumed release isotopic mix; 

the selected mix should be the same as that employed to calculate monitor readings for ICs 

AA1 and AG1. Acceptable sources of this information include, but are not limited to, the 

RETS/ODCM and values used in the site’s emergency dose assessment methodology. 
NEI 99-01 (Revision 6)

November 2012

43

 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 ECL. 

The “site-specific dose receptor point” is the distance(s) and/or locations used by the licensee to 

distinguish between on-site and offsite doses. The selected distance(s) and/or locations should 

reflect the content of the emergency plan, and the procedural methodology used to determine 

offsite doses and Protective Action Recommendations. 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 with a margin sufficient to ensure that an accurate monitor reading 

is available. For example, an EAL monitor reading might be set at 90% to 95% of the highest 

accurate monitor reading. This provision notwithstanding, 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 survey 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 EALs. 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 within 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 system; approval will be 

considered on a case-by-case basis. 

ECL Assignment Attributes: 3.1.3.C 
NEI 99-01 (Revision 6)

November 2012

44 

AS2

[See Developer Notes] 

ECL: Site Area Emergency

Initiating Condition: Spent fuel pool level at (site-specific Level 3 description).

Operating Mode Applicability: All

Example Emergency Action Levels: 

(1) Lowering of spent fuel pool level to (site-specific Level 3 value).

Basis:

This IC addresses a significant loss of spent fuel pool inventory control and makeup capability 

leading to IMMINENT fuel damage. This condition entails major failures of plant functions 

needed for protection of the public and thus warrant a Site Area Emergency declaration. 

It is recognized that this IC would likely not be met until well after another Site Area Emergency 

IC was met; however, it is included to provide classification diversity. 

Escalation of the emergency classification level would be via IC AG1 or AG2.

Developer Notes:

In accordance with the discussion in Section 1.4, NRC Order EA-12-051, it is recommended that 

this IC and EAL be implemented when the enhanced spent fuel pool level instrumentation is 

available for use. The “site-specific Level 3 value” is usually that spent fuel pool level where 

fuel remains covered and actions to implement make-up water addition should no longer be 

deferred. This site-specific level is determined in accordance with NRC Order EA-12-051 and 

NEI 12-02, and applicable owner’s group guidance.

Developers should modify the EAL and/or Basis section to reflect any site-specific constraints or 

limitations associated with the design or operation of instrumentation used to determine the 

Level 3 value.

ECL Assignment Attributes: 3.1.3.B 
NEI 99-01 (Revision 6)

November 2012

45

AG1

ECL: General Emergency

Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 

1,000 mrem TEDE or 5,000 mrem thyroid CDE. 

Operating Mode Applicability: All

Example Emergency Action Levels: (1 or 2 or 3) 

Notes: 

 The Emergency Director should declare the General Emergency promptly upon determining 

that the applicable time has been exceeded, or will likely be exceeded. 

 If an ongoing release is detected and the release start time is unknown, assume that the 

release duration has exceeded 15 minutes. 

 If the effluent flow past an effluent monitor is known to have stopped due to actions to 

isolate the release path, then the effluent monitor reading is no longer valid for classification 

purposes. 

 The pre-calculated effluent monitor values presented in EAL #1 should be used for 

emergency classification assessments until the results from a dose assessment using actual 

meteorology are available. 

(1) Reading on ANY of the following radiation monitors greater than the reading shown for 

15 minutes or longer:

(site-specific monitor list and threshold values) 

(2) Dose assessment using actual meteorology indicates doses greater than 1,000 mrem 

TEDE or 5,000 mrem thyroid CDE at or beyond (site-specific dose receptor point). 

(3) Field survey results indicate EITHER of the following at or beyond (site-specific dose 

receptor point): 

 Closed window dose rates greater than 1,000 mR/hr expected to continue for 60 

minutes or longer. 

 Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrem for 

one hour of inhalation. 

Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite 

doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both 

monitored and un-monitored releases. Releases of this magnitude will require implementation of 

protective actions for the public. 

Radiological effluent EALs are also included to provide a basis for classifying events and 

conditions that cannot be readily or appropriately classified on the basis of plant conditions 
NEI 99-01 (Revision 6)

November 2012

46 

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. 

Developer Notes:

The effluent ICs/EALs are included to provide a basis for classifying events that cannot be 

readily classified on the basis of plant conditions alone. The inclusion of both types of ICs/EALs 

more fully addresses the spectrum of possible events and accidents. 

While this IC may not be met absent challenges to multiple fission product barriers, it provides 

classification diversity and may be used to classify events that would not reach the same ECL 

based on plant status or the fission product matrix alone. For many of the DBAs analyzed in the 

Updated Final Safety Analysis Report, the discriminator will not be the number of fission 

product barriers challenged, but rather the amount of radioactivity released to the environment. 

The EPA PAGs are expressed in terms of the sum of the effective dose equivalent (EDE) and the 

committed effective dose equivalent (CEDE), or as the thyroid committed dose equivalent 

(CDE). For the purpose of these IC/EALs, the dose quantity total effective dose equivalent 

(TEDE), as defined in 10 CFR § 20, is used in lieu of “...sum of EDE and CEDE....”. 

The EPA PAG guidance provides for the use of adult thyroid dose conversion factors; however, 

some states have decided to base protective actions on child thyroid CDE. Nuclear power plant 

ICs/EALs need to be consistent with the protective action methodologies employed by the States 

within their EPZs. The thyroid CDE dose used in the IC and EALs should be adjusted as 

necessary to align with State protective action decision-making criteria. 

The “site-specific monitor list and threshold values” should be determined with consideration of 

the following: 

 Selection of the appropriate installed gaseous effluent monitors. 

 The effluent monitor readings should correspond to a dose of 1,000 mrem TEDE or 5,000 

mrem thyroid CDE at the “site-specific dose receptor point” (consistent with the calculation 

methodology employed) for one hour of exposure. 

 Monitor readings will be calculated using a set of assumed meteorological data or 

atmospheric dispersion factors; the data or factors selected for use should be the same as 

those employed to calculate the monitor readings for ICs AA1 and AS1. Acceptable sources 

of this information include, but are not limited to, the RETS/ODCM and values used in the 

site’s emergency dose assessment methodology.

 The calculation of monitor readings will also require use of an assumed release isotopic mix; 

the selected mix should be the same as that employed to calculate monitor readings for ICs 

AA1 and AS1. Acceptable sources of this information include, but are not limited to, the 
NEI 99-01 (Revision 6)

November 2012

47

RETS/ODCM and values used in the site’s emergency dose assessment methodology. 

 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 ECL. 

The “site-specific dose receptor point” is the distance(s) and/or locations used by the licensee to 

distinguish between on-site and offsite doses. The selected distance(s) and/or locations should 

reflect the content of the emergency plan, and procedural methodology used to determine offsite 

doses and Protective 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 with a margin sufficient to ensure that an accurate monitor reading 

is available. For example, an EAL monitor reading might be set at 90% to 95% of the highest 

accurate monitor reading. This provision notwithstanding, 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 survey 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 EALs. 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 within 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 system; approval will be 

considered on a case-by-case basis. 

ECL Assignment Attributes: 3.1.4.C 
NEI 99-01 (Revision 6)

November 2012

48 

AG2 

[See Developer Notes] 

ECL: General Emergency

Initiating Condition: Spent fuel pool level cannot be restored to at least (site-specific Level 3 

description) for 60 minutes or longer.

Operating Mode Applicability: All

Example Emergency Action Levels: 

Note: The Emergency Director should declare the General Emergency promptly upon 

determining that 60 minutes has been exceeded, or will likely be exceeded. 

(1) Spent fuel pool level cannot be restored to at least (site-specific Level 3 value) for 60 

minutes or longer. 

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. 

Developer Notes:

In accordance with the discussion in Section 1.4, NRC Order EA-12-051, it is recommended that 

this IC and EAL be implemented when the enhanced spent fuel pool level instrumentation is 

available for use. The “site-specific Level 3 value” is usually that spent fuel pool level where 

fuel remains covered and actions to implement make-up water addition should no longer be 

deferred. This site-specific level is determined in accordance with NRC Order EA-12-051 and 

NEI 12-02, and applicable owner’s group guidance.

Developers should modify the EAL and/or Basis section to reflect any site-specific constraints or 

limitations associated with the design or operation of instrumentation used to determine the 

Level 3 value.

ECL Assignment Attributes: 3.1.4.C 
NEI 99-01 (Revision 6)

November 2012

49

7 COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS

Table C-1: Recognition Category “C” Initiating Condition Matrix 

UNUSUAL EVENT ALERT SITE AREA 

EMERGENCY

GENERAL 

EMERGENCY

CU1 UNPLANNED 

loss of (reactor 

vessel/RCS [PWR] or 

RPV [BWR]) inventory 

for 15 minutes or 

longer. 

Op. Modes: Cold 

Shutdown, Refueling

CA1 Loss of 

(reactor vessel/RCS 

[PWR] or RPV 

[BWR]) inventory. 

Op. Modes: Cold 

Shutdown, Refueling

CS1 Loss of (reactor 

vessel/RCS [PWR] or 

RPV [BWR]) 

inventory affecting 

core decay heat 

removal capability. 

Op. Modes: Cold 

Shutdown, Refueling

CG1 Loss of (reactor 

vessel/RCS [PWR] or 

RPV [BWR]) 

inventory affecting 

fuel clad integrity with 

containment 

challenged. 

Op. Modes: Cold 

Shutdown, Refueling

CU2 Loss of all but 

one AC power source 

to emergency buses for 

15 minutes or longer. 

Op. Modes: Cold 

Shutdown, Refueling, 

Defueled

CA2 Loss of all 

offsite and all onsite 

AC power to 

emergency buses for 

15 minutes or longer. 

Op. Modes: Cold 

Shutdown, Refueling, 

Defueled

CU3 UNPLANNED

increase in RCS 

temperature. 

Op. Modes: Cold 

Shutdown, Refueling 

CA3 Inability to 

maintain the plant in 

cold shutdown. 

Op. Modes: Cold 

Shutdown, Refueling

CU4 Loss of Vital 

DC power for 15 

minutes or longer. 

Op. Modes: Cold 

Shutdown, Refueling

CU5 Loss of all 

onsite or offsite 

communications 

capabilities.

Op. Modes: Cold 

Shutdown, Refueling, 

Defueled

Table intended for use by 

EAL developers. 

Inclusion in licensee 

documents is not required. 
NEI 99-01 (Revision 6)

November 2012

50 

UNUSUAL EVENT ALERT SITE AREA 

EMERGENCY

GENERAL 

EMERGENCY

CA6 Hazardous 

event affecting a 

SAFETY SYSTEM 

needed for the current 

operating mode. 

Op. Modes: Cold 

Shutdown, Refueling

Table intended for use by 

EAL developers. 

Inclusion in licensee 

documents is not required. 
NEI 99-01 (Revision 6)

November 2012

51

CU1

ECL: Notification of Unusual Event

Initiating Condition: UNPLANNED loss of (reactor vessel/RCS [PWR] or RPV [BWR]) 

inventory for 15 minutes or longer. 

Operating Mode Applicability: Cold Shutdown, Refueling

Example Emergency Action Levels: (1 or 2)

Note: The Emergency Director should declare the Unusual Event promptly upon determining 

that 15 minutes has been exceeded, or will likely be exceeded. 

(1) UNPLANNED loss of reactor coolant results in (reactor vessel/RCS [PWR] or RPV 

[BWR]) level less than a required lower limit for 15 minutes or longer. 

(2) a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level cannot be monitored. 

 AND

b. UNPLANNED increase in (site-specific sump and/or tank) levels. 

Basis:

This IC addresses the inability to restore and maintain water level to a required minimum level 

(or the lower limit of a level band), or a loss of the ability to monitor (reactor vessel/RCS [PWR] 

or RPV [BWR]) level concurrent with indications of coolant leakage. Either of these conditions 

is considered to be a potential degradation of the level of safety of the plant. 

Refueling evolutions that decrease 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 #1 recognizes that the minimum required (reactor vessel/RCS [PWR] or RPV [BWR]) level 

can change several times during the course of a refueling outage as different plant configurations 

and system lineups are implemented. This EAL is met if the minimum level, specified for the 

current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is 

typically specified in the applicable operating procedure but may be specified in another 

controlling document. 

The 15-minute threshold duration allows sufficient time for prompt operator actions to restore 

and maintain the expected water level. This criterion excludes transient conditions causing a 

brief lowering of water level. 

EAL #2 addresses a condition where all means to determine (reactor vessel/RCS [PWR] or RPV 

[BWR]) level have been lost. In this condition, operators may determine that an inventory loss is 

occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes 

must be evaluated against other potential sources of water flow to ensure they are indicative of 

leakage from the (reactor vessel/RCS [PWR] or RPV [BWR]). 
NEI 99-01 (Revision 6)

November 2012

52 

Continued loss of RCS inventory may result in escalation to the Alert emergency classification 

level via either IC CA1 or CA3. 

Developer Notes:

EAL #1 – It is recognized that the minimum allowable reactor vessel/RCS/RPV level may have 

many values over the course of a refueling outage. Developers should solicit input from licensed 

operators concerning the optimum wording for this EAL statement. In particular, determine if 

the generic wording is adequate to ensure accurate and timely classification, or if specific 

setpoints can be included without making the EAL statement unwieldy or potentially inconsistent 

with actions that may be taken during an outage. If specific setpoints are included, these should 

be drawn from applicable operating procedures or other controlling documents. 

EAL #2.b – Enter any “site-specific sump and/or tank” levels that could be expected to increase

if there were a loss of inventory (i.e., the lost inventory would enter the listed sump or tank). 

ECL Assignment Attributes: 3.1.1.A 
NEI 99-01 (Revision 6)

November 2012

53

CU2 

ECL: Notification of Unusual Event

Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes 

or longer. 

Operating Mode Applicability: Cold Shutdown, Refueling, Defueled

Example Emergency Action Levels: 

Note: The Emergency Director should declare the Unusual Event promptly upon determining 

that 15 minutes has been exceeded, or will likely be exceeded. 

(1) a. AC power capability to (site-specific emergency buses) is reduced to a single 

power source for 15 minutes or longer. 

AND

b. Any additional single power source failure will result in loss of all AC power to 

SAFETY SYSTEMS. 

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 safetyrelated equipment.

When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an 

Alert because of the increased time available to restore another power source to service. 

Additional time is available due to the reduced core decay heat load, and the lower temperatures 

and pressures in various plant systems. Thus, when in these modes, this condition is considered 

to be a potential degradation of the level of safety of the plant. 

An “AC power source” is a source recognized in AOPs and EOPs, and capable of supplying 

required power to an emergency bus. Some examples of this condition are presented below.

• A loss of all offsite power with a concurrent failure of all but one emergency power source 

(e.g., an onsite diesel generator). 

• A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel 

generators) with 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 single train of 

emergency buses being back-fed from an offsite power source. 

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

The subsequent loss of the remaining single power source would escalate the event to an Alert in 

accordance with IC CA2. 
NEI 99-01 (Revision 6)

November 2012

54 

Developer 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 required power to 

an AC emergency bus. For example, if a backup power 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 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. 

Developers should modify the bulleted examples 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 power circuit constitutes a 

single power source. For example, three independent 345kV offsite power circuits (i.e., 

incoming power lines) comprise three separate power sources. Independence may be determined 

from a review of the site-specific UFSAR, SBO 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 source is recognized in AOPs and EOPS, or beyond design basis accident 

response guidelines (e.g., FLEX support guidelines). Such power sources should generally meet 

the “Alternate ac source” definition provided in 10 CFR 50.2. 

At multi-unit stations, the EALs may credit compensatory measures that are proceduralized and 

can be implemented within 15 minutes. Consider capabilities such as power source cross-ties, 

“swing” 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 via 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. 

ECL Assignment Attributes: 3.1.1.A 
NEI 99-01 (Revision 6)

November 2012

55

CU3 

ECL: Notification of Unusual Event

Initiating Condition: UNPLANNED increase in RCS temperature. 

Operating Mode Applicability: Cold Shutdown, Refueling 

Example Emergency Action Levels: (1 or 2)

Note: The Emergency Director should declare the Unusual Event promptly upon determining 

that 15 minutes has been exceeded, or will likely be exceeded.

(1) UNPLANNED increase in RCS temperature to greater than (site-specific Technical 

Specification cold shutdown temperature limit). 

(2) Loss of ALL RCS temperature and (reactor vessel/RCS [PWR] or RPV [BWR]) level 

indication for 15 minutes or longer. 

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 #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. 

EAL #2 reflects a condition where there has been a significant loss of instrumentation capability 

necessary to monitor RCS conditions and operators would be unable to monitor key parameters 

necessary to assure core decay heat removal. During this condition, there is no immediate threat 

of fuel damage because the core decay heat load has been reduced since the cessation of power 

operation. 

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

indication. 
NEI 99-01 (Revision 6)

November 2012

56 

Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on 

exceeding plant configuration-specific time criteria.

Developer Notes:

For EAL #1, enter the “site-specific Technical Specification cold shutdown temperature limit” 

where indicated. 

ECL Assignment Attributes: 3.1.1.A 
NEI 99-01 (Revision 6)

November 2012

57

CU4 

ECL: Notification of Unusual Event

Initiating Condition: Loss of Vital DC power for 15 minutes or longer. 

Operating Mode Applicability: Cold Shutdown, Refueling 

Example Emergency Action Levels: 

Note: The Emergency Director should declare the Unusual Event promptly upon determining 

that 15 minutes has been exceeded, or will likely be exceeded.

(1) Indicated voltage is less than (site-specific bus voltage value) on required Vital DC buses

for 15 minutes or longer. 

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 B is in-service 

(operable), then a loss of Vital DC power affecting Train B would require the declaration of an 

Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency 

classification.

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

Depending upon the event, escalation of the emergency classification level would be via IC CA1 

or CA3, or an IC in Recognition Category A. 

Developer Notes:

The “site-specific bus voltage value” should be based on the minimum bus voltage necessary for 

adequate operation of SAFETY SYSTEM equipment. This voltage 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. 

The typical value for an entire battery set is approximately 105 VDC. For a 60 cell string of 

batteries, the cell voltage is approximately 1.75 Volts per cell. For a 58 string battery set, the 

minimum voltage is approximately 1.81 Volts per cell. 

ECL Assignment Attributes: 3.1.1.A 
NEI 99-01 (Revision 6)

November 2012

58 

CU5 

ECL: Notification of Unusual Event

Initiating Condition: Loss of all onsite or offsite communications capabilities.

Operating Mode Applicability: Cold Shutdown, Refueling, Defueled 

Example Emergency Action Levels: (1 or 2 or 3) 

(1) Loss of ALL of the following onsite communication methods: 

(site-specific list of communications methods)

(2) Loss of ALL of the following ORO communications methods: 

(site-specific list of communications methods) 

(3) Loss of ALL of the following NRC communications methods: 

(site-specific list of communications methods)

Basis:

This IC addresses a significant loss of on-site or offsite communications capabilities. While not 

a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs 

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 #1 addresses a total loss of the communications methods used in support of routine plant 

operations. 

EAL #2 addresses a total loss of the communications methods used to notify all OROs of an 

emergency declaration. The OROs referred to here are (see Developer Notes).

EAL #3 addresses a total loss of the communications methods used to notify the NRC of an 

emergency declaration.

Developer 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 
NEI 99-01 (Revision 6)

November 2012

59

Emergency Plan. The listing should include installed plant equipment and components, and not 

items owned and maintained by individuals. Example methods are ring-down/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 with 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 NRC as described in the site 

Emergency Plan. The listing should include installed plant equipment and components, and not 

items owned and maintained by individuals. These methods are typically the dedicated 

Emergency Notification System (ENS) telephone line and commercial telephone lines.

ECL Assignment Attributes: 3.1.1.C 
NEI 99-01 (Revision 6)

November 2012

60 

CA1

ECL: Alert

Initiating Condition: Loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory. 

Operating Mode Applicability: Cold Shutdown, Refueling 

Example Emergency Action Levels: (1 or 2)

Note: The Emergency Director should declare the Alert promptly upon determining that 15 

minutes has been exceeded, or will likely be exceeded. 

(1) Loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory as indicated by level less 

than (site-specific level).

(2) a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level cannot be monitored for 15 

minutes or longer 

AND

b. UNPLANNED increase in (site-specific sump and/or tank) levels due to a loss of 

(reactor vessel/RCS [PWR] or RPV [BWR]) inventory. 

Basis:

This IC addresses conditions that are precursors to a loss of the ability to adequately cool 

irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents 

a potential substantial reduction in the level of plant safety. 

For EAL #1, a lowering of water level below (site-specific level) indicates that operator actions 

have not been successful in restoring and maintaining (reactor vessel/RCS [PWR] or RPV 

[BWR]) water level. The heat-up rate of the coolant will increase as the available water 

inventory is reduced. A continuing decrease in water level will lead to core uncovery. 

Although related, EAL #1 is concerned with the loss of RCS inventory and not the potential 

concurrent effects on systems needed for decay heat removal (e.g., loss of a Residual Heat 

Removal suction point). An increase in RCS temperature caused by a loss of decay heat removal 

capability is evaluated under IC CA3.

For EAL #2, the inability to monitor (reactor vessel/RCS [PWR] or RPV [BWR]) level may be 

caused by instrumentation and/or power failures, or water level dropping below the range of 

available instrumentation. If water level cannot be monitored, operators may determine that an 

inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank 

level changes must be evaluated against other potential sources of water flow to ensure they are 

indicative of leakage from the (reactor vessel/RCS [PWR] or RPV [BWR]). 

The 15-minute duration for the loss of level indication was chosen because it is half of the EAL 

duration specified in IC CS1
NEI 99-01 (Revision 6)

November 2012

61

If the (reactor vessel/RCS [PWR] or RPV [BWR]) inventory level continues to lower, then 

escalation to Site Area Emergency would be via IC CS1.

Developer Notes:

For EAL #1 – the “site-specific level” should be based on either: 

 [BWR] Low-Low ECCS actuation setpoint/Level 2. This setpoint was chosen because it is a 

standard operationally significant setpoint at which some (typically high pressure ECCS) 

injection systems would automatically start and is a value significantly below the low RPV 

water level RPS actuation setpoint specified in IC CU1. 

 [PWR] The minimum allowable level that supports operation of normally used decay heat 

removal systems (e.g., Residual Heat Removal or Shutdown Cooling). If multiple levels 

exist, specify each along with the appropriate mode or configuration dependency criteria. 

For EAL #2 - The type and range of RCS level instrumentation may vary during an outage as the 

plant moves through various operating modes and refueling evolutions, particularly for a PWR. 

As appropriate to the plant design, alternate means of determining RCS level are installed to 

assure that the ability to monitor level within the range required by operating procedures will not 

be interrupted. The instrumentation range necessary to support implementation of operating 

procedures in the Cold Shutdown and Refueling modes may be different (e.g., narrower) than 

that required during modes higher than Cold Shutdown. 

Enter any “site-specific sump and/or tank” levels that could be expected to increase if there were 

a loss of inventory (i.e., the lost inventory would enter the listed sump or tank). 

ECL Assignment Attributes: 3.1.2.B 
NEI 99-01 (Revision 6)

November 2012

62 

CA2 

ECL: Alert

Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 

minutes or longer. 

Operating Mode Applicability: Cold Shutdown, Refueling, Defueled 

Example Emergency Action Levels: 

Note: The Emergency Director should declare the Alert promptly upon determining that 15 

minutes has been exceeded, or will likely be exceeded. 

(1) Loss of ALL offsite and ALL onsite AC Power to (site-specific emergency buses) for 15 

minutes or longer. 

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 CS1 or AS1. 

Developer 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 example, if a backup power 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 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. 

The EAL and/or Basis section may specify use of a non-safety-related power source provided 

that operation of this source is controlled in accordance with abnormal or emergency operating 

procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines). 

Such power sources should generally meet the “Alternate ac source” definition provided in 10 

CFR 50.2. 
NEI 99-01 (Revision 6)

November 2012

63

At multi-unit stations, the EALs may credit compensatory measures that are proceduralized and 

can be implemented within 15 minutes. Consider capabilities such as power source cross-ties, 

“swing” 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 via 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. 

ECL Assignment Attributes: 3.1.2.B 
NEI 99-01 (Revision 6)

November 2012

64 

CA3 

ECL: Alert

Initiating Condition: Inability to maintain the plant in cold shutdown. 

Operating Mode Applicability: Cold Shutdown, Refueling 

Example Emergency Action Levels: (1 or 2)

Note: The Emergency Director should declare the Alert promptly upon determining that the 

applicable time has been exceeded, or will likely be exceeded. 

(1) UNPLANNED increase in RCS temperature to greater than (site-specific Technical 

Specification cold shutdown temperature limit) for greater than the duration specified in 

the following table. 

Table: RCS Heat-up Duration Thresholds

RCS Status Containment Closure Status Heat-up Duration

Intact (but not at reduced 

inventory [PWR]) Not applicable 60 minutes* 

Not intact (or at reduced 

inventory [PWR])

Established 20 minutes* 

Not Established 0 minutes 

* If an RCS heat removal system is in operation within this time frame and RCS 

temperature is being reduced, the EAL is not applicable. 

(2) UNPLANNED RCS pressure increase greater than (site-specific pressure reading). (This 

EAL does not apply during water-solid plant conditions. [PWR])

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.

The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when 

CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS inventory is 

reduced (e.g., mid-loop operation in PWRs). 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 
NEI 99-01 (Revision 6)

November 2012

65

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 or is at 

reduced inventory [PWR], and CONTAINMENT CLOSURE is not established, no heat-up 

duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be 

released directly into the Containment atmosphere and subsequently to the environment, and 2) 

there is reduced reactor coolant inventory above the top of irradiated fuel. 

EAL #2 provides a pressure-based indication of RCS heat-up. 

Escalation of the emergency classification level would be via IC CS1 or AS1. 

Developer Notes:

For EAL #1 – Enter the “site-specific Technical Specification cold shutdown temperature limit” 

where indicated. The RCS should be considered intact or not intact in accordance with sitespecific criteria.

For EAL #2 - The “site-specific pressure reading” should be the lowest change in pressure that 

can be accurately determined using installed instrumentation, but not less than 10 psig. 

For PWRs, this IC and its associated EALs address the concerns raised by Generic Letter 88-17, 

Loss of Decay Heat Removal. A number of phenomena such as pressurization, vortexing, steam 

generator U-tube draining, RCS level differences when operating at a mid-loop condition, decay 

heat removal system design, and level instrumentation problems can lead to conditions where 

decay heat removal is lost and core uncovery can occur. NRC analyses show that there are 

sequences that can cause core uncovery in 15 to 20 minutes, and severe core damage within an 

hour after decay heat removal is lost. The allowed time frames are consistent with the guidance 

provided by Generic Letter 88-17 and believed to be conservative given that a low pressure 

Containment barrier to fission product release is established. 

ECL Assignment Attributes: 3.1.2.B 
NEI 99-01 (Revision 6)

November 2012

66 

CA6 

ECL: Alert

Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current 

operating mode. 

Operating Mode Applicability: Cold Shutdown, Refueling 

Example Emergency Action Levels: 

(1) a. The occurrence of ANY of the following hazardous events: 

 Seismic event (earthquake)

 Internal or external flooding event 

 High winds or tornado strike 

 FIRE

 EXPLOSION

 (site-specific hazards)

 Other events with similar hazard characteristics as determined by the Shift 

Manager

 AND

b. EITHER of the following: 

 1. Event damage has caused indications of degraded performance in at least 

one train of a SAFETY SYSTEM needed for the current operating mode. 

 OR 

 2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM 

component or structure needed for the current operating mode. 

Basis:

This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure 

containing SAFETY SYSTEM components, needed for the current operating mode. This 

condition significantly reduces the margin to a loss or potential loss of a fission product barrier, 

and therefore represents an actual or potential substantial degradation of the level of safety of the 

plant. 

EAL 1.b.1 addresses damage to a SAFETY SYSTEM train that is in service/operation since 

indications for it will be readily available. The indications of degraded performance should be 

significant enough to cause concern regarding the operability or reliability of the SAFETY 

SYSTEM train. 

EAL 1.b.2 addresses damage to a SAFETY SYSTEM component that is not in service/operation 

or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM 

components. Operators will make this determination based on the totality of available event and 
NEI 99-01 (Revision 6)

November 2012

67

damage report information. This is intended to be a brief assessment not requiring lengthy 

analysis or quantification of the damage.

Escalation of the emergency classification level would be via IC CS1 or AS1. 

Developer Notes:

For (site-specific hazards), developers should consider including other significant, site-specific 

hazards to the bulleted list contained in EAL 1.a (e.g., a seiche). 

Nuclear power plant SAFETY SYSTEMS are comprised of two or more separate and redundant 

trains of equipment in accordance with site-specific design criteria.

ECL Assignment Attributes: 3.1.2.B 
NEI 99-01 (Revision 6)

November 2012

68 

CS1

ECL: Site Area Emergency

Initiating Condition: Loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory affecting 

core decay heat removal capability.

Operating Mode Applicability: Cold Shutdown, Refueling 

Example Emergency Action Levels: (1 or 2 or 3)

Note: The Emergency Director should declare the Site Area Emergency promptly upon 

determining that 30 minutes has been exceeded, or will likely be exceeded. 

(1) a. CONTAINMENT CLOSURE not established.

AND

b. (Reactor vessel/RCS [PWR] or RPV [BWR]) level less than (site-specific level).

(2) a. CONTAINMENT CLOSURE established.

AND

b. (Reactor vessel/RCS [PWR] or RPV [BWR]) level less than (site-specific level). 

(3) a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level cannot be monitored for 30 

minutes or longer. 

AND

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

 (Site-specific radiation monitor) reading greater than (site-specific value)

 Erratic source range monitor indication [PWR] 

 UNPLANNED increase in (site-specific sump and/or tank) levels of sufficient 

magnitude to indicate core uncovery 

 (Other site-specific indications)

Basis:

This IC addresses a significant and prolonged loss of (reactor vessel/RCS [PWR] or RPV [BWR]) 

inventory control and makeup capability leading to IMMINENT fuel damage. The lost 

inventory may be due to a RCS component failure, a loss of configuration control or prolonged 

boiling of reactor coolant. These conditions entail major failures of plant functions needed for 

protection of the public and thus warrant a Site Area Emergency declaration. 

Following an extended loss of core decay heat removal and inventory makeup, decay heat will 

cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor 

vessel level cannot be restored, fuel damage is probable. 
NEI 99-01 (Revision 6)

November 2012

69

Outage/shutdown contingency plans typically provide for re-establishing or verifying 

CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control 

functions. The difference in the specified RCS/reactor vessel levels of EALs 1.b and 2.b reflect 

the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a 

fission product release to the environment. 

In EAL 3.a, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total 

loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate 

reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account 

for various accident progression and instrumentation uncertainties). It also allows sufficient time 

for performance of actions to terminate leakage, recover inventory control/makeup equipment 

and/or restore level monitoring. 

The inability to monitor (reactor vessel/RCS [PWR] or RPV [BWR]) level may be caused by 

instrumentation and/or power failures, or water level dropping below the range of available 

instrumentation. If water level cannot be monitored, operators may determine that an inventory 

loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level 

changes must be evaluated against other potential sources of water flow to ensure they are 

indicative of leakage from the (reactor vessel/RCS [PWR] or RPV [BWR]). 

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 CG1 or AG1. 

Developer Notes:

Accident analyses suggest that fuel damage may occur within one hour of uncovery depending 

upon the amount of time since shutdown; refer to Generic Letter 88-17, SECY 91-283, NUREG1449 and NUMARC 91-06. 

The type and range of RCS level instrumentation may vary during an outage as the plant moves 

through various operating modes and refueling evolutions, particularly for a PWR. As 

appropriate to the plant design, alternate means of determining RCS level are installed to assure 

that the ability to monitor level within the range required by operating procedures will not be 

interrupted. The instrumentation range necessary to support implementation of operating 

procedures in the Cold Shutdown and Refueling modes may be different (e.g., narrower) than 

that required during modes higher than Cold Shutdown. 

PWR

For EAL #1.b – the “site-specific level” is 6" below the bottom ID of the RCS loop. This is the 

level at 6” below the bottom ID of the reactor vessel penetration and not the low point of the 

loop. If the availability of on-scale level indication is such that this level value can be 

determined during some shutdown modes or conditions, but not others, then specify the modedependent and/or configuration states during which the level indication is applicable. If the 

design and operation of water level instrumentation is such that this level value cannot be 
NEI 99-01 (Revision 6)

November 2012

70 

determined at any time during Cold Shutdown or Refueling modes, then do not include EAL #1 

(classification will be accomplished in accordance with EAL #3). 

For EAL #2.b – The “site-specific level” should be approximately the top of active fuel. If the 

availability of on-scale level indication is such that this level value can be determined during 

some shutdown modes or conditions, but not others, then specify the mode-dependent and/or 

configuration states during which the level indication is applicable. If the design and operation 

of water level instrumentation is such that this level value cannot be determined at any time 

during Cold Shutdown or Refueling modes, then do not include EAL #2 (classification will be 

accomplished in accordance with EAL #3). 

For EAL #3.b – first bullet - As water level in the reactor vessel lowers, the dose rate above the 

core will increase. Enter a “site-specific radiation monitor” that could be used to detect core 

uncovery and the associated “site-specific value” indicative of core uncovery. It is recognized 

that the condition described by this IC may result in a radiation value beyond the operating or 

display range of the installed radiation monitor. In those cases, EAL values should be 

determined with a margin sufficient to ensure that an accurate monitor reading is available. For 

example, an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor 

reading. This provision notwithstanding, 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. 

To further promote accurate classification, developers should consider if some combination of 

monitors could be specified in the EAL to build-in an appropriate level of corroboration between 

monitor readings into the classification assessment.

For EAL #3.b – second bullet - Post-TMI accident studies indicated that the installed PWR 

nuclear instrumentation will operate erratically when the core is uncovered and that this should 

be used as a tool for making such determinations. 

For EAL #3.b – third bullet – Enter any ‘site-specific sump and/or tank” levels that could be 

expected to change if there were a loss of RCS/reactor vessel inventory of sufficient magnitude 

to indicate core uncovery. Specific level values may be included if desired.

For EAL #3.b – fourth bullet - Developers should determine if other reliable indicators exist to 

identify fuel uncovery (e.g., remote viewing using cameras). The goal is to identify any unique 

or site-specific indications, not already used elsewhere, that will promote timely and accurate 

emergency classification.

BWR

For EAL #1.b – “site-specific level” is the Low-Low-Low ECCS actuation setpoint / Level 1. 

The BWR Low-Low-Low ECCS actuation setpoint / Level 1 was chosen because it is a standard 

operationally significant setpoint at which some (typically low pressure ECCS) injection systems 

would automatically start and attempt to restore RPV level. This is a RPV water level value that 

is observable below the Low-Low/Level 2 value specified in IC CA1, but significantly above the 

Top of Active Fuel (TOAF) threshold specified in EAL #2. 

For EAL #2.b – The “site-specific level” should be for the top of active fuel. 
NEI 99-01 (Revision 6)

November 2012

71

For EAL #3.b – first bullet - As water level in the reactor vessel lowers, the dose rate above the 

core will increase. Enter a “site-specific radiation monitor” that could be used to detect core 

uncovery and the associated “site-specific value” indicative of core uncovery. It is recognized 

that the condition described by this IC may result in a radiation value beyond the operating or 

display range of the installed radiation monitor. In those cases, EAL values should be 

determined with a margin sufficient to ensure that an accurate monitor reading is available. For 

example, an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor 

reading. This provision notwithstanding, 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. 

To further promote accurate classification, developers should consider if some combination of 

monitors could be specified in the EAL to build-in an appropriate level of corroboration between 

monitor readings into the classification assessment. 

For BWRs that do not have installed radiation monitors capable of indicating core uncovery, 

alternate site-specific level indications of core uncovery should be used if available.

For EAL #3.b – second bullet - Because BWR source range monitor (SRM) nuclear 

instrumentation detectors are typically located below core mid-plane, this may not be a viable 

indicator of core uncovery for BWRs.

For EAL #3.b – third bullet – Enter any “site-specific sump and/or tank” levels that could be 

expected to change if there were a loss of RPV inventory of sufficient magnitude to indicate core 

uncovery. Specific level values may be included if desired. 

For EAL #3.b – fourth bullet - Developers should determine if other reliable indicators exist to 

identify fuel uncovery (e.g., remote viewing using cameras). The goal is to identify any unique 

or site-specific indications, not already used elsewhere, that will promote timely and accurate 

emergency classification. 

ECL Assignment Attributes: 3.1.3.B 
NEI 99-01 (Revision 6)

November 2012

72 

CG1

ECL: General Emergency

Initiating Condition: Loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory affecting 

fuel clad integrity with containment challenged.

Operating Mode Applicability: Cold Shutdown, Refueling 

Example Emergency Action Levels: (1 or 2)

Note: The Emergency Director should declare the General Emergency promptly upon 

determining that 30 minutes has been exceeded, or will likely be exceeded. 

(1) a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level less than (site-specific level) 

for 30 minutes or longer. 

AND

 b. ANY indication from the Containment Challenge Table (see below). 

(2) a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level cannot be monitored for 30 

minutes or longer. 

 AND

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

 (Site-specific radiation monitor) reading greater than (site-specific value)

 Erratic source range monitor indication [PWR] 

 UNPLANNED increase in (site-specific sump and/or tank) levels of sufficient 

magnitude to indicate core uncovery 

 (Other site-specific indications)

AND

c. ANY indication from the Containment Challenge Table (see below). 

Containment Challenge Table

 CONTAINMENT CLOSURE not established* 

 (Explosive mixture) exists inside containment 

 UNPLANNED increase in containment pressure 

 Secondary containment radiation monitor reading above (site-specific value) [BWR] 

* If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, 

then declaration of a General Emergency is not required. 
NEI 99-01 (Revision 6)

November 2012

73

Basis:

This IC addresses the inability to restore and maintain reactor vessel level above the top of active 

fuel with containment challenged. This condition represents actual or IMMINENT substantial 

core degradation or melting with potential for loss of containment integrity. Releases can be 

reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate 

site area. 

Following an extended loss of core decay heat removal and inventory makeup, decay heat will 

cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor 

vessel level cannot be restored, fuel damage is probable. 

With CONTAINMENT CLOSURE not established, there is a high potential for a direct and 

unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is reestablished prior to exceeding the 30-minute time limit, then declaration of a General Emergency 

is not required. 

The existence of an explosive mixture means, at a minimum, that the containment atmospheric 

hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration 

limit). A hydrogen burn will raise containment pressure and could result in collateral equipment 

damage leading to a loss of containment integrity. It therefore represents a challenge to 

Containment integrity. 

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core 

uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas 

monitors are out-of-service during an event leading to fuel cladding damage, it may not be 

possible to obtain a containment hydrogen gas concentration reading as ambient conditions 

within the containment will preclude personnel access. During periods when installed 

containment hydrogen gas monitors are out-of-service, operators may use the other listed 

indications to assess whether or not containment is challenged. 

In EAL 2.b, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total 

loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate 

reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account 

for various accident progression and instrumentation uncertainties). It also allows sufficient time 

for performance of actions to terminate leakage, recover inventory control/makeup equipment 

and/or restore level monitoring. 

The inability to monitor (reactor vessel/RCS [PWR] or RPV [BWR]) level may be caused by 

instrumentation and/or power failures, or water level dropping below the range of available 

instrumentation. If water level cannot be monitored, operators may determine that an inventory 

loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level 

changes must be evaluated against other potential sources of water flow to ensure they are 

indicative of leakage from the (reactor vessel/RCS [PWR] or RPV [BWR]).

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 
NEI 99-01 (Revision 6)

November 2012

74 

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. 

Developer Notes:

Accident analyses suggest that fuel damage may occur within one hour of uncovery depending 

upon the amount of time since shutdown; refer to Generic Letter 88-17, SECY 91-283, NUREG1449 and NUMARC 91-06. 

The type and range of RCS level instrumentation may vary during an outage as the plant moves 

through various operating modes and refueling evolutions, particularly for a PWR. As 

appropriate to the plant design, alternate means of determining RCS level are installed to assure 

that the ability to monitor level within the range required by operating procedures will not be 

interrupted. The instrumentation range necessary to support implementation of operating 

procedures in the Cold Shutdown and Refueling modes may be different (e.g., narrower) than 

that required during modes higher than Cold Shutdown. 

For EAL #1.a – The “site-specific level” should be approximately the top of active fuel. If the 

availability of on-scale level indication is such that this level value can be determined during 

some shutdown modes or conditions, but not others, then specify the mode-dependent and/or 

configuration states during which the level indication is applicable. If the design and operation 

of water level instrumentation is such that this level value cannot be determined at any time 

during Cold Shutdown or Refueling modes, then do not include EAL #1 (classification will be 

accomplished in accordance with EAL #2). 

For EAL #2.b - first bullet - As water level in the reactor vessel lowers, the dose rate above the 

core will increase. Enter a “site-specific radiation monitor” that could be used to detect core 

uncovery and the associated “site-specific value” indicative of core uncovery. It is recognized 

that the condition described by this IC may result in a radiation value beyond the operating or 

display range of the installed radiation monitor. In those cases, EAL values should be 

determined with a margin sufficient to ensure that an accurate monitor reading is available. For 

example, an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor 

reading. This provision notwithstanding, 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. 

To further promote accurate classification, developers should consider if some combination of 

monitors could be specified in the EAL to build-in an appropriate level of corroboration between 

monitor readings into the classification assessment. 

For BWRs that do not have installed radiation monitors capable of indicating core uncovery, 

alternate site-specific level indications of core uncovery should be used if available.

For EAL #2.b - second bullet - Post-TMI accident studies indicated that the installed PWR 

nuclear instrumentation will operate erratically when the core is uncovered and that this should 

be used as a tool for making such determinations. Because BWR Source Range Monitor (SRM) 

nuclear instrumentation detectors are typically located below core mid-plane, this may not be a 

viable indicator of core uncovery for BWRs. 
NEI 99-01 (Revision 6)

November 2012

75

For EAL #2.b – third bullet - Enter any “site-specific sump and/or tank” levels that could be 

expected to change if there were a loss of inventory of sufficient magnitude to indicate core 

uncovery. Specific level values may be included if desired. 

For EAL #2.b – fourth bullet - Developers should determine if other reliable indicators exist to 

identify fuel uncovery (e.g., remote viewing using cameras). The goal is to identify any unique 

or site-specific indications, not already used elsewhere, that will promote timely and accurate 

emergency classification.

For the Containment Challenge Table:

Site shutdown contingency plans typically provide for re-establishing CONTAINMENT 

CLOSURE following a loss of RCS heat removal or inventory control functions. 

For “Explosive mixture”, developers may enter the minimum containment atmospheric hydrogen

concentration necessary to support a hydrogen burn (i.e., the lower deflagration limit). A 

concurrent containment oxygen concentration may be included if the plant has this indication 

available in the Control Room. 

For BWRs, the use of secondary containment radiation monitors should provide indication of 

increased release that may be indicative of a challenge to secondary containment. The “sitespecific value” should be based on the EOP maximum safe values because these values are 

easily recognizable and have a defined basis.

ECL Assignment Attributes: 3.1.4.B 
NEI 99-01 (Revision 6)

November 2012

76 

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

Table E-1: Recognition Category “E” Initiating Condition Matrix 

UNUSUAL EVENT

E-HU1 Damage to a loaded cask 

CONFINEMENT BOUNDARY.

Op. Modes: All

Table intended for use by 

EAL developers. 

Inclusion in licensee 

documents is not required. 
ISFSI MALFUNCTION

77

E-HU1

ECL: Notification of Unusual Event

Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY.

Operating Mode Applicability: All

Example Emergency Action Levels: 

(1) Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact 

radiation reading greater than (2 times the site-specific cask specific technical 

specification allowable radiation level) on the surface of the spent fuel cask.

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 A IC AU1, is used here to 

distinguish between non-emergency and emergency conditions. The emphasis for this 

classification is the degradation in the level of safety of the spent fuel cask and not the magnitude 

of the associated dose or dose rate. It is recognized that in the case of extreme damage to a 

loaded cask, the fact that the “on-contact” dose rate limit is exceeded may be determined based 

on measurement of a dose rate at some distance from the cask.

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

Developer Notes:

The results of the ISFSI Safety Analysis Report (SAR) [per NUREG 1536], or a SAR referenced 

in the cask Certificate of Compliance and the related NRC Safety Evaluation Report, identify the 

natural phenomena events and accident conditions that could potentially affect the 

CONFINEMENT BOUNDARY. This EAL addresses damage that could result from the range of

identified natural or man-made events (e.g., a dropped or tipped over cask, EXPLOSION, FIRE, 

EARTHQUKE, etc.). 

The allowable radiation level for a spent fuel cask can be found in the cask’s technical 

specification located in the Certificate of Compliance. 

ECL Assignment Attributes: 3.1.1.B 
NEI 99-01 (Revision 6)

November 2012

78 

9 FISSION PRODUCT BARRIER ICS/EALS

Table 9-F-1: Recognition Category “F” Initiating 

Condition Matrix 

ALERT

FA1

Any Loss or any Potential Loss of either the 

Fuel Clad or RCS barrier. 

Op. Modes: Power Operation, Hot Standby, 

Startup, Hot Shutdown

SITE AREA EMERGENCY

FS1

Loss or Potential Loss of any two barriers.

Op. Modes: Power Operation, Hot Standby, 

Startup, Hot Shutdown

GENERAL EMERGENCY

FG1

Loss of any two barriers and Loss or 

Potential Loss of the third barrier. 

Op. Modes: Power Operation, Hot Standby, 

Startup, Hot Shutdown

See Table 9-F-2 for BWR EALs 

See Table 9-F-3 for PWR EALs

Developer Note: The adjacent logic flow diagram is for 

use by developers and is not required for site-specific 

implementation; however, a site-specific scheme must 

include some type of user-aid to facilitate timely and 

accurate classification of fission product barrier losses 

and/or potential losses. Such aids are typically comprised 

of logic flow diagrams, “scoring” criteria or checkboxtype matrices. The user-aid logic must be consistent with 

that of the adjacent diagram. 

3/3

2/3

1/2

Loss of at least 2 

Barriers?

LOSS POTENTIAL 

LOSS

FUEL CLAD

LOSS POTENTIAL 

LOSS

RCS

LOSS POTENTIAL 

LOSS

CONTAINMENT

LOSS POTENTIAL 

LOSS

FUEL CLAD

LOSS POTENTIAL 

LOSS

RCS

LOSS POTENTIAL 

LOSS

CONTAINMENT

LOSS POTENTIAL 

LOSS

FUEL CLAD

LOSS POTENTIAL 

LOSS

RCS

FG1 - Loss of ANY Two Barriers AND Loss or 

Potential Loss of Third Barrier -- YES --

FS1 - Loss or Potential Loss of ANY Two Barriers

-- NO -- 

FA1 - ANY Loss or ANY Potential Loss of EITHER

Fuel Clad OR RCS
NEI 99-01 (Revision 6)

November 2012

79

Developer Notes

1. The logic used for these initiating conditions reflects the following considerations: 

• The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Containment Barrier. 

• Unusual Event ICs associated with fission product barriers are addressed in Recognition Category S. 

2. For accident conditions involving a radiological release, evaluation of the fission product barrier thresholds will need to be performed in 

conjunction with dose assessments to ensure correct and timely escalation of the emergency classification. For example, an evaluation of the 

fission product barrier thresholds may result in a Site Area Emergency classification while a dose assessment may indicate that an EAL for 

General Emergency IC AG1 has been exceeded.

3. The fission product barrier thresholds specified within a scheme are expected to reflect plant-specific design and operating characteristics. 

This may require that developers create different thresholds than those provided in the generic guidance. 

4. Alternative presentation methods for the Recognition Category F ICs and fission product barrier thresholds are acceptable and include flow 

charts, block diagrams, and checklist-type tables. Developers must ensure that the site-specific method addresses all possible threshold 

combinations and classification outcomes shown in the BWR or PWR EAL fission product barrier tables. The NRC staff considers the 

presentation method of the Recognition Category F information to be an important user aid and may request a change to a particular proposed 

method if, among other reasons, the change is necessary to promote consistency across the industry. 

5. As used in this Recognition Category, the term RCS leakage encompasses not just those types defined in Technical Specifications but also 

includes the loss of RCS mass to any location– inside containment, a secondary-side system (i.e., PWR steam generator tube leakage), an 

interfacing system, or outside of containment. The release of liquid or steam mass from the RCS due to the as-designed/expected operation of 

a relief valve is not considered to be RCS leakage. 

6. At the Site Area Emergency level, classification decision-makers should maintain cognizance of how far present conditions are from meeting 

a threshold that would require a General Emergency declaration. For example, if the Fuel Clad and RCS fission product barriers were both 

lost, then there should be frequent assessments of containment radioactive inventory and integrity. Alternatively, if both the Fuel Clad and 

RCS fission product barriers were potentially lost, the Emergency Director would have more assurance that there was no immediate need to 

escalate to a General Emergency.

7. The ability to escalate to a higher emergency classification level in response to degrading conditions should be maintained. For example, a 

steady increase in RCS leakage would represent an increasing risk to public health and safety. 
NEI 99-01 (Revision 6)

November 2012

80 
NEI 99-01 (Revision 6)

November 2012

81

Table 9-F-2: BWR EAL Fission Product Barrier Table

Thresholds for LOSS or POTENTIAL LOSS of Barriers

FA1 ALERT

Any Loss or any Potential Loss of either the 

Fuel Clad or RCS barrier.

FS1 SITE AREA EMERGENCY

Loss or Potential Loss of any two barriers.

FG1 GENERAL EMERGENCY

Loss of any two barriers and Loss or 

Potential Loss of the third barrier.

Fuel Clad Barrier RCS Barrier Containment Barrier

LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS

1. RCS Activity 1. Primary Containment Pressure 1. Primary Containment Conditions

A. (Site-specific 

indications that 

reactor coolant 

activity is greater 

than 300 μCi/gm 

dose equivalent I131). 

Not Applicable A. Primary 

containment 

pressure greater 

than (site-specific 

value) due to RCS 

leakage.

Not Applicable A. UNPLANNED 

rapid drop in 

primary 

containment 

pressure following 

primary 

containment 

pressure rise 

OR

B. Primary 

containment 

pressure response 

not consistent with 

LOCA conditions.

A. Primary 

containment 

pressure greater 

than (sitespecific value)

OR

B. (site-specific 

explosive 

mixture) exists 

inside primary 

containment 

OR

C. HCTL exceeded.

2. RPV Water Level 2. RPV Water Level 2. RPV Water Level

A. Primary 

containment 

flooding required. 

A. RPV water level 

cannot be restored 

and maintained 

above (site-specific 

RPV water level 

corresponding to 

the top of active 

fuel) or cannot be 

A. RPV water level 

cannot be restored 

and maintained 

above (sitespecific RPV 

water level 

corresponding to 

the top of active 

Not Applicable Not Applicable A. Primary 

containment 

flooding 

required. 
NEI 99-01 (Revision 6)

November 2012

82 

Fuel Clad Barrier RCS Barrier Containment Barrier

LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS

determined. fuel) or cannot be 

determined.

3. Not Applicable 3. RCS Leak Rate 3. Primary Containment Isolation Failure 

Not Applicable Not Applicable A. UNISOLABLE 

break in ANY of 

the following: 

(site-specific 

systems with 

potential for highenergy line breaks) 

OR

B. Emergency RPV 

Depressurization. 

A. UNISOLABLE 

primary system 

leakage that 

results in 

exceeding 

EITHER of the 

following: 

1. Max Normal 

Operating 

Temperature

OR

2. Max Normal 

Operating Area 

Radiation 

Level.

A. UNISOLABLE 

direct downstream 

pathway to the 

environment exists 

after primary 

containment 

isolation signal

OR

B. Intentional primary 

containment 

venting per EOPs 

OR

C. UNISOLABLE 

primary system 

leakage that results 

in exceeding 

EITHER of the 

following: 

1. Max Safe 

Operating 

Temperature.

OR

2. Max Safe 

Operating Area 

Radiation 

Level.

Not Applicable 
NEI 99-01 (Revision 6)

November 2012

83

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. Primary 

containment 

radiation monitor 

reading greater 

than (site-specific 

value).

Not Applicable A. Primary 

containment 

radiation monitor 

reading greater 

than (site-specific 

value).

Not Applicable Not Applicable A. Primary 

containment 

radiation monitor 

reading greater 

than (site-specific 

value).

5. Other Indications 5. Other Indications 5. Other Indications

A. (site-specific as 

applicable)

A. (site-specific as 

applicable)

A. (site-specific as 

applicable)

A. (site-specific as 

applicable)

A. (site-specific as 

applicable)

A. (site-specific as 

applicable)

6. Emergency Director Judgment 6. Emergency Director Judgment 6. Emergency Director Judgment

A. ANY condition in 

the opinion of the 

Emergency 

Director that 

indicates Loss of 

the Fuel Clad 

Barrier.

A. ANY condition in 

the opinion of the 

Emergency 

Director that 

indicates Potential 

Loss of the Fuel 

Clad Barrier.

A. ANY condition in 

the opinion of the 

Emergency 

Director that 

indicates Loss of 

the RCS Barrier.

A. ANY condition in 

the opinion of the 

Emergency 

Director that 

indicates Potential 

Loss of the RCS 

Barrier.

A. ANY condition in 

the opinion of the 

Emergency 

Director that 

indicates Loss of 

the Containment 

Barrier.

A. ANY condition in 

the opinion of the 

Emergency 

Director that 

indicates Potential 

Loss of the 

Containment 

Barrier.
NEI 99-01 (Revision 6)

November 2012

84 

Basis Information For 

BWR EAL Fission Product Barrier Table 9-F-2 

BWR 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 1.A 

This threshold indicates that RCS radioactivity concentration is greater than 300 μCi/gm 

dose equivalent I-131. Reactor coolant activity above this level is greater than that 

expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel 

clad damage. Since this condition indicates that a significant amount of fuel clad damage 

has occurred, it represents a loss of the Fuel Clad Barrier.

There is no Potential Loss threshold associated with RCS Activity. 

Developer Notes:

Threshold values should be determined assuming RCS radioactivity concentration equals 

300 μCi/gm dose equivalent I-131. Other site-specific units may be used (e.g., μCi/cc). 

Depending upon site-specific capabilities, this threshold may have a sample analysis 

component and/or a radiation monitor reading component. 

Add this paragraph (or similar wording) to the Basis if the threshold includes a sample 

analysis component, “It is recognized that sample collection and analysis of reactor 

coolant with highly elevated activity levels could require several hours to complete. 

Nonetheless, a sample-related threshold is included as a backup to other indications.” 

2. RPV Water Level

Loss 2.A 

The Loss threshold represents the EOP requirement for primary containment flooding. 

This is identified in the BWROG EPGs/SAGs when the phrase, “Primary Containment 

Flooding Is Required," appears. Since a site-specific RPV water level is not specified 

here, the Loss threshold phrase, “Primary containment flooding required,” also 

accommodates the EOP need to flood the primary containment when RPV water level 

cannot be determined and core damage due to inadequate core cooling is believed to be 

occurring.

Potential Loss 2.A 

This water level corresponds to the top of the active fuel and is used in the EOPs to 

indicate a challenge to core cooling.
NEI 99-01 (Revision 6)

November 2012

85

BWR FUEL CLAD BARRIER THRESHOLDS:

The RPV water level threshold is the same as RCS barrier Loss threshold 2.A. Thus, this 

threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of the RCS barrier 

that appropriately escalates the emergency classification level to a Site Area Emergency. 

This threshold is considered to be exceeded when, as specified in the site-specific 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 of RPV water level until pressure drops below the shutoff heads 

of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only 

after either: 1) the RPV has been depressurized, or required emergency RPV 

depressurization has been attempted, giving the operator an opportunity to assess the 

capability of low-pressure injection sources to restore RPV water level or 2) no low 

pressure RPV injection systems are available, precluding RPV depressurization in an 

attempt to minimize loss 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 SA5 or SS5 will dictate the need for 

emergency classification.

Since the loss of ability to determine if adequate core cooling is being provided presents a 

significant challenge to the fuel clad barrier, a potential loss of the fuel clad barrier is 

specified.
NEI 99-01 (Revision 6)

November 2012

86 

BWR FUEL CLAD BARRIER THRESHOLDS:

Developer Notes:

Loss 2.A 

The phrase, “Primary containment flooding required,” should be modified to agree with 

the site-specific EOP phrase indicating exit from all EOPs and entry to the SAGs (e.g., 

drywell flooding required, etc.). 

Potential Loss 2.A 

The decision that "RPV water level cannot be determined" is directed by guidance given 

in the RPV water level control sections of the EOPs. 

3. Not Applicable (included for numbering consistency between barrier tables)

4. Primary Containment Radiation 

Loss 4.A 

The radiation monitor reading corresponds to an instantaneous release of all reactor 

coolant mass into the primary containment, assuming that reactor coolant activity equals 

300 μCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater 

than that expected for iodine spikes and corresponds to an approximate range of 2% to 

5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad 

damage has occurred, it represents a loss of the Fuel Clad Barrier. 

The radiation monitor reading in this threshold is higher than that specified for RCS 

Barrier Loss threshold 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. 

Developer Notes:

The reading should be determined assuming the instantaneous release and dispersal of the 

reactor coolant noble gas and iodine inventory, with RCS radioactivity concentration 

equal to 300 μCi/gm dose equivalent I-131, into the primary containment atmosphere. 
NEI 99-01 (Revision 6)

November 2012

87

BWR FUEL CLAD BARRIER THRESHOLDS:

5. Other Indications

Loss and/or Potential Loss 5.A 

This subcategory addresses other site-specific thresholds that may be included to indicate 

loss or potential loss of the Fuel Clad barrier based on plant-specific design 

characteristics not considered in the generic guidance. 

Developer Notes:

Loss and/or Potential Loss 5.A 

Developers should determine if other reliable indicators exist to evaluate the status of this 

fission product barrier (e.g., review accident analyses described in the site Final Safety 

Analysis Report, as updated). The goal is to identify any unique or site-specific 

indications that will promote timely and accurate assessment of barrier status. 

Any added thresholds should represent approximately the same relative threat to the 

barrier as the other thresholds in this column. Basis information for the other thresholds 

may be used to gauge the relative barrier threat level. 

6. Emergency Director Judgment 

Loss 6.A 

This threshold addresses any other factors that are to be used by the Emergency Director 

in determining whether the 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.

Developer Notes:

None 
NEI 99-01 (Revision 6)

November 2012

88 

BWR 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 Pressure

Loss 1.A 

The (site-specific value) primary containment pressure is the drywell high pressure 

setpoint which indicates a LOCA by automatically initiating the ECCS or equivalent 

makeup system.

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

Developer Notes:

None 

2. RPV Water Level

Loss 2.A 

This water level corresponds to the top of active fuel 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 site-specific 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 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. 
NEI 99-01 (Revision 6)

November 2012

89

BWR RCS BARRIER THRESHOLDS: 

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. 

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 SA5 or SS5 will dictate the need for 

emergency classification.

There is no RCS Potential Loss threshold associated with RPV Water Level.

3. RCS Leak Rate

Loss Threshold 3.A 

Large high-energy lines that rupture outside primary containment can discharge 

significant amounts of inventory and jeopardize the pressure-retaining capability of the 

RCS until they are isolated. If it is determined that the ruptured line cannot be promptly 

isolated from the Control Room, the RCS barrier Loss threshold is met.

Loss Threshold 3.B 

Emergency RPV Depressurization in accordance with the EOPs is indicative of a loss of 

the RCS barrier. If Emergency RPV Depressurization is performed, the plant operators 

are directed to open safety relief valves (SRVs) and keep them open. Even though the 

RCS is being vented into the suppression pool, a Loss of the RCS barrier exists due to the 

diminished effectiveness of the RCS to retain fission products within its boundary. 

Potential Loss Threshold 3.A 

Potential loss of RCS based on primary system leakage outside the primary containment 

is determined from EOP temperature or radiation Max Normal Operating values in areas 

such as main steam line tunnel, RCIC, HPCI, etc., which indicate a direct path from the 

RCS to areas outside primary containment.

A Max Normal Operating 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. 
NEI 99-01 (Revision 6)

November 2012

90 

BWR RCS BARRIER THRESHOLDS: 

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 Max Normal Operating 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.

Developer Notes:

Loss Threshold 3.A 

The list of systems included in this threshold should be the high energy lines which, if 

ruptured and remain unisolated, can rapidly depressurize the RPV. These lines are 

typically isolated by actuation of the Leak Detection system.

Large high-energy line breaks such as Main Steam Line (MSL), High Pressure Coolant 

Injection (HPCI), Feedwater, Reactor Water Cleanup (RWCU), Isolation Condenser (IC) 

or Reactor Core Isolation Cooling (RCIC) that are UNISOLABLE represent a significant 

loss of the RCS barrier.

4. Primary Containment Radiation

Loss 4.A 

The radiation monitor reading corresponds to an instantaneous release of all reactor 

coolant mass into the primary containment, assuming that reactor coolant activity equals

Technical Specification allowable limits. This value is lower than that specified for Fuel 

Clad Barrier Loss threshold 4.A since it indicates a loss of the RCS Barrier only. 

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

Developer Notes:

The reading should be determined assuming the instantaneous release and dispersal of the 

reactor coolant noble gas and iodine inventory, with RCS activity at Technical 

Specification allowable limits, into the primary containment atmosphere. Using RCS 

activity at Technical Specification allowable limits aligns this threshold with IC SU3. 

Also, RCS activity at this level will typically result in primary containment radiation 

levels that can be more readily detected by primary containment radiation monitors, and 

more readily differentiated from those caused by piping or component “shine” sources. If 

desired, a plant may use a lesser value of RCS activity for determining this value. 
NEI 99-01 (Revision 6)

November 2012

91

BWR RCS BARRIER THRESHOLDS: 

In some cases, the site-specific physical location and sensitivity of the primary 

containment radiation monitor(s) may be such that radiation from a cloud of released 

RCS gases cannot be distinguished from radiation emanating from piping and 

components containing elevated reactor coolant activity. If so, refer to the Developer 

Guidance for Loss/Potential Loss 5.A and determine if an alternate indication is 

available.

5. Other Indications

Loss and/or Potential Loss 5.A 

This subcategory addresses other site-specific thresholds that may be included to indicate 

loss or potential loss of the RCS barrier based on plant-specific design characteristics not 

considered in the generic guidance. 

Developer Notes:

Loss and/or Potential Loss 5.A 

Developers should determine if other reliable indicators exist to evaluate the status of this 

fission product barrier (e.g., review accident analyses described in the site Final Safety 

Analysis Report, as updated). The goal is to identify any unique or site-specific 

indications that will promote timely and accurate assessment of barrier status. 

Any added thresholds should represent approximately the same relative threat to the 

barrier as the other thresholds in this column. Basis information for the other thresholds 

may be used to gauge the relative barrier threat level. 

6. Emergency Director Judgment

Loss 6.A 

This threshold addresses any other factors that are to be used by the Emergency Director 

in determining whether the RCS barrier is lost. 

Potential Loss 6.A 

This threshold addresses any other factors that may be used by the Emergency Director in 

determining whether the RCS Barrier is potentially lost. The Emergency Director should 

also consider whether or not to declare the barrier potentially lost in the event that barrier 

status cannot be monitored.

Developer Notes:

None 
NEI 99-01 (Revision 6)

November 2012

92 

BWR 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 1.A and 1.B 

Rapid UNPLANNED loss of primary containment pressure (i.e., not attributable to 

drywell spray or condensation effects) following an initial pressure increase indicates a 

loss of primary containment integrity. Primary containment pressure should increase as a 

result of mass and energy release into the primary containment from a LOCA. Thus, 

primary containment 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. 

Potential Loss 1.A 

The threshold pressure is the primary containment 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 1.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 Temperature Limit (HCTL) 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
NEI 99-01 (Revision 6)

November 2012

93

BWR CONTAINMENT BARRIER THRESHOLDS:

 Suppression chamber pressure above Primary Containment Pressure Limit A, while 

the rate of energy transfer from the RPV to the containment is greater than the 

capacity of the containment vent.

The HCTL is a function 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.

Developer Notes:

Potential Loss 1.B 

BWR EPGs/SAGs specifically define the limits associated with explosive mixtures in 

terms of deflagration concentrations of hydrogen and oxygen. For Mk I/II containments 

the deflagration limits are “6% hydrogen and 5% oxygen in the drywell or suppression 

chamber”. For Mk III containments, the limit is the “Hydrogen Deflagration 

Overpressure Limit”. The threshold term “explosive mixture” is synonymous with the 

EPG/SAG “deflagration limits”.

Potential Loss 1.C 

Since the HCTL is defined assuming a range of suppression pool water levels as low as 

the elevation of the downcomer openings in Mk I/II containments, or 2 feet above the 

elevation of the horizontal vents in a Mk III containment, it is unnecessary to consider 

separate Containment barrier Loss or Potential Loss thresholds for abnormal suppression 

pool water level conditions. If desired, developers may include a separate Containment 

Potential Loss threshold based on the inability to maintain suppression pool water level 

above the downcomer openings in Mk I/II containments, or 2 feet above the elevation of 

the horizontal vents in a Mk III containment with RPV pressure above the minimum 

decay heat removal pressure, if it will simplify the assessment of the suppression pool 

level component of the HCTL. 

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 SAGs 

are entered. Entry into SAGs is a logical escalation in response to the inability to restore 

and maintain adequate core cooling.
NEI 99-01 (Revision 6)

November 2012

94 

BWR CONTAINMENT BARRIER THRESHOLDS:

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.

Developer Notes:

The phrase, “Primary containment flooding required,” should be modified to agree with 

the site-specific EOP phrase indicating exit from all EOPs and entry to the SAGs (e.g., 

drywell flooding required, etc.). 

3. Primary Containment Isolation Failure 

These thresholds address incomplete containment isolation that allows an UNISOLABLE 

direct release to the environment.

Loss 3.A 

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 A ICs.

Loss 3.B 

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. 
NEI 99-01 (Revision 6)

November 2012

95

Loss 3.C 

The Max Safe Operating Temperature and the Max Safe Operating 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. 

BWR CONTAINMENT BARRIER THRESHOLDS:

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 Primary Containment Isolation 

Failure. 

Developer Notes:

Loss 3.B 

Consideration may be given to specifying the specific procedural step within the Primary 

Containment Control EOP that defines intentional venting of the Primary Containment 

regardless of offsite radioactivity release rate. 

4. Primary Containment Radiation 

There is no Loss threshold associated with Primary Containment Radiation. 

Potential Loss 4.A 

The radiation monitor reading corresponds to an instantaneous release of all reactor 

coolant mass into the primary containment, assuming that 20% of the fuel cladding has 

failed. This level of fuel clad failure is well above that used to determine the analogous 

Fuel Clad 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. 
NEI 99-01 (Revision 6)

November 2012

96 

Developer Notes:

NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power 

Plant Accidents, provides the basis for using the 20% fuel cladding failure value. Unless 

there is a site-specific analysis justifying a different value, the reading should be 

determined assuming the instantaneous release and dispersal of the reactor coolant noble 

gas and iodine inventory associated with 20% fuel clad failure into the primary 

containment atmosphere. 

BWR CONTAINMENT BARRIER THRESHOLDS:

5. Other Indications

Loss and/or Potential Loss 5.A 

This subcategory addresses other site-specific thresholds that may be included to indicate 

loss or potential loss of the Containment barrier based on plant-specific design 

characteristics not considered in the generic guidance. 

Developer Notes:

Loss and/or Potential Loss 5.A 

Developers should determine if other reliable indicators exist to evaluate the status of this 

fission product barrier (e.g., review accident analyses described in the site Final Safety 

Analysis Report, as updated). The goal is to identify any unique or site-specific 

indications that will promote timely and accurate assessment of barrier status. 

Any added thresholds should represent approximately the same relative threat to the 

barrier as the other thresholds in this column. Basis information for the other thresholds 

may be used to gauge the relative barrier threat level. 

6. Emergency Director Judgment

Loss 6.A 

This threshold addresses any other factors that are to be used by the Emergency Director 

in determining whether the 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. 

Developer Notes:

None 
NEI 99-01 (Revision 6)

November 2012

97

Table 9-F-3: PWR EAL Fission Product Barrier Table

Thresholds for LOSS or POTENTIAL LOSS of Barriers

FA1 ALERT

Any Loss or any Potential Loss of either 

the Fuel Clad or RCS barrier.

FS1 SITE AREA EMERGENCY

Loss or Potential Loss of any two barriers.

FG1 GENERAL EMERGENCY

Loss of any two barriers and Loss or 

Potential Loss of the third barrier.

Fuel Clad Barrier RCS Barrier Containment Barrier

LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS

1. RCS or SG Tube Leakage 1. RCS or SG Tube Leakage 1. RCS or SG Tube Leakage

Not Applicable A. RCS/reactor 

vessel level less 

than (site-specific 

level).

A. An automatic or 

manual ECCS (SI) 

actuation is 

required by 

EITHER of the 

following: 

1. UNISOLABLE 

RCS leakage

OR

2. SG tube 

RUPTURE. 

A. Operation of a 

standby charging 

(makeup) pump is 

required by 

EITHER of the 

following: 

1. UNISOLABLE 

RCS leakage

OR

2. SG tube 

leakage.

OR

B. RCS cooldown 

rate greater than 

(site-specific 

pressurized 

thermal shock 

criteria/limits 

defined by sitespecific 

indications). 

A. A leaking or 

RUPTURED SG 

is FAULTED 

outside of 

containment. 

Not Applicable
NEI 99-01 (Revision 6)

November 2012

98 

Fuel Clad Barrier RCS Barrier Containment Barrier

LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS

2. Inadequate Heat Removal 2. Inadequate Heat Removal 2. Inadequate Heat Removal

A. Core exit 

thermocouple 

readings greater 

than (sitespecific 

temperature 

value).

A. Core exit 

thermocouple 

readings greater 

than (site-specific 

temperature 

value).

OR

B. Inadequate RCS 

heat removal 

capability via 

steam generators 

as indicated by 

(site-specific 

indications). 

Not Applicable A. Inadequate RCS 

heat removal 

capability via 

steam generators 

as indicated by 

(site-specific 

indications). 

Not Applicable A. 1. (Site-specific 

criteria for 

entry into core 

cooling 

restoration 

procedure) 

 AND

 2. Restoration 

procedure not 

effective within 

15 minutes. 

3. RCS Activity / Containment Radiation 3. RCS Activity / Containment Radiation 3. RCS Activity / Containment Radiation

A. Containment 

radiation monitor 

reading greater 

than (site-specific 

value).

OR

B. (Site-specific 

indications that 

reactor coolant 

activity is greater 

than 300 μCi/gm 

dose equivalent I131).

Not Applicable A. Containment 

radiation monitor 

reading greater 

than (site-specific 

value).

Not Applicable Not Applicable A. Containment 

radiation monitor 

reading greater 

than (site-specific 

value).
NEI 99-01 (Revision 6)

November 2012

99

Fuel Clad Barrier RCS Barrier Containment Barrier

LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS

4. Containment Integrity or Bypass 4. Containment Integrity or Bypass 4. Containment Integrity or Bypass

Not Applicable Not Applicable Not Applicable Not Applicable A. Containment 

isolation is 

required

AND 

EITHER of the 

following:

1. Containment 

integrity has 

been lost based 

on Emergency 

Director 

judgment. 

OR

2. UNISOLABLE 

pathway from 

the containment 

to the 

environment 

exists. 

OR

B. Indications of RCS 

leakage outside of 

containment. 

A. Containment 

pressure greater 

than (site-specific 

value)

OR

B. Explosive mixture 

exists inside 

containment 

OR

C. 1. Containment 

pressure greater 

than (sitespecific 

pressure

setpoint) 

AND

2. Less than one 

full train of 

(site-specific 

system or 

equipment) is 

operating per 

design for 15 

minutes or 

longer.
NEI 99-01 (Revision 6)

November 2012

100 

Fuel Clad Barrier RCS Barrier Containment Barrier

LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS

5. Other Indications 5. Other Indications 5. Other Indications

A. (site-specific as 

applicable)

A. (site-specific as 

applicable)

A. (site-specific as 

applicable)

A. (site-specific as 

applicable)

A. (site-specific as 

applicable)

A. (site-specific as 

applicable)

6. Emergency Director Judgment 6. Emergency Director Judgment 6. Emergency Director Judgment

A. ANY condition 

in the opinion of 

the Emergency 

Director that 

indicates Loss of 

the Fuel Clad 

Barrier.

A. ANY condition in 

the opinion of the 

Emergency 

Director that 

indicates 

Potential Loss of 

the Fuel Clad 

Barrier.

A. ANY condition in 

the opinion of the 

Emergency 

Director that 

indicates Loss of 

the RCS Barrier.

A. ANY condition in 

the opinion of the 

Emergency 

Director that 

indicates 

Potential Loss of 

the RCS Barrier.

A. ANY condition in 

the opinion of the 

Emergency 

Director that 

indicates Loss of 

the Containment 

Barrier.

A. ANY condition in 

the opinion of the 

Emergency Director 

that indicates 

Potential Loss of 

the Containment 

Barrier.
NEI 99-01 (Revision 6)

November 2012

101

Basis Information For

PWR EAL Fission Product Barrier Table 9-F-3 

Developer Notes:

Threshold Parameters and Values

Each PWR owner’s group has developed a methodology for guiding the development and 

implementation of EOPs (i.e., assessing plant parameters, and determining and prioritizing 

operator actions). Many of the thresholds contained in the PWR EAL Fission Product Barrier 

Table reflect conditions that are specifically addressed in EOPs (e.g., a loss of heat removal 

capability by the steam generators). When developing a site-specific threshold, developers 

should use the parameters and values specified within their EOPs that align with the condition 

described by the generic threshold and basis, and related developer notes. This approach will 

ensure consistency between the site-specific EOPs and emergency classification scheme, and 

thus facilitate more timely and accurate classification assessments.

In support of EOP development and implementation, the Westinghouse Owners Group (WOG) 

developed a defined set of Critical Safety Functions as part of their Emergency Response 

Guidelines. The WOG approach structures EOPs to maintain and/or restore these Critical Safety 

Functions, and to do so in a prioritized and systematic manner. The WOG Critical Safety 

Functions are presented below.

 Subcriticality

 Core Cooling 

 Heat Sink

 RCS Integrity

 Containment 

 RCS Inventory 

The WOG ERGs provide a methodology for monitoring the status of the Critical Safety 

Functions and classifying the significance of a challenge to a function; this methodology is 

referred to as the Critical Safety Function Status Trees (CSFSTs). For plants that have 

implemented the WOG ERGs, the guidance in NEI 99-01 allows for use of certain CSFST 

assessment results as EALs and fission product barrier loss/potential loss thresholds. In this 

manner, an emergency classification assessment may flow directly from a CSFST assessment.

It is important to understand that the CSFSTs are evaluated using plant parameters, and that they 

are simply a vendor-specific method for collectively evaluating a set of parameters for purposes 

of driving emergency operating procedure usage. For the emergency conditions of interest, the 

generic thresholds within the PWR EAL Fission Product Barrier Table specify the plant 

parameters that define a potential loss or loss of a fission product barrier; however, as described 

in the associated Developer Notes, a CSFST terminus may be used as well. For this reason, 

inclusion of the CSFST-related thresholds would be redundant to the parameter-based thresholds 

for plants that employ the WOG ERGs. 

Sites that employ the WOG ERGs may, at their discretion, include the CSFST-based loss and 

potential loss thresholds as described in the Developer Notes. Developers at these sites should 

consult with their classification decision-makers to determine if inclusion would assist with 
NEI 99-01 (Revision 6)

November 2012

102 

timely and accurate emergency classification. This decision should consider the effects of any 

site-specific changes to the generic WOG CSFST evaluation logic and setpoints, as well as those 

arising from user rules applicable to emergency operating procedures (e.g., exceptions to 

procedure entry or transition due to specific accident conditions or loss of a support system).

The CSFST thresholds may be addressed in one of 3 ways: 

1) Not incorporated; thresholds will use parameters and values as discussed in the Developer 

Notes. 

2) Incorporated along with parameter and value thresholds (e.g., a fuel clad loss would have 2 

thresholds such as “CETs > 1200o

F” and “Core Cooling Red entry conditions met”.

3) Used in lieu of parameters and values for all thresholds. 

With one exception, if a decision is made to include the CSFST-based thresholds, then all such 

allowed thresholds must be used in the table (e.g., it is not permissible to use only the C Orange 

terminus as a potential loss of the fuel clad barrier threshold and disregard all other CSFSTbased thresholds). The one exception is the RCS Integrity (P) CSFST. Because of the 

complexity of the P Red decision-point that relies on an assessment a pressure-temperature 

curve, a P Red condition may be used as an RCS potential loss threshold without the need to 

incorporate the other CSFST-based thresholds. 
NEI 99-01 (Revision 6)

November 2012

103

PWR FUEL CLAD BARRIER THRESHOLDS:

The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.

1. RCS or SG Tube Leakage

There is no Loss threshold associated with RCS or SG Tube Leakage. 

Potential Loss 1.A 

This reading indicates a reduction in reactor vessel water level sufficient to allow the 

onset of heat-induced cladding damage. 

Developer Notes:

Potential Loss 1.A 

Enter the site-specific reactor vessel water level value(s) used by EOPs to identify a 

degraded core cooling condition (e.g., requires prompt restoration action). The reactor 

vessel level that corresponds to approximately the top of active fuel may also be used. 

For plants that have implemented Westinghouse Owners Group Emergency Response 

Guidelines, enter the reactor vessel level(s) used for the Core Cooling Orange Path

(including dependencies upon the status of RCPs, if applicable). 

Westinghouse ERG Plants 

Developers should consider including a threshold the same as, or similar to, “Core 

Cooling Orange entry conditions met” in accordance with the guidance at the front of this 

section. 

2. Inadequate Heat Removal 

Loss 2.A 

This reading indicates temperatures within the core are sufficient to cause significant 

superheating of reactor coolant. 

Potential Loss 2.A 

This reading indicates temperatures within the core are sufficient to allow the onset of 

heat-induced cladding damage. 

Potential Loss 2.B 

This condition indicates an extreme challenge to the ability to remove RCS heat using the 

steam generators (i.e., loss of an effective secondary-side heat sink). This condition 

represents a potential loss of the Fuel Clad Barrier. In accordance with EOPs, there may 

be unusual accident conditions during which operators intentionally reduce the heat 
NEI 99-01 (Revision 6)

November 2012

104 

removal capability of the steam generators; during these conditions, classification using 

threshold is not warranted. 

PWR FUEL CLAD BARRIER THRESHOLDS:

Meeting this threshold results in a Site Area Emergency because this threshold is 

identical to RCS Barrier Potential Loss threshold 2.A; both will be met. This condition 

warrants a Site Area Emergency declaration because inadequate RCS heat removal may 

result in fuel heat-up sufficient to damage the cladding and increase RCS pressure to the 

point where mass will be lost from the system. 

Developer Notes:

Some site-specific EOPs and/or EOP user guidelines may establish decision-making 

criteria concerning the number or other attributes of thermocouple readings necessary to 

drive actions (e.g., 5 CETs reading greater than 1,200o

F is required before transitioning to 

an inadequate core cooling procedure). To maintain consistency with EOPs, these 

decision-making criteria may be used in the core exit thermocouple reading thresholds. 

Loss 2.A 

Enter a site-specific temperature value that corresponds to significant in-core 

superheating of reactor coolant. 1,200o

F may also be used. 

For plants that have implemented Westinghouse Owners Group Emergency Response 

Guidelines, enter the parameters and values used in the Core Cooling Red Path. 

Potential Loss 2.A 

Enter a site-specific temperature value that corresponds to core conditions at the onset of 

heat-induced cladding damage (e.g., the temperature allowing for the formation of 

superheated steam assuming that the RCS is intact). 700oF may also be used. 

For plants that have implemented Westinghouse Owners Group Emergency Response 

Guidelines, enter the parameters and values used in the Core Cooling Orange Path. 

Potential Loss 2.B 

Enter the site-specific parameters and values that define an extreme challenge to the 

ability to remove heat from the RCS via the steam generators. These will typically be 

parameters and values that would require operators to take prompt action to address this 

condition. 

For plants that have implemented Westinghouse Owners Group Emergency Response 

Guidelines, enter the parameters and values used in the Heat Sink Red Path. 
NEI 99-01 (Revision 6)

November 2012

105

Westinghouse ERG Plants 

As a loss indication, developers should consider including a threshold the same as, or 

similar to, “Core Cooling Red entry conditions met” in accordance with the guidance at 

the front of this section. 

PWR FUEL CLAD BARRIER THRESHOLDS:

As a potential loss indication, developers should consider including a threshold the same 

as, or similar to, “Core Cooling Orange entry conditions met” in accordance with the 

guidance at the front of this section. 

As a potential loss indication, developers should consider including a threshold the same 

as, or similar to, “Heat Sink Red entry conditions met” in accordance with the guidance

at the front of this section. 

3. RCS Activity / Containment Radiation

Loss 3.A 

The radiation monitor reading corresponds to an instantaneous release of all reactor 

coolant mass into the containment, assuming that reactor coolant activity equals 

300μCi/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 reading in this threshold is higher than that specified for RCS 

Barrier Loss threshold 3.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. 

Loss 3.B 

This threshold indicates that RCS radioactivity concentration is greater than 300 μCi/gm 

dose equivalent I-131. Reactor coolant activity above this level is greater than that 

expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel 

clad damage. Since this condition indicates that a significant amount of fuel clad damage 

has occurred, it represents a loss of the Fuel Clad Barrier. 

There is no Potential Loss threshold associated with RCS Activity / Containment 

Radiation. 

Developer Notes:

Loss 3.A 
NEI 99-01 (Revision 6)

November 2012

106 

The reading should be determined assuming the instantaneous release and dispersal of the 

reactor coolant noble gas and iodine inventory, with RCS radioactivity concentration 

equal to 300 μCi/gm dose equivalent I-131, into the containment atmosphere. 

PWR FUEL CLAD BARRIER THRESHOLDS:

Loss 3.B 

Threshold values should be determined assuming RCS radioactivity concentration equals 

300 μCi/gm dose equivalent I-131. Other site-specific units may be used (e.g., μCi/cc). 

Depending upon site-specific capabilities, this threshold may have a sample analysis 

component and/or a radiation monitor reading component. 

Add this paragraph (or similar wording) to the Basis if the threshold includes a sample 

analysis component, “It is recognized that sample collection and analysis of reactor 

coolant with highly elevated activity levels could require several hours to complete. 

Nonetheless, a sample-related threshold is included as a backup to other indications.” 

4. Containment Integrity or Bypass 

Not Applicable (included for numbering consistency) 

5. Other Indications

Loss and/or Potential Loss 5.A 

This subcategory addresses other site-specific thresholds that may be included to indicate 

loss or potential loss of the Fuel Clad barrier based on plant-specific design 

characteristics not considered in the generic guidance. 

Developer Notes:

Loss and/or Potential Loss 5.A 

Developers should determine if other reliable indicators exist to evaluate the status of this 

fission product barrier (e.g., review accident analyses described in the site Final Safety 

Analysis Report, as updated). The goal is to identify any unique or site-specific 

indications that will promote timely and accurate assessment of barrier status. 

Any added thresholds should represent approximately the same relative threat to the 

barrier as the other thresholds in this column. Basis information for the other thresholds 

may be used to gauge the relative barrier threat level. 
NEI 99-01 (Revision 6)

November 2012

107

6. Emergency Director Judgment

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 lost. 

PWR FUEL CLAD BARRIER THRESHOLDS:

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. 

Developer Notes:

None 
NEI 99-01 (Revision 6)

November 2012

108 

PWR RCS BARRIER THRESHOLDS:

The RCS Barrier includes the RCS primary side and its connections up to and including the 

pressurizer safety and relief valves, and other connections up to and including the primary 

isolation valves. 

1. RCS or SG Tube Leakage

Loss 1.A 

This threshold is based on an UNISOLABLE RCS leak of sufficient size to require an 

automatic or manual actuation of the Emergency Core Cooling System (ECCS). This 

condition clearly represents a loss of the RCS Barrier. 

This threshold is applicable to unidentified and pressure boundary leakage, as well as 

identified leakage. It is also applicable to UNISOLABLE RCS leakage through an 

interfacing system. The mass loss may be into any location – inside containment, to the 

secondary-side (i.e., steam generator tube leakage) or outside of containment. 

A steam generator with primary-to-secondary leakage of sufficient magnitude to require a 

safety injection is considered to be RUPTURED. If a RUPTURED steam generator is 

also FAULTED outside of containment, the declaration escalates to a Site Area 

Emergency since the Containment Barrier Loss threshold 1.A will also be met.

Potential Loss 1.A 

This threshold is based on an UNISOLABLE RCS leak that results in the inability to 

maintain pressurizer level within specified limits by operation of a normally used 

charging (makeup) pump, but an ECCS (SI) actuation has not occurred. The threshold is 

met when an operating procedure, or operating crew supervision, directs that a standby 

charging (makeup) pump be placed in service to restore and maintain pressurizer level. 

This threshold is applicable to unidentified and pressure boundary leakage, as well as 

identified leakage. It is also applicable to UNISOLABLE RCS leakage through an 

interfacing system. The mass loss may be into any location – inside containment, to the 

secondary-side (i.e., steam generator tube leakage) or outside of containment. 

If a leaking steam generator is also FAULTED outside of containment, the declaration 

escalates to a Site Area Emergency since the Containment Barrier Loss threshold 1.A will 

also be met.

Potential Loss 1.B 

This condition indicates an extreme challenge to the integrity of the RCS pressure 

boundary due to pressurized thermal shock – a transient that causes rapid RCS cooldown 

while the RCS is in Mode 3 or higher (i.e., hot and pressurized). 
NEI 99-01 (Revision 6)

November 2012

109

PWR RCS BARRIER THRESHOLDS:

Developer Notes:

Loss 1.A 

Actuation of the ECCS may also be referred to as Safety Injection (SI) actuation or other 

appropriate site-specific term.

Potential Loss 1.A 

Depending upon charging pump flow capacities and RCS volume control parameters, 

developers may use an RCS leak rate value of 50 gpm, or an appropriate site-specific 

value, as an alternate Potential Loss threshold. If used, the threshold wording should 

reflect that the determination of the leak rate value excludes normal reductions in RCS 

inventory (e.g., by the letdown system or RCP seal leakoff).

Potential Loss 1.B 

Enter the site-specific indications that define an extreme challenge to the integrity of the 

RCS pressure boundary due to pressurized thermal shock – a transient that causes rapid 

RCS cooldown while the RCS is in Mode 3 or higher (i.e., hot and pressurized). These 

will typically be parameters and values that would require operators to take prompt action 

to address a pressurized thermal shock condition. Developers should also determine if 

the threshold needs to reflect any dependencies used as EOP transition/entry decision 

points or condition validation criteria (e.g., an EOP used to respond to an excessive RCS 

cooldown may not be entered or immediately exited if RCS pressure is below a certain 

value).

For plants that have implemented Westinghouse Owners Group Emergency Response 

Guidelines, enter the parameters and values used in the RCS Integrity Red Path. Because 

of the complexity of certain decision-points within the Red Path of this CSFST, 

developers at these plants may elect to not include the specific parameters and values, 

and instead follow the guidance below. 

Westinghouse ERG Plants 

As a potential loss indication, developers should consider including a threshold the same 

as, or similar to, “RCS Integrity Red entry conditions met” in accordance with the 

guidance at the front of this section. As noted above, developers should ensure that the 

threshold wording reflects any EOP transition/entry decision points or condition 

validation criteria. For example, a threshold might read “RCS Integrity (P) Red entry 

conditions met with RCS pressure > 300 psig.” 

2. Inadequate Heat Removal 

There is no Loss threshold associated with Inadequate Heat Removal. 
NEI 99-01 (Revision 6)

November 2012

110 

PWR RCS BARRIER THRESHOLDS:

Potential Loss 2.A 

This condition indicates an extreme challenge to the ability to remove RCS heat using the 

steam generators (i.e., loss of an effective secondary-side heat sink). This condition 

represents a potential loss of the RCS Barrier. In accordance with EOPs, there may be 

unusual accident conditions during which operators intentionally reduce the heat removal 

capability of the steam generators; during these conditions, classification using threshold 

is not warranted. 

Meeting this threshold results in a Site Area Emergency because this threshold is 

identical to Fuel Clad Barrier Potential Loss threshold 2.B; both will be met. This 

condition warrants a Site Area Emergency declaration because inadequate RCS heat 

removal may result in fuel heat-up sufficient to damage the cladding and increase RCS 

pressure to the point where mass will be lost from the system. 

Developer Notes:

Potential Loss 2.A 

Enter the site-specific parameters and values that define an extreme challenge to the 

ability to remove heat from the RCS via the steam generators. These will typically be 

parameters and values that would require operators to take prompt action to address this 

condition. 

For plants that have implemented Westinghouse Owners Group Emergency Response 

Guidelines, enter the parameters and values used in the Heat Sink Red Path. 

Westinghouse ERG Plants 

Developers should consider including a threshold the same as, or similar to, “Heat Sink 

Red entry conditions met” in accordance with the guidance at the front of this section. 

3. RCS Activity / Containment Radiation 

Loss 3.A 

The radiation monitor reading corresponds to an instantaneous release of all reactor 

coolant mass into the containment, assuming that reactor coolant activity equals

Technical Specification allowable limits. This value is lower than that specified for Fuel 

Clad Barrier Loss threshold 3.A since it indicates a loss of the RCS Barrier only.

There is no Potential Loss threshold associated with RCS Activity / Containment 

Radiation. 
NEI 99-01 (Revision 6)

November 2012

111

PWR RCS BARRIER THRESHOLDS:

Developer Notes:

Loss 3.A 

The reading should be determined assuming the instantaneous release and dispersal of the 

reactor coolant noble gas and iodine inventory, with RCS activity at Technical 

Specification allowable limits, into the containment atmosphere. Using RCS activity at 

Technical Specification allowable limits aligns this threshold with IC SU3. Also, RCS 

activity at this level will typically result in containment radiation levels that can be more 

readily detected by containment radiation monitors, and more readily differentiated from 

those caused by piping or component “shine” sources. If desired, a plant may use a lesser 

value of RCS activity for determining this value. 

In some cases, the site-specific physical location and sensitivity of the containment 

radiation monitor(s) may be such that radiation from a cloud of released RCS gases 

cannot be distinguished from radiation emanating from piping and components 

containing elevated reactor coolant activity. If so, refer to the Developer Notes for 

Loss/Potential Loss 5.A and determine if an alternate indication is available.

4. Containment Integrity or Bypass 

Not Applicable (included for numbering consistency) 

5. Other Indications

Loss and/or Potential Loss 5.A 

This subcategory addresses other site-specific thresholds that may be included to indicate 

loss or potential loss of the RCS barrier based on plant-specific design characteristics not 

considered in the generic guidance. 

Developer Notes:

Loss and/or Potential Loss 5.A 

Developers should determine if other reliable indicators exist to evaluate the status of this 

fission product barrier (e.g., review accident analyses described in the site Final Safety 

Analysis Report, as updated). The goal is to identify any unique or site-specific 

indications that will promote timely and accurate assessment of barrier status. 

Any added thresholds should represent approximately the same relative threat to the 

barrier as the other thresholds in this column. Basis information for the other thresholds 

may be used to gauge the relative barrier threat level. 
NEI 99-01 (Revision 6)

November 2012

112 

PWR RCS BARRIER THRESHOLDS:

6. Emergency Director Judgment

Loss 6.A 

This threshold addresses any other factors that may 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. 

Developer Notes:

None 
NEI 99-01 (Revision 6)

November 2012

113

PWR CONTAINMENT BARRIER THRESHOLDS:

The Containment Barrier includes the containment building and connections up to and including 

the outermost containment isolation valves. This barrier also includes the main steam, feedwater, 

and blowdown line extensions outside the containment building up to and including the 

outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for 

escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.

1. RCS or SG Tube Leakage

Loss 1.A 

This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also 

FAULTED outside of containment. The condition of the SG, whether leaking or 

RUPTURED, is determined in accordance with the thresholds for RCS Barrier Potential 

Loss 1.A and Loss 1.A, respectively. This condition represents a bypass of the 

containment barrier. 

FAULTED is a defined term within the NEI 99-01 methodology; this determination is 

not necessarily dependent upon entry into, or diagnostic steps within, an EOP. For 

example, if the pressure in a steam generator is decreasing uncontrollably [part of the 

FAULTED definition] and the faulted steam generator isolation procedure is not entered 

because EOP user rules are dictating implementation of another procedure to address a 

higher priority condition, the steam generator is still considered FAULTED for 

emergency classification purposes.

The FAULTED criterion establishes an appropriate lower bound on the size of a steam 

release that may require an emergency classification. Steam releases of this size are 

readily observable with normal Control Room indications. The lower bound for this 

aspect of the containment barrier is analogous to the lower bound criteria specified in IC 

SU3 for the fuel clad barrier (i.e., RCS activity values) and IC SU4 for the RCS barrier 

(i.e., RCS leak rate values).

 

This threshold also applies to prolonged steam releases necessitated by operational 

considerations such as the forced steaming of a leaking or RUPTURED steam generator 

directly to atmosphere to cooldown the plant, or to drive an auxiliary (emergency) feed 

water pump. These types of conditions will result in a significant and sustained release of

radioactive steam to the environment (and are thus similar to a FAULTED condition). 

The inability to isolate the steam flow without an adverse effect on plant cooldown meets 

the intent of a loss of containment. 

Steam releases associated with the expected operation of a SG power operated relief 

valve or safety relief valve do not meet the intent of this threshold. Such releases may 

occur intermittently for a short period of time following a reactor trip as operators process 

through emergency operating procedures to bring the plant to a stable condition and 

prepare to initiate a plant cooldown. Steam releases associated with the unexpected 

operation of a valve (e.g., a stuck-open safety valve) do meet this threshold. 
NEI 99-01 (Revision 6)

November 2012

114 

PWR CONTAINMENT BARRIER THRESHOLDS:

Following an SG tube leak or rupture, there may be minor radiological releases through a 

secondary-side system component (e.g., air ejectors, glad seal exhausters, valve packing, 

etc.). These types of releases do not constitute a loss or potential loss of containment but 

should be evaluated using the Recognition Category A ICs. 

The emergency classification levels resulting from primary-to-secondary leakage, with or 

without a steam release from the FAULTED SG, are summarized below.

Affected SG is FAULTED 

Outside of Containment?

P-to-S Leak Rate Yes No

Less than or equal to 25 gpm 

(or other value per SU4 

Developer Notes)

No classification No classification

Greater than 25 gpm (or other 

value per SU4 Developer 

Notes)

Unusual Event per SU4 Unusual Event per SU4 

Requires operation of a 

standby charging (makeup) 

pump (RCS Barrier Potential 

Loss)

Site Area Emergency 

per FS1 Alert per FA1

Requires an automatic or 

manual ECCS (SI) actuation 

(RCS Barrier Loss)

Site Area Emergency 

per FS1 Alert per FA1

There is no Potential Loss threshold associated with RCS or SG Tube Leakage. 

Developer Notes:

Loss 1.A 

A steam generator power operated relief valve may also be referred to as an atmospheric 

steam dump valve or other appropriate site-specific term.

Developers may include an additional site-specific threshold(s) to address prolonged 

steam releases necessitated by operational considerations if AOPs or EOPs could require 

that a leaking or RUPTURED steam generator be used to support plant cooldown. 

Developers may wish to consider incorporating the above table into user aids (e.g., a 

wallboard) or other locations within their basis document. 
NEI 99-01 (Revision 6)

November 2012

115

PWR CONTAINMENT BARRIER THRESHOLDS:

2. Inadequate Heat Removal 

There is no Loss threshold associated with Inadequate Heat Removal. 

Potential Loss 2.A 

This condition represents an IMMINENT core melt sequence which, if not corrected, 

could lead to vessel failure and an increased potential for containment failure. For this 

condition to occur, there must already have been a loss of the RCS Barrier and the Fuel 

Clad Barrier. If implementation of a procedure(s) to restore adequate core cooling is not 

effective (successful) within 15 minutes, it is assumed that the event trajectory will likely 

lead to core melting and a subsequent challenge of the Containment Barrier. 

The restoration procedure is considered “effective” if core exit thermocouple readings are

decreasing and/or if reactor vessel level is increasing. Whether or not the procedure(s)

will be effective should be apparent within 15 minutes. The Emergency Director should 

escalate the emergency classification level as soon as it is determined that the 

procedure(s) will not be effective. 

Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration 

procedures can arrest core degradation in a significant fraction of core damage scenarios, 

and that the likelihood of containment failure is very small in these events. Given this, it 

is appropriate to provide 15 minutes beyond the required entry point to determine if 

procedural actions can reverse the core melt sequence.

Developer Notes:

Some site-specific EOPs and/or EOP user guidelines may establish decision-making 

criteria concerning the number or other attributes of thermocouple readings necessary to 

drive actions (e.g., 5 CETs reading greater than 1,200o

F is required before transitioning to 

an inadequate core cooling procedure). To maintain consistency with EOPs, these 

decision-making criteria may be used in the core exit thermocouple reading thresholds. 

Potential Loss 2.A.1 

Enter site-specific criteria requiring entry into a core cooling restoration procedure or 

prompt implementation of core cooling restoration actions. A reading of 1,200oF on the 

CETs may also be used. 

For plants that have implemented Westinghouse Owners Group Emergency Response 

Guidelines, enter the parameters and values used in the Core Cooling Red Path. 
NEI 99-01 (Revision 6)

November 2012

116 

PWR CONTAINMENT BARRIER THRESHOLDS:

Westinghouse ERG Plants 

Developers should consider including a threshold the same as, or similar to, “Core 

Cooling Red entry conditions met for 15 minutes or longer” in accordance with the 

guidance at the front of this section. 

3. RCS Activity / Containment Radiation 

There is no Loss threshold associated with RCS Activity / Containment Radiation. 

Potential Loss 3.A 

The radiation monitor reading corresponds to an instantaneous release of all reactor 

coolant mass into the containment, assuming that 20% of the fuel cladding has failed. 

This level of fuel clad failure is well above that used to determine the analogous Fuel 

Clad Barrier Loss and RCS Barrier Loss thresholds. 

NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power 

Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in 

order for there to be a major release of radioactivity requiring offsite protective actions. 

For this condition to exist, there must already have been a loss of the RCS Barrier and the 

Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of 

containment which would then escalate the emergency classification level to a General 

Emergency.

Developer Notes:

Potential Loss 3.A 

NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power 

Plant Accidents, provides the basis for using the 20% fuel cladding failure value. Unless 

there is a site-specific analysis justifying a different value, the reading should be 

determined assuming the instantaneous release and dispersal of the reactor coolant noble 

gas and iodine inventory associated with 20% fuel clad failure into the containment 

atmosphere.

4. Containment Integrity or Bypass 

Loss 4.A 

These thresholds address a situation where containment isolation is required and one of 

two conditions exists as discussed below. Users are reminded that there may be accident 

and release conditions that simultaneously meet both thresholds 4.A.1 and 4.A.2. 
NEI 99-01 (Revision 6)

November 2012

117

PWR CONTAINMENT BARRIER THRESHOLDS:

4.A.1 – Containment integrity has been lost, i.e., the actual containment atmospheric leak 

rate likely exceeds that associated with allowable leakage (or sometimes referred to as 

design leakage). Following the release of RCS mass into containment, containment 

pressure will fluctuate based on a variety of factors; a loss of containment integrity

condition may (or may not) be accompanied by a noticeable drop in containment 

pressure. Recognizing the inherent difficulties in determining a containment leak rate 

during accident conditions, it is expected that the Emergency Director will assess this 

threshold using judgment, and with due consideration given to current plant conditions, 

and available operational and radiological data (e.g., containment pressure, readings on 

radiation monitors outside containment, operating status of containment pressure control 

equipment, etc.). 

Refer to the middle piping run of Figure 9-F-4. Two simplified examples are provided. 

One is leakage from a penetration and the other is leakage from an in-service system

valve. Depending upon radiation monitor locations and sensitivities, the leakage could be 

detected by any of the four monitors depicted in the figure. 

Another example would be a loss or potential loss of the RCS barrier, and the 

simultaneous occurrence of two FAULTED locations on a steam generator where one 

fault is located inside containment (e.g., on a steam or feedwater line) and the other 

outside of containment. In this case, the associated steam line provides a pathway for the 

containment atmosphere to escape to an area outside the containment. 

Following the leakage of RCS mass into containment and a rise in containment pressure, 

there may be minor radiological releases associated with allowable (design) containment 

leakage through various penetrations or system components. These releases do not 

constitute a loss or potential loss of containment but should be evaluated using the 

Recognition Category A ICs. 

4.A.2 – Conditions are such that there is an UNISOLABLE pathway for the migration of 

radioactive material from the containment atmosphere to the environment. As used here, 

the term “environment” includes the atmosphere of a room or area, outside the 

containment, that may, in turn, communicate with the outside-the-plant atmosphere (e.g., 

through discharge of a ventilation system or atmospheric leakage). Depending upon a 

variety of factors, this condition may or may not be accompanied by a noticeable drop in 

containment pressure. 

Refer to the top piping run of Figure 9-F-4. In this simplified example, the inboard and 

outboard isolation valves remained open after a containment isolation was required (i.e., 

containment isolation was not successful). There is now an UNISOLABLE pathway 

from the containment to the environment. 

The existence of a filter is not considered in the threshold assessment. Filters do not 

remove fission product noble gases. In addition, a filter could become ineffective due to 

iodine and/or particulate loading beyond design limits (i.e., retention ability has been 

exceeded) or water saturation from steam/high humidity in the release stream.
NEI 99-01 (Revision 6)

November 2012

118 

PWR CONTAINMENT BARRIER THRESHOLDS:

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

Refer to the bottom piping run of Figure 9-F-4. In this simplified example, leakage in an 

RCP seal cooler is allowing radioactive material to enter the Auxiliary Building. The 

radioactivity would be detected by the Process Monitor. If there is no leakage from the 

closed water cooling system to the Auxiliary Building, then no threshold has been met. If 

the pump or system piping developed a leak that allowed steam/water to enter the 

Auxiliary Building, then threshold 4.B would be met. Depending upon radiation monitor 

locations and sensitivities, this leakage could be detected by any of the four monitors 

depicted in the figure and cause threshold 4.A.1 to be met as well. 

Following the leakage of RCS mass into containment and a rise in containment pressure, 

there may be minor radiological releases associated with allowable (design) containment 

leakage through various penetrations or system components. Minor releases may also 

occur if a containment isolation valve(s) fails to close but the containment atmosphere 

escapes to a closed system. These releases do not constitute a loss or potential loss of 

containment but should be evaluated using the Recognition Category A ICs. 

The status of the containment barrier during an event involving steam generator tube 

leakage is assessed using Loss Threshold 1.A.

Loss 4.B

Containment sump, temperature, pressure and/or radiation levels will increase if reactor 

coolant mass is leaking into the containment. If these parameters have not increased, 

then the reactor coolant mass may be leaking outside of containment (i.e., a containment 

bypass sequence). Increases in sump, temperature, pressure, flow and/or radiation level 

readings outside of the containment may indicate that the RCS mass is being lost outside 

of containment. 

Unexpected elevated readings and alarms on radiation monitors with detectors outside 

containment should be corroborated with other available indications to confirm that the 

source is a loss of RCS mass outside of containment. If the fuel clad barrier has not been 

lost, radiation monitor readings outside of containment may not increase significantly;

however, other unexpected changes in sump levels, area temperatures or pressures, flow 

rates, etc. should be sufficient to determine if RCS mass is being lost outside of the 

containment. 

Refer to the middle piping run of Figure 9-F-4. In this simplified example, a leak has 

occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building. 

Depending upon radiation monitor locations and sensitivities, the leakage could be 

detected by any of the four monitors depicted in the figure and cause threshold 4.A.1 to 

be met as well. 

To ensure proper escalation of the emergency classification, the RCS leakage outside of 

containment must be related to the mass loss that is causing the RCS Loss and/or 

Potential Loss threshold 1.A to be met. 
NEI 99-01 (Revision 6)

November 2012

119

PWR CONTAINMENT BARRIER THRESHOLDS:

Potential Loss 4.A 

If containment pressure exceeds the design pressure, there exists a potential to lose the 

Containment Barrier. To reach this level, there must be an inadequate core cooling 

condition for an extended period of time; therefore, the RCS and Fuel Clad barriers 

would already be lost. Thus, this threshold is a discriminator between a Site Area 

Emergency and General Emergency since there is now a potential to lose the third 

barrier.

Potential Loss 4.B 

The existence of an explosive mixture means, at a minimum, that the containment 

atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the 

lower deflagration limit). A hydrogen burn will raise containment pressure and could 

result in collateral equipment damage leading to a loss of containment integrity. It 

therefore represents a potential loss of the Containment Barrier.

Potential Loss 4.C 

This threshold describes a condition where containment pressure is greater than the 

setpoint at which containment energy (heat) removal systems are designed to 

automatically actuate, and less than one full train of equipment is capable of operating 

per design. The 15-minute criterion is included to allow operators time to manually start 

equipment that may not have automatically started, if possible. This threshold represents 

a potential loss of containment in that containment heat removal/depressurization systems 

(e.g., containment sprays, ice condenser fans, etc., but not including containment venting 

strategies) are either lost or performing in a degraded manner. 

Developer Notes:

Loss 4.A.1 

Developers may include a list of site-specific radiation monitors to better define this 

threshold. Expected monitor alarms or readings may also be included. 

Potential Loss 4.A 

The site-specific pressure is the containment design pressure. 

For plants that have implemented Westinghouse Owners Group Emergency Response 

Guidelines, the pressure value in Potential Loss 4.A is that used for the Containment Red 

Path. If the Containment CSFST contains more than one Red Path due to other 

dependencies (e.g., status of containment isolation), enter the highest containment 

pressure value shown on the tree. This is typically the containment design pressure. 
NEI 99-01 (Revision 6)

November 2012

120 

 PWR CONTAINMENT BARRIER THRESHOLDS:

Potential Loss 4.B 

Developers may enter the minimum containment atmospheric hydrogen concentration 

necessary to support a hydrogen burn (i.e., the lower deflagration limit). A concurrent 

containment oxygen concentration may be included if the plant has this indication 

available in the Control Room. 

Potential Loss 4.C 

Enter the site-specific pressure setpoint 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 

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).

This threshold is not applicable to the U.S. Evolutionary Power Reactor (EPR) design. 

Westinghouse ERG Plants 

As a potential loss indication, developers should consider including a threshold the same 

as, or similar to, “Containment Red entry conditions met” in accordance with the 

guidance at the front of this section. 

5. Other Indications

Loss and/or Potential Loss 5.A 

This subcategory addresses other site-specific thresholds that may be included to indicate 

loss or potential loss of the Containment barrier based on plant-specific design 

characteristics not considered in the generic guidance. 

Developer Notes:

Loss and/or Potential Loss 5.A 

If site emergency operating procedures provide for venting of the containment as a means 

of preventing catastrophic failure, a Loss threshold should be included for the 

containment barrier. This threshold would be met as soon as such venting is 

IMMINENT. Containment venting as part of recovery actions is classified in accordance 

with the radiological effluent ICs.

Developers should determine if other reliable indicators exist to evaluate the status of this 

fission product barrier (e.g., review accident analyses described in the site Final Safety 

Analysis Report, as updated). The goal is to identify any unique or site-specific 

indications that will promote timely and accurate assessment of barrier status. 
NEI 99-01 (Revision 6)

November 2012

121

PWR CONTAINMENT BARRIER THRESHOLDS:

Any added thresholds should represent approximately the same relative threat to the 

barrier as the other thresholds in this column. Basis information for the other thresholds 

may be used to gauge the relative barrier threat level. 

6. Emergency Director Judgment

Loss 6.A 

This threshold addresses any other factors that may 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. 

Developer Notes:

None 
NEI 99-01 (Revision 6)

November 2012

122 

Figure 9-F-4: PWR Containment Integrity or Bypass Examples 

Inside Containment

Vent Auxiliary Building

Filter

Closed Cooling 

Water System

RCP Seal 

Cooling 

Interface leakage point

Open valve

4.A.1 -

Airborne 

release from 

penetration

Penetration

Damper

Damper

Process 

Monitor 

Airborne 

Monitor

Area 

Monitor

Effluent 

Monitor

4.B - RCS 

leakage 

outside CNMT

Open valve

Open valve

Open valve

Open valve

Open valve

4.A.2 - Airborne 

release from 

pathway

4.A.1 -

Airborne 

release from 

valve

Pump
NEI 99-01 (Revision 6)

October 2011

123

10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS

Table H-1: Recognition Category “H” Initiating Condition Matrix

UNUSUAL EVENT ALERT SITE AREA 

EMERGENCY

GENERAL 

EMERGENCY

HU1 Confirmed 

SECURITY 

CONDITION or threat.

Op. Modes: All

HA1 HOSTILE 

ACTION within the 

OWNER 

CONTROLLED AREA 

or airborne attack threat

within 30 minutes. 

Op. Modes: All

HS1 HOSTILE 

ACTION within the 

PROTECTED AREA.

Op. Modes: All

HG1 HOSTILE 

ACTION resulting in 

loss of physical control 

of the facility. 

Op. Modes: All

HU2 Seismic event

greater than OBE levels.

Op. Modes: All

HU3 Hazardous event.

Op. Modes: All

HU4 FIRE potentially 

degrading the level of 

safety of the plant. 

Op. Modes: All

HA5 Gaseous release 

impeding access to 

equipment necessary for 

normal plant operations, 

cooldown or shutdown. 

Op. Modes: All 

HA6 Control Room 

evacuation resulting in 

transfer of plant control 

to alternate locations. 

Op. Modes: All 

HS6 Inability to 

control a key safety 

function from outside the 

Control Room. 

Op. Modes: All

HU7 Other conditions 

exist which in the 

judgment of the 

Emergency Director 

warrant declaration of a

(NO)UE.

Op. Modes: All

HA7 Other conditions 

exist which in the 

judgment of the 

Emergency Director 

warrant declaration of an 

Alert.

Op. Modes: All

HS7 Other conditions 

exist which in the 

judgment of the 

Emergency Director 

warrant declaration of a 

Site Area Emergency.

Op. Modes: All 

HG7 Other conditions 

exist which in the 

judgment of the 

Emergency Director 

warrant declaration of a 

General Emergency.

Op. Modes: All 

Table intended for use by 

EAL developers. 

Inclusion in licensee 

documents is not required. 
NEI 99-01 (Revision 6)

November 2012

124 

HU1 

ECL: Notification of Unusual Event

Initiating Condition: Confirmed SECURITY CONDITION or threat.

Operating Mode Applicability: All

Example Emergency Action Levels: (1 or 2 or 3) 

(1) A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by 

the (site-specific security shift supervision). 

(2) Notification of a credible security threat directed at the site. 

(3) A validated notification from the NRC providing information of an aircraft threat. 

Basis:

This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, 

and thus represent a potential degradation in the level of plant safety. Security events which do 

not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 

10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs 

HA1, HS1 and HG1. 

Timely and accurate communications between Security Shift Supervision and the Control Room 

is essential for proper classification of a security-related event. Classification of these events 

will initiate appropriate threat-related notifications to plant personnel and OROs.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for 

the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and 

Independent Spent Fuel Storage Installation Security Program]. 

EAL #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 10 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 #3 addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters 

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

The status and size of the plane may also be provided by NORAD through the NRC. Validation 

of the threat is performed in accordance with (site-specific procedure). 

Emergency plans and implementing procedures are public documents; therefore, EALs should 

not incorporate Security-sensitive information. This includes information that may be 

advantageous to a potential adversary, such as the particulars concerning a specific threat or 

threat location. Security-sensitive information should be contained in non-public documents 

such as the Security Plan.
NEI 99-01 (Revision 6)

November 2012

125

Escalation of the emergency classification level would be via IC HA1. 

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, 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 

example, an EAL may be worded as “Security event #2, #5 or #9 is reported by the (site-specific 

security shift supervision).” 

ECL Assignment Attributes: 3.1.1.A 
NEI 99-01 (Revision 6)

November 2012

126 

HU2 

ECL: Notification of Unusual Event

Initiating Condition: Seismic event greater than OBE levels. 

Operating Mode Applicability: All

Example Emergency Action Levels: 

(1) Seismic event greater than Operating Basis Earthquake (OBE) as indicated by: 

(site-specific indication that a seismic event met or exceeded OBE limits) 

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)7

. An earthquake greater than an OBE but 

less than a Safe Shutdown Earthquake (SSE)8 should have no significant impact on safetyrelated systems, structures and components; however, some time may be required for the plant 

staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and 

post-event inspections). Given the time necessary to perform walk-downs and inspections, and 

fully understand any impacts, this event represents a potential degradation of the level of safety 

of the plant. 

Event verification with external sources should not be necessary during or following an OBE. 

Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a 

seismic event (e.g., typical lateral accelerations are in excess of 0.08g). The Shift Manager or 

Emergency Director may seek external verification if deemed appropriate (e.g., a call to the 

USGS, check internet news sources, etc.); however, the verification action must not preclude a 

timely emergency declaration. 

Depending upon the plant mode at the time of the event, escalation of the emergency 

classification level would be via IC CA6 or SA9. 

Developer Notes:

This “site-specific indication that a seismic event met or exceeded OBE 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 which require lengthy times to assess (e.g., processing of scratch plates or 

 

7 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. 8 An SSE is vibratory ground motion for which certain (generally, safety-related) structures, systems, and 

components must be designed to remain functional. 
NEI 99-01 (Revision 6)

November 2012

127

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 OBE indications within the Control Room, 

developers should use the following alternate EAL (or similar wording). 

(1) a. Control Room personnel feel an actual or potential seismic event. 

 AND

 b. The occurrence of a seismic event is confirmed in manner deemed appropriate by 

the Shift Manager or Emergency Director. 

The EAL 1.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 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. It is recognized that this alternate EAL wording may cause a site 

to declare an Unusual Event while another site, similarly affected but with readily assessable 

OBE indications in the Control Room, may not. 

The above alternate wording may also be used to develop a compensatory EAL for use during 

periods when a seismic monitoring system capable of detecting an OBE is out-of-service for 

maintenance or repair.

ECL Assignment Attributes: 3.1.1.A 
NEI 99-01 (Revision 6)

November 2012

128 

HU3

ECL: Notification of Unusual Event

Initiating Condition: Hazardous event. 

Operating Mode Applicability: All

Example Emergency Action Levels: (1 or 2 or 3 or 4 or 5) 

Note: EAL #4 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle 

breakdowns or accidents. 

(1) A tornado strike within the PROTECTED AREA. 

(2) Internal room or area flooding of a magnitude sufficient to require manual or automatic 

electrical isolation of a SAFETY SYSTEM component needed for the current operating 

mode. 

(3) Movement of personnel within the PROTECTED AREA is impeded due to an offsite 

event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).

(4) A hazardous event that results in on-site conditions sufficient to prohibit the plant staff 

from accessing the site via personal vehicles.

(5) (Site-specific list of natural or technological hazard events) 

Basis:

This IC addresses hazardous events that are considered to represent a potential degradation of the 

level of safety of the plant.

EAL #1 addresses a tornado striking (touching down) within the Protected Area. 

EAL #2 addresses flooding of a building room or area that results in operators isolating power to 

a SAFETY SYSTEM component due to water level or other wetting concerns. Classification is 

also required if the water level or related wetting causes an automatic isolation of a SAFETY 

SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant 

classification, operability of the affected component must be required by Technical 

Specifications for the current operating mode. 

EAL #3 addresses a hazardous materials event originating at an offsite location and of sufficient 

magnitude to impede the movement of personnel within the PROTECTED AREA. 

EAL #4 addresses a hazardous event that causes an on-site impediment to vehicle movement and 

significant enough to prohibit the plant staff from accessing the site using personal vehicles. 

Examples of such an event include site flooding caused by a hurricane, heavy rains, up-river 

water releases, dam failure, etc., or an on-site train derailment blocking the access road. 
NEI 99-01 (Revision 6)

November 2012

129

This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehicle 

breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew 

strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest 

floods of 1993, or the flooding around Ft. Calhoun Station in 2011. 

EAL #5 addresses (site-specific description). 

Escalation of the emergency classification level would be based on ICs in Recognition 

Categories A, F, S or C. 

Developer Notes:

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 within a relatively short time frame. 

In these cases, a damage assessment can be performed soon after the event, and the plant staff 

will be able to identify potential or actual impacts to plant systems and structures. This will 

enable prompt definition and implementation of compensatory or corrective measures with no 

appreciable increase in risk to the public. 

To the extent 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 expected 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. 

ECL Assignment Attributes: 3.1.1.A and 3.1.1.C 
NEI 99-01 (Revision 6)

November 2012

130 

HU4

ECL: Notification of Unusual Event

Initiating Condition: FIRE potentially degrading the level of safety of the plant. 

Operating Mode Applicability: All

Example Emergency Action Levels: (1 or 2 or 3 or 4) 

Note: The Emergency Director should declare the Unusual Event promptly upon 

determining that the applicable time has been exceeded, or will likely be exceeded.

(1) a. A FIRE is NOT extinguished within 15-minutes of ANY of the following FIRE 

detection indications: 

 Report from the field (i.e., visual observation) 

 Receipt of multiple (more than 1) fire alarms or indications 

 Field verification of a single fire alarm

AND

b. The FIRE is located within ANY of the following plant rooms or areas:

(site-specific list of plant rooms or areas) 

(2) a. Receipt of a single fire alarm (i.e., no other indications of a FIRE).

 AND

 b. The FIRE is located within ANY of the following plant rooms or areas: 

 (site-specific list of plant rooms or areas) 

 AND

 c. The existence of a FIRE is not verified within 30-minutes of alarm receipt.

(3) A FIRE within the plant or ISFSI [for plants with an ISFSI outside the plant Protected 

Area] PROTECTED AREA not extinguished within 60-minutes of the initial report, 

alarm or indication. 

(4) A FIRE within the plant or ISFSI [for plants with an ISFSI outside the plant Protected 

Area] PROTECTED AREA that requires firefighting support by an offsite fire response 

agency to extinguish. 

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.
NEI 99-01 (Revision 6)

November 2012

131

EAL #1 

The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES 

that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other 

indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression 

system, etc. 

Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, 

indication, or report. For EAL assessment purposes, the emergency declaration clock starts at 

the time that the initial alarm, indication, or report was received, and not the time that a 

subsequent verification action was performed. Similarly, the fire duration clock also starts at the 

time of receipt of the initial alarm, indication or report.

EAL #2 

This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified 

(i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take 

prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 

30-minute clock starts at the time that the initial alarm was received, and not the time that a 

subsequent verification action was performed. 

A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure 

or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to 

verify the validity of the alarm. The 30-minute period is a reasonable amount of time to 

determine if an actual FIRE exists; however, after that time, and absent information to the 

contrary, it is assumed that an actual FIRE is in progress.

If an actual FIRE is verified by a report from the field, then EAL #1 is immediately applicable, 

and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the 

report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this 

verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable 

and no emergency declaration is warranted. 

EAL #3 

In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant PROTECTED 

AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. 

This basis extends to a FIRE occurring within the PROTECTED AREA of an ISFSI located 

outside the plant PROTECTED AREA. [Sentence for plants with an ISFSI outside the plant 

Protected Area] 

EAL #4 

If a FIRE within the plant or ISFSI [for plants with an ISFSI outside the plant Protected Area] 

PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency 

(e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The 

dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it 

is needed to actively support firefighting efforts because the fire is beyond the capability of the 
NEI 99-01 (Revision 6)

November 2012

132 

Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on 

stand-by, or supporting post-extinguishment recovery or investigation actions. 

Basis-Related Requirements from Appendix R 

Appendix R to 10 CFR 50, states in part:

Criterion 3 of Appendix A to this part specifies that "Structures, systems, and 

components important to safety shall be designed and located to minimize, consistent 

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

When considering the effects of fire, those systems associated with achieving and 

maintaining safe shutdown conditions assume major importance to safety because 

damage to them can lead to core damage resulting from loss of coolant through boil-off.

Because fire may affect safe shutdown systems and because the loss of function of 

systems used to mitigate the consequences of design basis accidents under post-fire 

conditions does not per se impact public safety, the need to limit fire damage to systems 

required to achieve and maintain safe shutdown conditions is greater than the need to 

limit fire damage to those systems required to mitigate the consequences of design basis 

accidents.

In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour 

fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one 

redundant train (G.2.c). As used in EAL #2, the 30-minutes to verify a single alarm is well 

within this worst-case 1-hour time period. 

Depending upon the plant mode at the time of the event, escalation of the emergency 

classification level would be via IC CA6 or SA9. 

Developer 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.

ECL Assignment Attributes: 3.1.1.A 
NEI 99-01 (Revision 6)

November 2012

133

HU7 

ECL: Notification of Unusual Event

Initiating Condition: Other conditions exist which in the judgment of the Emergency Director 

warrant declaration of a (NO)UE.

Operating Mode Applicability: All

Example 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 occurs.

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.
NEI 99-01 (Revision 6)

November 2012

134 

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)

(1) A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED 

AREA as reported by the (site-specific security shift supervision). 

(2) A validated notification from NRC of an aircraft attack threat within 30 minutes of the 

site. 

Basis:

This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED 

AREA or notification of an aircraft attack threat. This event will require rapid response and 

assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the 

need to prepare the plant and staff for a potential aircraft impact. 

Timely and accurate communications between Security Shift Supervision and the Control Room 

is essential for proper classification of a security-related event. 

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for 

the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and 

Independent Spent Fuel Storage Installation Security Program]. 

As time and conditions allow, these events require a heightened state of readiness by the plant 

staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). 

The Alert declaration will also heighten the awareness of Offsite Response Organizations, 

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 #1 is applicable for any HOSTILE ACTION occurring, or that has occurred, in the 

OWNER CONTROLLED AREA. This includes any action directed against an ISFSI that is 

located outside the plant PROTECTED AREA. 

EAL #2 addresses the threat from the impact of an aircraft on the plant, and the anticipated 

arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related 

notifications are made in a timely manner so that plant personnel and OROs are in a heightened 
NEI 99-01 (Revision 6)

November 2012

135

state of readiness. This EAL is met when the threat-related information has been validated in 

accordance with (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 

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 HS1. 

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.

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 

example, an EAL may be worded 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 OWNER CONTROLLED AREA.

ECL Assignment Attributes: 3.1.2.D 
NEI 99-01 (Revision 6)

November 2012

136 

HA5 

ECL: Alert

Initiating Condition: Gaseous release impeding access to equipment necessary for normal plant 

operations, cooldown or shutdown. 

Operating Mode Applicability: All

Example Emergency Action Levels: 

Note: If the equipment in the listed room or area was already inoperable or out-of-service

before the event occurred, then no emergency classification is warranted. 

(1) a. Release of a toxic, corrosive, asphyxiant or flammable gas into any of the 

following plant rooms or areas:

(site-specific list of plant rooms or areas with entry-related mode applicability

identified) 

AND

b. Entry into the room or area is prohibited or impeded.

Basis:

This IC addresses an event involving a release of a hazardous gas that precludes or impedes 

access to equipment necessary to maintain normal plant operation, or required for a normal plant 

cooldown and shutdown. This condition represents an actual or potential substantial degradation 

of the level of safety of the plant. 

An Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally 

required during the plant operating mode in effect at the time of the gaseous release. The 

emergency classification is not contingent upon whether entry is actually necessary at the time of 

the release.

Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the 

Emergency Director’s judgment that the gas concentration in the affected room/area is sufficient 

to preclude or significantly impede procedurally required access. This judgment may be based 

on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, 

advice from a subject matter expert or operating experience with the same or similar hazards. 

Access should be considered as impeded if extraordinary measures are necessary to facilitate 

entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such 

as SCBAs, that is not routinely employed). 

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

 The plant is in an operating mode different than the mode specified for the affected 

room/area (i.e., entry is not required during the operating mode in effect at the time of the 

gaseous release). For example, the plant is in Mode 1 when the gaseous release occurs, and 
NEI 99-01 (Revision 6)

November 2012

137

the procedures used for normal operation, cooldown and shutdown do not require entry into 

the affected room until Mode 4. 

 The gas release is a planned activity that includes compensatory measures which address the 

temporary inaccessibility of a room or area (e.g., fire suppression system testing). 

 The action for which room/area entry is required is of an administrative or record keeping 

nature (e.g., normal rounds or routine inspections). 

 The access control measures are of a conservative or precautionary nature, and would not 

actually prevent or impede a required action. 

An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. 

Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This 

reduces the concentration of oxygen below the normal level of around 19%, which can lead to 

breathing difficulties, unconsciousness or even death. 

This EAL does not apply to firefighting activities that automatically or manually activate a fire 

suppression system in an area, or to intentional inerting of containment (BWR only). 

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

ICs. 

Developer Notes:

The “site-specific list of plant rooms or areas with entry-related mode applicability identified” 

should specify those rooms or areas that contain equipment which require a manual/local action 

as specified in operating procedures used for normal plant operation, cooldown and shutdown. 

Do not include rooms or areas in which actions of a contingent or emergency nature would be 

performed (e.g., an action to address an off-normal or emergency condition such as emergency 

repairs, corrective measures or emergency operations). In addition, the list should specify the 

plant mode(s) during which entry would be required for each room or area. 

The list should not include rooms or areas for which entry is required solely to perform actions 

of an administrative or record keeping nature (e.g., normal rounds or routine inspections). 

The list need not include the Control Room if adequate engineered safety/design features are in 

place to preclude a Control Room evacuation due to the release of a hazardous gas. Such 

features may include, but are not limited to, capability to draw air from multiple air intakes at 

different and separate locations, inner and outer atmospheric boundaries, or the capability to 

acquire and maintain positive pressure within the Control Room envelope. 

If the equipment in the listed room or area was already inoperable, or out-of-service, before the 

event occurred, then no emergency should be declared since the event will have no adverse 

impact beyond that already allowed by Technical Specifications at the time of the event. 

ECL Assignment Attributes: 3.1.2.B 
NEI 99-01 (Revision 6)

November 2012

138 

HA6 

ECL: Alert

Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate 

locations. 

Operating Mode Applicability: All

Example Emergency Action Levels: 

(1) An event has resulted in plant control being transferred from the Control Room to (sitespecific remote shutdown panels and local control stations). 

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 HS6. 

Developer Notes:

The “site-specific remote shutdown 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. 

ECL Assignment Attributes: 3.1.2.B 
NEI 99-01 (Revision 6)

November 2012

139

HA7 

ECL: Alert

Initiating Condition: Other conditions exist which in the judgment of the Emergency Director 

warrant declaration of an Alert.

Operating Mode Applicability: All

Example 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 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. 

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.
NEI 99-01 (Revision 6)

November 2012

140 

HS1

ECL: Site Area Emergency

Initiating Condition: HOSTILE ACTION within the PROTECTED AREA.

Operating Mode Applicability: All

Example Emergency Action Levels: 

(1) A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as 

reported by the (site-specific security shift supervision). 

Basis:

This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. 

This event will require rapid response and assistance due to the possibility for damage to plant 

equipment. 

Timely and accurate communications between Security Shift Supervision and the Control Room 

is essential for proper classification of a security-related event.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for 

the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and 

Independent Spent Fuel Storage Installation Security Program]. 

As time and conditions allow, these events require a heightened state of readiness by the plant 

staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). 

The Site Area Emergency declaration will mobilize ORO resources and have them available to 

develop and implement public protective actions in the unlikely event that the attack is 

successful in impairing multiple safety functions. 

This IC does not apply to a HOSTILE ACTION directed at an ISFSI PROTECTED AREA 

located outside the plant PROTECTED AREA; such an attack should be assessed using IC HA1. 

It also does not apply to incidents that are accidental events, acts of civil disobedience, or 

otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include 

the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. 

Reporting of these types of events is adequately addressed by other EALs, or the requirements of 

10 CFR § 73.71 or 10 CFR § 50.72. 

Emergency plans and implementing procedures are public documents; therefore, EALs should 

not incorporate Security-sensitive information. This includes information that may be 

advantageous to a potential adversary, such as the particulars concerning a specific threat or 

threat location. Security-sensitive information should be contained in non-public documents 

such as the Security Plan. 

Escalation of the emergency classification level would be via IC HG1. 
NEI 99-01 (Revision 6)

November 2012

141

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.

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 

example, an EAL may be worded 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.

ECL Assignment Attributes: 3.1.3.D 
NEI 99-01 (Revision 6)

November 2012

142 

HS6 

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 Emergency promptly upon 

determining that (site-specific number of minutes) has been exceeded, or will likely be 

exceeded. 

(1) a. An event has resulted in plant control being transferred from the Control Room to 

(site-specific remote shutdown panels and local control stations). 

AND

 b. Control of ANY of the following key safety functions is not reestablished within 

(site-specific number of minutes). 

 Reactivity control 

 Core cooling [PWR] / RPV water level [BWR] 

 RCS heat removal

Basis:

This IC addresses an evacuation of the Control Room that results in transfer of plant control to 

alternate locations, and the control of a key safety function cannot be reestablished in a timely 

manner. The failure to gain control of a key safety function following a transfer of plant control 

to alternate locations is a precursor to a challenge to one or more fission product barriers within a 

relatively short period of time.

The determination of whether or not “control” is established at the remote safe shutdown 

location(s) is based on Emergency Director judgment. The Emergency Director is expected to 

make a reasonable, informed judgment within (the site-specific time for transfer) minutes 

whether or not the operating staff has control of key safety functions from the remote safe 

shutdown location(s). 

Escalation of the emergency classification level would be via IC FG1 or CG1. 

Developer Notes:

The “site-specific remote shutdown 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. 

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 
NEI 99-01 (Revision 6)

November 2012

143

period may be used with appropriate basis/justification. 

ECL Assignment Attributes: 3.1.3.B 
NEI 99-01 (Revision 6)

November 2012

144 

HS7 

ECL: Site Area Emergency

Initiating Condition: Other conditions exist which in the judgment of the Emergency Director 

warrant declaration of a Site Area Emergency.

Operating Mode Applicability: All

Example 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 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. 

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. 
NEI 99-01 (Revision 6)

November 2012

145

HG1

ECL: General Emergency

Initiating Condition: HOSTILE ACTION resulting in loss of physical control of the facility. 

Operating Mode Applicability: All

Example Emergency Action Levels: 

(1) a. A HOSTILE ACTION is occurring or has occurred within the PROTECTED 

AREA as reported by the (site-specific security shift supervision). 

AND

b. EITHER of the following has occurred: 

1. ANY of the following safety functions cannot be controlled or maintained. 

 Reactivity control

 Core cooling [PWR] / RPV water level [BWR] 

 RCS heat removal

OR

2. Damage to spent fuel has occurred or is IMMINENT. 

Basis:

This IC addresses an event in which a HOSTILE FORCE has taken physical control of the 

facility to the extent that the plant staff can no longer operate equipment necessary to maintain 

key safety functions. It also addresses a HOSTILE ACTION leading to a loss of physical control 

that results in actual or IMMINENT damage to spent fuel due to 1) damage to a spent fuel pool 

cooling system (e.g., pumps, heat exchangers, controls, etc.) or, 2) loss of spent fuel pool 

integrity such that sufficient water level cannot be maintained.

Timely and accurate communications between Security Shift Supervision and the Control Room 

is essential for proper classification of a security-related event.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for 

the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and 

Independent Spent Fuel Storage Installation Security Program]. 

Emergency plans and implementing procedures are public documents; therefore, EALs 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. 
NEI 99-01 (Revision 6)

November 2012

146 

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.

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 

example, an EAL may be worded 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.

ECL Assignment Attributes: 3.1.4.D 

 
NEI 99-01 (Revision 6)

November 2012

147

HG7 

ECL: General Emergency

Initiating Condition: Other conditions exist which in the judgment of the Emergency Director 

warrant declaration of a General Emergency.

Operating Mode Applicability: All

Example 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 involve actual or IMMINENT substantial 

core degradation or melting with potential for loss of containment integrity or HOSTILE 

ACTION that results in an actual loss of physical control of the facility. Releases can be 

reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite 

for more than the immediate site area.

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.
NEI 99-01 (Revision 6)

November 2012

148 

11 SYSTEM MALFUNCTION ICS/EALS

Table S-1: Recognition Category “S” Initiating Condition Matrix 

UNUSUAL EVENT ALERT SITE AREA 

EMERGENCY

GENERAL 

EMERGENCY

SU1 Loss of all offsite 

AC power capability to 

emergency buses for 15 

minutes or longer. 

Op. Modes: Power 

Operation, Startup, Hot 

Standby, Hot Shutdown

SA1 Loss of all but 

one AC power source to 

emergency buses for 15 

minutes or longer. 

Op. Modes: Power 

Operation, Startup, Hot 

Standby, Hot Shutdown

SS1 Loss of all offsite 

and all onsite AC power 

to emergency buses for 15 

minutes or longer. 

Op. Modes: Power 

Operation, Startup, Hot 

Standby, Hot Shutdown

SG1 Prolonged loss of 

all offsite and all onsite 

AC power to emergency 

buses.

Op. Modes: Power 

Operation, Startup, Hot 

Standby, Hot Shutdown

SU2 UNPLANNED 

loss of Control Room 

indications for 15 

minutes or longer. 

Op. Modes: Power 

Operation, Startup, Hot 

Standby, Hot Shutdown

SA2 UNPLANNED 

loss of Control Room 

indications for 15 

minutes or longer with a 

significant transient in 

progress. 

Op. Modes: Power 

Operation, Startup, Hot 

Standby, Hot Shutdown

SU3 Reactor coolant 

activity greater than 

Technical Specification 

allowable limits. 

Op. Modes: Power 

Operation, Startup, Hot 

Standby, Hot Shutdown

SU4 RCS leakage for 

15 minutes or longer. 

Op. Modes: Power 

Operation, Startup, Hot 

Standby, Hot Shutdown 

SU5 Automatic or

manual (trip [PWR] / 

scram [BWR]) fails to 

shutdown the reactor. 

Op. Modes: Power 

Operation

SA5 Automatic or 

manual (trip [PWR] / 

scram [BWR]) fails to 

shutdown the reactor, 

and subsequent manual 

actions taken at the 

reactor control consoles 

are not successful in 

shutting down the 

reactor. 

Op. Modes: Power 

Operation

SS5 Inability to 

shutdown the reactor 

causing a challenge to 

(core cooling [PWR] / 

RPV water level [BWR]) 

or RCS heat removal. 

Op. Modes: Power 

Operation Table intended for use by 

EAL developers. 

Inclusion in licensee 

documents is not required. 
NEI 99-01 (Revision 6)

November 2012

149

UNUSUAL EVENT ALERT SITE AREA 

EMERGENCY

GENERAL 

EMERGENCY

SU6 Loss of all onsite 

or offsite 

communications 

capabilities.

Op. Modes: Power 

Operation, Startup, Hot 

Standby, Hot Shutdown

SU7 Failure to isolate 

containment or loss of 

containment pressure 

control. [PWR] 

Op. Modes: Power 

Operation, Startup, Hot 

Standby, Hot Shutdown

SS8 Loss of all Vital

DC power for 15 minutes 

or longer. 

Op. Modes: Power 

Operation, Startup, Hot 

Standby, Hot Shutdown 

SG8 Loss of all AC 

and Vital DC power 

sources for 15 minutes or 

longer. 

Op. Modes: Power 

Operation, Startup, Hot 

Standby, Hot Shutdown

SA9 Hazardous event 

affecting a SAFETY 

SYSTEM needed for the 

current operating mode. 

Op. Modes: Power 

Operation, Startup, Hot 

Standby, Hot Shutdown

Table intended for use by 

EAL developers. 

Inclusion in licensee 

documents is not required. 
NEI 99-01 (Revision 6)

November 2012

150 

SU1

ECL: Notification of Unusual Event

Initiating Condition: Loss of all offsite AC power capability to emergency buses for 15 

minutes or longer. 

Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown 

Example Emergency Action Levels: 

Note: The Emergency Director should declare the Unusual Event promptly upon determining 

that 15 minutes has been exceeded, or will likely be exceeded. 

(1) Loss of ALL offsite AC power capability to (site-specific emergency buses) for 15 

minutes or longer. 

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 buses. This condition 

represents a potential reduction in the level of safety of the plant. 

For emergency classification purposes, “capability” means that an offsite AC power source(s) is 

available to the emergency buses, whether or not the buses are powered from it. 

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

power. 

Escalation of the emergency classification level would be via IC SA1. 

Developer 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 within 15 minutes. Consider capabilities such as power source cross-ties, 

“swing” 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 via 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. 

ECL Assignment Attributes: 3.1.1.A 
NEI 99-01 (Revision 6)

November 2012

151

SU2 

ECL: Notification of Unusual Event

Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or 

longer. 

Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown 

Example Emergency Action Levels:

Note: The Emergency Director should declare the Unusual Event promptly upon determining 

that 15 minutes has been exceeded, or will likely be exceeded. 

(1) a. An UNPLANNED event results in the inability to monitor one or more of the 

following parameters from within the Control Room for 15 minutes or longer. 

[BWR parameter list] [PWR parameter list]

Reactor Power Reactor Power

RPV Water Level RCS Level

RPV Pressure RCS Pressure

Primary Containment Pressure In-Core/Core Exit Temperature

Suppression Pool Level Levels in at least (site-specific 

number) steam generators

Suppression Pool Temperature Steam Generator Auxiliary or 

Emergency Feed Water Flow

Basis:

This IC addresses the difficulty associated with monitoring normal plant conditions without the 

ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition 

is a precursor to a more significant event and represents a potential degradation in the level of 

safety of the plant. 

As used in this EAL, an “inability to monitor” means that values for one or more of the listed 

parameters cannot be determined from within the Control Room. This situation would require a 

loss of all of the Control Room sources for the given parameter(s). For example, the reactor 

power level cannot be determined from any analog, digital and recorder source within the 

Control Room. 

An event involving a loss of plant indications, annunciators and/or display systems is evaluated

in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an 

NRC event report is required. The event would be reported if it significantly impaired the 

capability to perform emergency assessments. In particular, emergency assessments necessary to

implement abnormal operating procedures, emergency operating procedures, and emergency 

plan implementing procedures addressing emergency classification, accident assessment, or 

protective action decision-making. 
NEI 99-01 (Revision 6)

November 2012

152 

This EAL is focused on a selected subset of plant parameters associated with the key safety 

functions of reactivity control, core cooling [PWR] / RPV level [BWR] and RCS heat removal. 

The loss of the ability to determine one or more of these parameters from within the Control 

Room is considered to be more significant than simply a reportable condition. In addition, if all 

indication sources for one or more of the listed parameters are lost, then the ability to determine 

the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the 

value for reactor vessel level [PWR] / RPV water level [BWR] cannot be determined from the 

indications and recorders on a main control board, the SPDS or the plant computer, the 

availability of other parameter values may be compromised as well.

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

indication. 

Escalation of the emergency classification level would be via IC SA2. 

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 

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 vessel level in the PWR parameter column 

entry for RCS Level. 

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 determine, 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 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. 

A loss of plant annunciators will be evaluated for reportability in accordance with 10 CFR 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 CFR 50.72 (and associated guidance in NUREG-1022). The reporting of this 
NEI 99-01 (Revision 6)

November 2012

153

event will ensure adequate plant staff and NRC 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 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. 

ECL Assignment Attributes: 3.1.1.A 
NEI 99-01 (Revision 6)

November 2012

154 

SU3 

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 Shutdown 

Example Emergency Action Levels: (1 or 2)

(1) (Site-specific radiation monitor) reading greater than (site-specific value). 

(2) Sample analysis indicates that a reactor coolant activity value is greater than an allowable 

limit specified in Technical Specifications. 

Basis:

This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in 

Technical Specifications. This condition is a precursor to a more significant event and represents

a potential degradation of the level of safety of the plant. 

Escalation of the emergency classification level would be via ICs FA1 or the Recognition 

Category A ICs. 

Developer Notes:

For EAL #1 – Enter the radiation monitor(s) that may be used to readily identify when RCS 

activity levels exceed Technical Specification allowable limits. This EAL may be developed 

using different methods and sites should use existing capabilities to address it (e.g., development 

of new capabilities is not required). Examples of existing 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. 

The monitor reading values should correspond to an RCS activity level approximately at

Technical Specification allowable limits.

If there is no existing 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 Technical Specifications and the associated allowable limit(s) (e.g., 

values for dose equivalent I-131 and gross activity, time-dependent or transient values, etc.). If 

this approach is selected, all RCS activity allowable limits should be included. 

ECL Assignment Attributes: 3.1.1.A and 3.1.1.B 
NEI 99-01 (Revision 6)

November 2012

155

SU4 

ECL: Notification of Unusual Event

Initiating Condition: RCS leakage for 15 minutes or longer. 

Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown 

Example Emergency Action Levels: (1 or 2 or 3) 

Note: The Emergency Director should declare the Unusual Event promptly upon determining 

that 15 minutes has been exceeded, or will likely be exceeded. 

(1) RCS unidentified or pressure boundary leakage greater than (site-specific value) for 15 

minutes or longer. 

(2) RCS identified leakage greater than (site-specific value) for 15 minutes or longer. 

(3) Leakage from the RCS to a location outside containment greater than 25 gpm for 15 

minutes or longer. 

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 #1 and EAL #2 are focused on a loss of mass from the RCS due to “unidentified leakage", 

"pressure boundary leakage" or "identified leakage” (as these leakage types are defined in the 

plant Technical Specifications). EAL #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 (e.g., 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 #1 uses a lower value that reflects the 

greater significance of unidentified or pressure boundary leakage. 

The release of mass from the RCS due to the as-designed/expected operation of a relief valve 

does not warrant an emergency classification. For PWRs, an emergency classification would be 

required if a mass loss is caused by a relief valve that is not functioning as designed/expected 

(e.g., a relief valve sticks open and the line flow cannot be isolated). For BWRs, a stuck-open 

Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified 

leakage by Technical Specifications and, therefore, is not applicable to this EAL. 

The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the 

leakage, if possible. 

 
NEI 99-01 (Revision 6)

November 2012

156 

Escalation of the emergency classification level would be via ICs of Recognition Category A or 

F. 

Developer Notes:

EAL #1 – For the site-specific leak rate value, enter the higher of 10 gpm or the value specified 

in the site’s Technical Specifications for this type of leakage.

EAL #2 – For the site-specific leak rate value, enter the higher of 25 gpm or the value specified 

in the site’s Technical Specifications for this type of leakage. 

For sites that have 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. 

ECL Assignment Attributes: 3.1.1.A 
NEI 99-01 (Revision 6)

November 2012

157

SU5 

ECL: Notification of Unusual Event

Initiating Condition: Automatic or manual (trip [PWR] / scram [BWR]) fails to shutdown the 

reactor. 

Operating Mode Applicability: Power Operation

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. 

Example Emergency Action Levels: (1 or 2)

(1) a. An automatic (trip [PWR] / scram [BWR]) did not shutdown the reactor.

AND

 b. A subsequent manual action taken at the reactor control consoles is successful in 

shutting down the reactor. 

(2) a. A manual trip ([PWR] / scram [BWR]) did not shutdown the reactor.

AND

 b. EITHER of the following: 

 1. A subsequent manual action taken at the reactor control consoles is

successful in shutting down the reactor. 

 OR

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

shutting down the reactor. 

Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor (trip 

[PWR] / scram [BWR]) that results in a reactor shutdown, and either a subsequent operator 

manual action taken at the reactor control consoles or an automatic (trip [PWR] / scram [BWR])

is successful in shutting down the reactor. This event is a precursor to a more significant 

condition and thus represents a potential degradation of the level of safety of the plant. 

Following the failure on an automatic reactor (trip [PWR] / scram [BWR]), operators will 

promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., 

initiate a manual reactor (trip [PWR] / scram [BWR])). If these manual actions are successful in 

shutting down the reactor, core heat generation will quickly fall to a level within the capabilities 

of the plant’s decay heat removal systems. 
NEI 99-01 (Revision 6)

November 2012

158 

If an initial manual reactor (trip [PWR] / scram [BWR]) is unsuccessful, operators will promptly 

take manual action at another location(s) on the reactor control consoles to shutdown the reactor 

(e.g., initiate a manual reactor (trip [PWR] / scram [BWR])) using a different switch). Depending 

upon several factors, the initial or subsequent effort to manually (trip [PWR] / scram [BWR]) the 

reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor (trip 

[PWR] / scram [BWR]) signal. If a subsequent manual or automatic (trip [PWR] / scram [BWR]) 

is successful in shutting down the reactor, core heat generation will quickly fall to a level within 

the capabilities of the plant’s decay heat removal systems. 

A manual action at the reactor control consoles is any operator action, or set of actions, which 

causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor (trip 

[PWR] / scram [BWR])). This action does not include manually driving in control rods or 

implementation of boron injection strategies. Actions taken at back-panels or other locations 

within the Control Room, or any location outside the Control Room, are not considered to be “at 

the reactor control consoles”.

Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. 

[BWR] 

The plant response to the failure of an automatic or manual reactor (trip [PWR] / scram [BWR]) 

will vary based upon several factors including the reactor power level prior to the event, 

availability of the condenser, performance of mitigation equipment and actions, other concurrent 

plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles 

are also unsuccessful in shutting down the reactor, then the emergency classification level will 

escalate to an Alert via IC SA5. Depending upon the plant response, escalation is also possible 

via IC FA1. Absent the plant conditions needed to meet either IC SA5 or FA1, an Unusual 

Event declaration is appropriate for this event. 

A reactor shutdown is determined in accordance with applicable Emergency Operating 

Procedure criteria.

Should a reactor (trip [PWR] / scram [BWR]) signal be generated as a result of plant work (e.g., 

RPS setpoint testing), the following classification guidance should be applied. 

 If the signal causes a plant transient that should have included an automatic reactor (trip 

[PWR] / scram [BWR]) and the RPS fails to automatically shutdown the reactor, then this IC 

and the EALs are applicable, and should be evaluated. 

 If the signal does not cause a plant transient and the (trip [PWR] / scram [BWR]) failure is 

determined through other means (e.g., assessment of test results), then this IC and the EALs 

are not applicable and no classification is warranted.

Developer Notes:

This IC is applicable in any Mode in which the actual reactor power level could exceed 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 lower bound of 

Power Operation (Mode 1) will need to include Startup (Mode 2) in the Operating Mode 

Applicability. For example, if the reactor is considered to be shutdown at 3% and Power 
NEI 99-01 (Revision 6)

November 2012

159

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 with the appropriate site-specific term (e.g., 

main control boards). 

ECL Assignment Attributes: 3.1.1.A 
NEI 99-01 (Revision 6)

November 2012

160 

SU6 

ECL: Notification of Unusual Event

Initiating Condition: Loss of all onsite or offsite communications capabilities.

Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown 

Example Emergency Action Levels: (1 or 2 or 3) 

(1) Loss of ALL of the following onsite communication methods: 

(site-specific list of communications methods)

(2) Loss of ALL of the following ORO communications methods: 

(site-specific list of communications methods) 

(3) Loss of ALL of the following NRC communications methods: 

(site-specific list of communications methods)

Basis:

This IC addresses a significant loss of on-site or offsite communications capabilities. While not 

a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs 

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 #1 addresses a total loss of the communications methods used in support of routine plant 

operations. 

EAL #2 addresses a total loss of the communications methods used to notify all OROs of an 

emergency declaration. The OROs referred to here are (see Developer Notes). 

EAL #3 addresses a total loss of the communications methods used to notify the NRC of an 

emergency declaration. 

Developer 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. 
NEI 99-01 (Revision 6)

November 2012

161

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 owned and maintained by individuals. Example methods are ring-down/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 with 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 NRC as described in the site 

Emergency Plan. The listing should include installed plant equipment and components, and not 

items owned and maintained by individuals. These methods are typically the dedicated 

Emergency Notification System (ENS) telephone line and commercial telephone lines. 

ECL Assignment Attributes: 3.1.1.C 
NEI 99-01 (Revision 6)

November 2012

162 

SU7 

ECL: Notification of Unusual Event

Initiating Condition: Failure to isolate containment or loss of containment pressure control. 

[PWR] 

Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown 

Example Emergency Action Levels: (1 or 2)

(1) a. Failure of containment to isolate when required by an actuation signal. 

AND

b. ALL required penetrations are not closed within 15 minutes of the actuation 

signal. 

(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:

This IC addresses a failure of one or more containment penetrations to automatically isolate 

(close) when required by an actuation signal. It also addresses an event that results in high 

containment pressure with a concurrent failure of containment pressure control systems. Absent 

challenges to another fission product barrier, either condition represents potential degradation of 

the level of safety of the plant. 

For EAL #1, the containment isolation signal must be generated as the result on an offnormal/accident condition (e.g., a safety injection or high containment pressure); a failure 

resulting from testing or maintenance does not warrant classification. The determination of 

containment and penetration status – isolated or not isolated – should be made in accordance 

with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute criterion is 

included to allow operators time to manually isolate the required penetrations, if possible. 

EAL #2 addresses a condition where containment pressure is greater than the setpoint at which 

containment energy (heat) removal systems are designed to automatically actuate, and less than 

one full train of equipment is capable of operating per design. The 15-minute criterion is 

included to allow operators time to manually start equipment that may not have automatically 

started, if possible. The inability to start the required equipment indicates that containment heat 

removal/depressurization systems (e.g., containment sprays or ice condenser fans) are either lost 

or performing in a degraded manner. 
NEI 99-01 (Revision 6)

November 2012

163

This event would escalate to a Site Area Emergency in accordance with IC FS1 if there were a 

concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers. 

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 Power Reactor (EPR) design. 

ECL Assignment Attributes: 3.1.1.A 
NEI 99-01 (Revision 6)

November 2012

164 

SA1

ECL: Alert

Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes 

or longer. 

Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown 

Example Emergency Action Levels: 

Note: The Emergency Director should declare the Alert promptly upon determining that 15 

minutes has been exceeded, or will likely be exceeded.

(1) a. AC power capability to (site-specific emergency buses) is reduced to a single 

power source for 15 minutes or longer. 

AND

b. Any additional single power source failure will result in a loss of all AC power to 

SAFETY SYSTEMS. 

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 safetyrelated equipment. This IC provides an escalation path from IC SU1. 

An “AC power source” is a source recognized in AOPs and EOPs, and capable of supplying 

required power to 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 power sources (e.g., onsite diesel 

generators) with 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 single train of 

emergency buses being back-fed from an offsite power source. 

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

Escalation of the emergency classification level would be via IC SS1.

Developer 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 required power to 

an AC emergency bus. For example, if a backup power source is comprised of two generators 
NEI 99-01 (Revision 6)

November 2012

165

(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 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. 

Developers should modify the bulleted examples 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 power circuit constitutes a 

single power source. For example, three independent 345kV offsite power circuits (i.e., 

incoming power lines) comprise three separate power sources. Independence may be determined 

from a review of the site-specific UFSAR, SBO 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 source is recognized in AOPs and EOPs, or beyond design basis accident 

response guidelines (e.g., FLEX support guidelines). Such power sources should generally meet 

the “Alternate ac source” definition provided in 10 CFR 50.2. 

At multi-unit stations, the EALs may credit compensatory measures that are proceduralized and 

can be implemented within 15 minutes. Consider capabilities such as power source cross-ties, 

“swing” 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 via 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. 

ECL Assignment Attributes: 3.1.2.B 
NEI 99-01 (Revision 6)

November 2012

166 

SA2

ECL: Alert

Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer 

with a significant transient in progress. 

Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown 

Example Emergency Action Levels: 

Note: The Emergency Director should declare the Alert promptly upon determining that 15 

minutes has been exceeded, or will likely be exceeded.

(1) a. An UNPLANNED event results in the inability to monitor one or more of the 

following parameters from within the Control Room for 15 minutes or longer. 

[BWR parameter list] [PWR parameter list]

Reactor Power Reactor Power

RPV Water Level RCS Level

RPV Pressure RCS Pressure

Primary Containment Pressure In-Core/Core Exit Temperature

Suppression Pool Level Levels in at least (site-specific 

number) steam generators

Suppression Pool Temperature Steam Generator Auxiliary or 

Emergency Feed Water Flow

 AND

a. ANY of the following transient events in progress. 

 Automatic or manual runback greater than 25% thermal reactor power 

 Electrical load rejection greater than 25% full electrical load 

 Reactor scram [BWR] / trip [PWR] 

 ECCS (SI) actuation 

 Thermal power oscillations greater than 10% [BWR] 

Basis:

This IC addresses the difficulty associated with monitoring rapidly changing plant conditions 

during a transient without the ability to obtain SAFETY SYSTEM parameters from within the 

Control Room. During this condition, the margin to a potential fission product barrier challenge

is reduced. It thus represents a potential substantial degradation in the level of safety of the 

plant. 

As used in this EAL, an “inability to monitor” means that values for one or more of the listed 

parameters cannot be determined from within the Control Room. This situation would require a 

loss of all of the Control Room sources for the given parameter(s). For example, the reactor 
NEI 99-01 (Revision 6)

November 2012

167

power level cannot be determined from any analog, digital and recorder source within the 

Control Room. 

An event involving a loss of plant indications, annunciators and/or display systems is evaluated 

in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an 

NRC event report is required. The event would be reported if it significantly impaired the 

capability to perform emergency assessments. In particular, emergency assessments necessary to 

implement abnormal operating procedures, emergency operating procedures, and emergency 

plan implementing procedures addressing emergency classification, accident assessment, or 

protective action decision-making. 

This EAL is focused on a selected subset of plant parameters associated with the key safety 

functions of reactivity control, core cooling [PWR] / RPV level [BWR] and RCS heat removal. 

The loss of the ability to determine one or more of these parameters from within the Control 

Room is considered to be more significant than simply a reportable condition. In addition, if all 

indication sources for one or more of the listed parameters are lost, then the ability to determine 

the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the 

value for reactor vessel level [PWR] / RPV water level [BWR] cannot be determined from the 

indications and recorders on a main control board, the SPDS or the plant computer, the 

availability of other parameter values may be compromised as well.

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

indication. 

Escalation of the emergency classification level would be via ICs FS1 or IC AS1. 

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 

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 vessel level in the PWR 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 expected 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 determine, 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 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. 
NEI 99-01 (Revision 6)

November 2012

168 

A loss of plant annunciators will be evaluated for reportability in accordance with 10 CFR 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 CFR 50.72 (and associated guidance in NUREG-1022). The reporting of this 

event will ensure adequate plant staff and NRC 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 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. 

ECL Assignment Attributes: 3.1.2.B 
NEI 99-01 (Revision 6)

November 2012

169

SA5

ECL: Alert

Initiating Condition: Automatic or manual (trip [PWR] / scram [BWR]) fails to shutdown the 

reactor, and subsequent manual actions taken at the reactor control consoles are not successful in 

shutting down the reactor. 

Operating Mode Applicability: Power Operation

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. 

Example Emergency Action Levels: 

(1) a. An automatic or manual (trip [PWR] / scram [BWR]) did not shutdown the 

reactor. 

AND

 b. Manual actions taken at the reactor control consoles are not successful in shutting 

down the reactor. 

Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor (trip 

[PWR] / scram [BWR]) that results in a reactor shutdown, and subsequent operator manual 

actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This 

condition represents an actual or potential substantial degradation of the level of safety of the 

plant. An emergency declaration is required even if the reactor is subsequently shutdown by an 

action taken away from the reactor control consoles since this event entails a significant failure 

of the RPS. 

A manual action at the reactor control consoles is any operator action, or set of actions, which 

causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor (trip 

[PWR] / scram [BWR])). This action does not include manually driving in control rods or 

implementation of boron injection strategies. 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. 

[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 
NEI 99-01 (Revision 6)

November 2012

170 

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 Emergency via IC SS5. 

Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent 

the plant conditions needed to meet either IC SS5 or FS1, an Alert declaration is appropriate for 

this event. 

It is recognized that plant responses or symptoms may also require an Alert declaration in 

accordance with the Recognition Category F ICs; however, this IC and EAL are included to 

ensure a timely emergency declaration.

A reactor shutdown is determined in accordance with applicable Emergency Operating

Procedure criteria.

Developer Notes:

This IC is applicable in any Mode in which the actual reactor power level could exceed 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 lower bound of 

Power Operation (Mode 1) will need to include Startup (Mode 2) in the Operating Mode 

Applicability. For example, 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).

The term “reactor control consoles” may be replaced with the appropriate site-specific term (e.g., 

main control boards). 

ECL Assignment Attributes: 3.1.2.B 
NEI 99-01 (Revision 6)

November 2012

171

SA9 

ECL: Alert

Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current 

operating mode. 

Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown 

Example Emergency Action Levels: 

(1) a. The occurrence of ANY of the following hazardous events: 

 Seismic event (earthquake)

 Internal or external flooding event 

 High winds or tornado strike 

 FIRE

 EXPLOSION

 (site-specific hazards)

 Other events with similar hazard characteristics as determined by the Shift 

Manager

 AND

b. EITHER of the following: 

 1. Event damage has caused indications of degraded performance in at least 

one train of a SAFETY SYSTEM needed for the current operating mode. 

 OR 

2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM 

component or structure needed for the current operating mode. 

Basis:

This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure 

containing SAFETY SYSTEM components, needed for the current operating mode. This 

condition significantly reduces the margin to a loss or potential loss of a fission product barrier, 

and therefore represents an actual or potential substantial degradation of the level of safety of the 

plant. 

EAL 1.b.1 addresses damage to a SAFETY SYSTEM train that is in service/operation since 

indications for it will be readily available. The indications of degraded performance should be 

significant enough to cause concern regarding the operability or reliability of the SAFETY 

SYSTEM train. 

EAL 1.b.2 addresses damage to a SAFETY SYSTEM component that is not in service/operation 

or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM 
NEI 99-01 (Revision 6)

November 2012

172 

components. Operators will make this determination based on the totality of available event and 

damage report information. This is intended to be a brief assessment not requiring lengthy 

analysis or quantification of the damage. 

Escalation of the emergency classification level would be via IC FS1 or AS1. 

Developer Notes:

For (site-specific hazards), developers should consider including other significant, site-specific 

hazards to the bulleted list contained in EAL 1.a (e.g., a seiche). 

Nuclear power plant SAFETY SYSTEMS are comprised of two or more separate and redundant 

trains of equipment in accordance with site-specific design criteria.

ECL Assignment Attributes: 3.1.2.B 
NEI 99-01 (Revision 6)

November 2012

173

SS1 

ECL: Site Area Emergency

Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 

minutes or longer. 

Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown 

Example Emergency Action Levels: 

Note: The Emergency Director should declare the Site Area Emergency promptly upon 

determining that 15 minutes has been exceeded, or will likely be exceeded. 

(1) Loss of ALL offsite and ALL onsite AC power to (site-specific emergency buses) for 15 

minutes or longer. 

Basis:

This IC addresses a total loss of AC power that compromises the performance of all SAFETY 

SYSTEMS requiring electric power including those necessary for emergency core cooling, 

containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. 

In addition, fission product barrier monitoring capabilities may be degraded under these 

conditions. This IC represents a condition that involves actual or likely major failures of plant 

functions needed for the protection of the public. 

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

Escalation of the emergency classification level would be via ICs AG1, FG1 or SG1.

Developer 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 example, if a backup power 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 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. 

The EAL and/or Basis section may specify use of a non-safety-related power source provided 

that operation of this source is controlled in accordance with abnormal or emergency operating 

procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines). 

Such power sources should generally meet the “Alternate ac source” definition provided in 10 

CFR 50.2. 
NEI 99-01 (Revision 6)

November 2012

174 

At multi-unit stations, the EALs may credit compensatory measures that are proceduralized and 

can be implemented within 15 minutes. Consider capabilities such as power source cross-ties, 

“swing” 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 via 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. 

ECL Assignment Attributes: 3.1.3.B 
NEI 99-01 (Revision 6)

November 2012

175

SS5 

ECL: Site Area Emergency

Initiating Condition: Inability to shutdown the reactor causing a challenge to (core cooling 

[PWR] / RPV water level [BWR]) or RCS heat removal. 

Operating Mode Applicability: Power Operation

Example Emergency Action Levels: 

(1) a. An automatic or manual (trip [PWR] / scram [BWR]) did not shutdown the 

reactor. 

 AND

 b. All manual actions to shutdown the reactor have been unsuccessful.

AND

c. EITHER of the following conditions exist: 

• (Site-specific indication of an inability to adequately remove heat from the 

core) 

• (Site-specific indication of an inability to adequately remove heat from the 

RCS) 

Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor (trip 

[PWR] / scram [BWR]) that results in a reactor shutdown, all subsequent operator actions to 

manually shutdown the reactor are unsuccessful, and continued power generation is challenging 

the capability to adequately remove heat from the core and/or the RCS. This condition will lead 

to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration 

of a Site Area Emergency. 

In some instances, the emergency classification resulting from this IC/EAL may be higher than 

that resulting from an assessment of the plant responses and symptoms against the Recognition 

Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not 

address the additional threat posed by a failure to 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.

A reactor shutdown is determined in accordance with applicable Emergency Operating 

Procedure criteria.

Escalation of the emergency classification level would be via IC AG1 or FG1.
NEI 99-01 (Revision 6)

November 2012

176 

Developer Notes:

This IC is applicable in any Mode in which the actual reactor power level could exceed 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 lower bound of 

Power Operation (Mode 1) will need to include Startup (Mode 2) in the Operating Mode 

Applicability. For example, 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 heat 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 exit 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 

exit thermocouple temperatures greater than 1,200o

F and/or a reactor vessel water level that 

corresponds to approximately 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 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 Capacity 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. 

ECL Assignment Attributes: 3.1.3.B 
NEI 99-01 (Revision 6)

November 2012

177

SS8 

ECL: Site Area Emergency

Initiating Condition: Loss of all Vital DC power for 15 minutes or longer. 

Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown 

Example Emergency Action Levels: 

Note: The Emergency Director should declare the Site Area Emergency promptly upon 

determining that 15 minutes has been exceeded, or will likely be exceeded. 

(1) Indicated voltage is less than (site-specific bus voltage value) on ALL (site-specific Vital

DC busses) for 15 minutes or longer. 

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 ICs AG1, FG1 or SG8. 

Developer Notes:

The “site-specific bus voltage value” should be based on the minimum bus voltage necessary for 

adequate operation of SAFETY SYSTEM equipment. This voltage 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. 

The typical value for an entire battery set is approximately 105 VDC. For a 60 cell string of 

batteries, the cell voltage is approximately 1.75 Volts per cell. For a 58 string battery set, the 

minimum voltage is approximately 1.81 Volts per cell. 

The “site-specific Vital DC busses” are the DC busses that provide monitoring and control 

capabilities for SAFETY SYSTEMS. 

ECL Assignment Attributes: 3.1.3.B 
NEI 99-01 (Revision 6)

November 2012

178 

SG1 

ECL: General Emergency

Initiating Condition: Prolonged loss of all offsite and all onsite AC power to emergency buses.

Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown 

Example Emergency Action Levels: 

Note: The Emergency Director should declare the General Emergency promptly upon 

determining that (site-specific hours) has been exceeded, or will likely be exceeded. 

(1) a. Loss of ALL offsite and ALL onsite AC power to (site-specific emergency 

buses).

AND

b. EITHER of the following: 

• Restoration of at least one AC emergency bus in less than (site-specific hours) 

is not likely.

• (Site-specific indication of an inability to adequately remove heat from the 

core) 

Basis:

This IC addresses a prolonged loss of all power sources to AC emergency buses. A loss of all 

AC power compromises the performance of all SAFETY SYSTEMS requiring electric power 

including those necessary for emergency core cooling, containment heat removal/pressure 

control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will 

lead to a loss of one or more fission product barriers. In addition, fission product barrier 

monitoring capabilities may be degraded under these conditions. 

The EAL should require declaration of a General Emergency prior to meeting the thresholds for 

IC FG1. This will allow additional time for implementation of offsite protective actions.

Escalation of the emergency classification from Site Area Emergency will occur if it is projected 

that power cannot be restored to at least one AC emergency bus by the end of the analyzed 

station blackout coping period. Beyond this time, plant responses and event trajectory are 

subject to greater uncertainty, and there is an increased likelihood of challenges to multiple 

fission product barriers. 

The estimate for restoring at least one emergency bus should be based on a realistic appraisal of 

the situation. Mitigation actions with a low probability of success should not be used as a basis 

for delaying a classification upgrade. The goal is to maximize the time available to prepare for, 

and implement, protective actions for the public. 

The EAL will also require a General Emergency declaration if the loss of AC power results in

parameters that indicate an inability to adequately remove decay heat from the core. 
NEI 99-01 (Revision 6)

November 2012

179

Developer Notes:

Although this IC and EAL may be viewed as redundant to the Fission Product Barrier ICs, 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 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. 

The “site-specific hours” to restore AC power to an emergency bus should be based on the 

station blackout coping analysis performed in accordance with 10 CFR § 50.63 and Regulatory 

Guide 1.155, Station Blackout. 

Site-specific indication of an inability to adequately remove heat 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 exit thermocouple temperature and/or 

reactor vessel water level that drive entry into a core cooling restoration procedure (or otherwise 

requires implementation of prompt restoration actions). Alternately, a site may use incore/core 

exit thermocouple temperatures greater than 1,200o

F and/or a reactor vessel water level that 

corresponds to approximately 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 Owners Group Emergency Response 

Guidelines, enter the parameters used in the Core Cooling Red Path.

ECL Assignment Attributes: 3.1.4.B 
NEI 99-01 (Revision 6)

November 2012

180 

SG8 

ECL: General Emergency

Initiating Condition: Loss of all AC and Vital DC power sources for 15 minutes or longer. 

Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown 

Example Emergency Action Levels: 

Note: The Emergency Director should declare the General Emergency promptly upon 

determining that 15 minutes has been exceeded, or will likely be exceeded.

(1) a. Loss of ALL offsite and ALL onsite AC power to (site-specific emergency 

 buses) for 15 minutes or longer. 

AND

b. Indicated voltage is less than (site-specific bus voltage value) on ALL (sitespecific Vital DC busses) for 15 minutes or longer. 

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. 

Developer 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. 

The “site-specific bus voltage value” should be based on the minimum bus voltage necessary for 

adequate operation of SAFETY SYSTEM equipment. This voltage 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. 

The typical value for an entire battery set is approximately 105 VDC. For a 60 cell string of 

batteries, the cell voltage is approximately 1.75 Volts per cell. For a 58 string battery set, the 

minimum voltage is approximately 1.81 Volts per cell. 

The “site-specific Vital DC busses” are the DC busses that provide monitoring and control 

capabilities for SAFETY SYSTEMS. 
NEI 99-01 (Revision 6)

November 2012

181

This IC and EAL were added to Revision 6 to address operating experience from the March, 

2011 accident at Fukushima Daiichi.

ECL Assignment Attributes: 3.1.4.B 

NEI 99-01 (Revision 6)

November 2012

A-1 

APPENDIX A – ACRONYMS AND ABBREVIATIONS

AC ...................................................................................................................... Alternating Current 

AOP................................................................................................. Abnormal Operating Procedure 

APRM ................................................................................................. Average Power Range Meter

ATWS ................................................................................... Anticipated Transient Without Scram

B&W ................................................................................................................ Babcock and Wilcox

BIIT ..................................................................................... Boron Injection Initiation Temperature 

BWR ............................................................................................................. Boiling Water Reactor

CDE...................................................................................................... Committed Dose Equivalent 

CFR ...................................................................................................... Code of Federal Regulations

CTMT/CNMT ............................................................................................................... Containment 

CSF ............................................................................................................. Critical Safety Function

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

DBA .............................................................................................................. Design Basis Accident

DC .............................................................................................................................. Direct Current

EAL ........................................................................................................... Emergency Action Level

ECCS............................................................................................ Emergency Core Cooling System 

ECL ................................................................................................ Emergency Classification Level

EOF ..................................................................................................Emergency Operations Facility

EOP ............................................................................................... Emergency Operating Procedure

EPA ............................................................................................. Environmental Protection Agency 

EPG ............................................................................................... Emergency Procedure Guideline

EPIP ................................................................................ Emergency Plan Implementing Procedure 

EPR ...................................................................................................... Evolutionary Power Reactor 

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

ERG................................................................................................ Emergency Response Guideline 

FEMA ............................................................................. Federal Emergency Management Agency

FSAR................................................................................................... Final Safety Analysis Report

GE ...................................................................................................................... General Emergency

HCTL .......................................................................................... Heat Capacity Temperature Limit 

HPCI .............................................................................................. High Pressure Coolant Injection 

HSI ............................................................................................................. Human System Interface

IC........................................................................................................................ Initiating Condition

ID ............................................................................................................................. Inside Diameter

IPEEE ............................. Individual Plant Examination of External Events (Generic Letter 88-20) 

ISFSI ........................................................................... Independent Spent Fuel Storage Installation

Keff .................................................................................... Effective Neutron Multiplication Factor

LCO............................................................................................... Limiting Condition of Operation

LOCA ........................................................................................................Loss of Coolant Accident 

MCR .................................................................................................................. Main Control Room 

MSIV.....................................................................................................Main Steam Isolation Valve

MSL ....................................................................................................................... Main Steam Line

mR, mRem, mrem, mREM ............................................................ milli-Roentgen Equivalent Man 

MW ....................................................................................................................................Megawatt

NEI ............................................................................................................. Nuclear Energy Institute 

NPP .................................................................................................................. Nuclear Power Plant

NRC .............................................................................................. Nuclear Regulatory Commission
NEI 99-01 (Revision 6)

November 2012

A-2 

NSSS ................................................................................................. Nuclear Steam Supply System

NORAD ................................................................. North American Aerospace Defense Command 

(NO)UE .......................................................................................... (Notification Of) Unusual Event 

NUMARC9 ............................................................... Nuclear Management and Resources Council

OBE....................................................................................................... Operating Basis Earthquake 

OCA ............................................................................................................. Owner Controlled Area 

ODCM/ODAM ...................................................... Offsite Dose Calculation (Assessment) Manual

ORO ................................................................................................ Off-site Response Organization

PA .............................................................................................................................. Protected Area

PACS.................................................................................... Priority Actuation and Control System 

PAG....................................................................................................... Protective Action Guideline 

PICS ................................................................................. Process Information and Control System 

PRA/PSA .................................... Probabilistic Risk Assessment / Probabilistic Safety Assessment

PWR ........................................................................................................ Pressurized Water Reactor

PS ......................................................................................................................... Protection System

PSIG ................................................................................................. Pounds per Square Inch Gauge 

R ......................................................................................................................................... Roentgen

RCC............................................................................................................ Reactor Control Console 

RCIC ............................................................................................... Reactor Core Isolation Cooling 

RCS ............................................................................................................. Reactor Coolant System

Rem, rem, REM ......................................................................................Roentgen Equivalent Man 

RETS ....................................................................... Radiological Effluent Technical Specifications

RPS ......................................................................................................... Reactor Protection System

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

RVLIS ...................................................................... Reactor Vessel Level Instrumentation System

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

SAR .............................................................................................................. Safety Analysis Report

SAS ........................................................................................................ Safety Automation System

SBO ......................................................................................................................... Station Blackout 

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

SG ...........................................................................................................................Steam Generator

SI .............................................................................................................................. Safety Injection

SICS ................................................................................... Safety Information and Control System 

SPDS ............................................................................................ Safety Parameter Display System

SRO ............................................................................................................ Senior Reactor Operator

TEDE ............................................................................................. Total Effective Dose Equivalent 

TOAF .................................................................................................................. Top of Active Fuel 

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

WOG .................................................................................................. Westinghouse Owners Group 

 

9 NUMARC was a predecessor organization of the Nuclear Energy Institute (NEI).
NEI 99-01 (Revision 6)

November 2012

B-1 

APPENDIX B – DEFINITIONS

The following definitions are taken from Title 10, Code of Federal Regulations, and related 

regulatory guidance documents. 

Alert: Events are in progress or have occurred which involve an actual or potential substantial 

degradation of the level of safety of the plant or a security event that involves probable life 

threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. 

Any releases are expected to be limited to small fractions of the EPA PAG exposure levels.

General Emergency: Events are in progress or have occurred which involve actual or 

IMMINENT substantial core degradation or melting with potential for loss of containment 

integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. 

Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than 

the immediate site area.

Notification of Unusual Event (NOUE)10: Events are in progress or have occurred which indicate 

a potential degradation of the level of safety of the plant or indicate a security threat to facility 

protection has been initiated. No releases of radioactive material requiring offsite response or 

monitoring are expected unless further degradation of safety systems occurs. 

Site Area Emergency: Events are in progress or have occurred which involve actual or likely 

major failures of plant functions needed for protection of the public or HOSTILE ACTION that 

results in intentional damage or malicious acts; 1) toward site personnel or equipment that could 

lead to the likely failure of or; 2) that prevent effective access to, equipment needed for the 

protection of the public. Any releases are 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 99-01 emergency 

classification scheme. 

Emergency Action Level (EAL): A pre-determined, site-specific, observable threshold for an 

Initiating Condition that, when met or exceeded, places the plant in a given emergency 

classification level.

Emergency Classification Level (ECL): One of a set of names or titles established by the US 

Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according 

to (1) potential or actual effects or consequences, and (2) resulting onsite and 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)

Fission Product Barrier Threshold: A pre-determined, site-specific, observable threshold 

indicating the loss or potential loss of a fission product barrier. 

 

10 This term is sometimes shortened to Unusual Event (UE) or other similar site-specific terminology.
NEI 99-01 (Revision 6)

November 2012

B-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. 

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-specific definition for this term.) Developer Note

– The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry 

storage.

CONTAINMENT CLOSURE: (Insert a site-specific definition for this term.) Developer Note – 

The procedurally defined conditions or actions taken to secure containment (primary or 

secondary for BWR) and its associated structures, systems, and components as a functional 

barrier to fission product release under shutdown conditions. 

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. 

FAULTED: The term applied to a steam generator that has a steam leak on the secondary side of 

sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator 

to become completely depressurized. Developer Note – This term is applicable to PWRs only. 

FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive 

belts or overheated electrical equipment do not constitute FIRES. Observation of flame is 

preferred but is NOT required if large quantities of smoke and heat are observed. 

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

by the station.

HOSTILE ACTION: An act toward a NPP 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 NPP. Non-terrorism-based EALs 

should be used to address such activities (i.e., this may include violent acts between individuals 

in the owner controlled area). 

HOSTILE FORCE: One or more individuals who are engaged in a determined assault, overtly or 

by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing 

destruction. 

IMMINENT: The trajectory of events or conditions is such that an EAL will be met within a 

relatively short period of time regardless of mitigation or corrective actions. 
NEI 99-01 (Revision 6)

November 2012

B-3 

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. 

NORMAL LEVELS: As applied to radiological IC/EALs, the highest reading in the past twentyfour hours excluding the current peak value. 

OWNER CONTROLLED AREA: (Insert a site-specific definition for this term.) Developer 

Note – This term is typically taken to mean the site property owned by, or otherwise under the 

control of, the licensee. In some cases, it may be appropriate for a licensee to define a smaller 

area with a perimeter closer to the plant Protected Area perimeter (e.g., a site with a large OCA 

where some portions of the boundary may be a significant distance from the Protected Area). In 

these cases, developers should consider using the boundary defined by the Restricted or Secured 

Owner Controlled Area (ROCA/SOCA). The area and boundary selected for scheme use must 

be consistent with the description of the same area and boundary contained in the Security Plan. 

PROJECTILE: An object directed toward a NPP that could cause concern for its continued 

operability, reliability, or personnel safety. 

PROTECTED AREA: (Insert a site-specific definition for this term.) Developer Note – This 

term is typically taken to mean the 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.) Developer Note – 

This description should include all the cavities, tubes, canals and pools through which irradiated 

fuel may be moved, but not including the reactor vessel. 

RUPTURE(D): The condition of a steam generator in which primary-to-secondary leakage is of 

sufficient magnitude to require a safety injection. Developer Note – 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 

classified as safety-related. Developer Note – This term may be modified to include the 

attributes of “safety-related” in accordance with 10 CFR 50.2 or other site-specific terminology, 

if desired.

SECURITY CONDITION: Any Security Event as 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.

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. 
NEI 99-01 (Revision 6)

November 2012

B-4 

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. 
NEI 99-01 (Revision 6)

November 2012

C-1 

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 

NRC for exemption from specific emergency planning requirements. These exemptions reflect the 

lowered radiological source term and risks associated with spent fuel pool storage relative to reactor 

at-power operation. Source terms and accident analyses associated with plausible accidents are 

documented in the station’s Final Safety Analysis Report (FSAR), as updated. As a result, each 

licensee will need to develop a site-specific emergency classification scheme using the NRCapproved 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 while 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 with 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 pool are the focus of the 

Recognition Category PD ICs and EALs. Radiological effluent IC and EALs were included to 

provide a basis for classifying events that cannot be readily classified based on an observable events 

or plant conditions alone. 

Appropriate ICs and EALs from Recognition Categories A, C, F, H, and S were modified and 

included in Recognition Category PD to address a spectrum of the events that may affect a spent fuel 

pool. The Recognition Category 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 

NRC-approved exemptions, and the source terms and accident analyses specific to the licensee. 

Security-related events will also need to be considered.
NEI 99-01 (Revision 6)

November 2012

C-2 

Table PD-1: Recognition Category “PD” Initiating Condition Matrix

UNUSUAL EVENT ALERT

PD-AU1 Release of gaseous or liquid 

radioactivity greater than 2 times the (site-specific 

effluent release controlling document) limits for 60 

minutes or longer. 

Op. Modes: Not Applicable 

PD-AA1 Release of gaseous or liquid 

radioactivity resulting in offsite dose greater than 10 

mrem TEDE or 50 mrem thyroid CDE. 

Op. Modes: Not Applicable

PD-AU2 UNPLANNED rise in plant 

radiation levels.

Op. Modes: Not Applicable

PD-AA2 UNPLANNED rise in plant radiation 

levels that impedes plant access required to maintain 

spent fuel integrity. 

Op. Modes: Not Applicable

PD-SU1 UNPLANNED spent fuel pool 

temperature rise.

Op. Modes: Not Applicable 

PD-HU1 Confirmed SECURITY 

CONDITION or threat. 

Op. Modes: Not Applicable

PD-HA1 HOSTILE ACTION within the 

OWNER CONTROLLED AREA or airborne attack 

threat within 30 minutes. 

Op. Modes: Not Applicable

PD-HU2 Hazardous event affecting SAFETY 

SYSTEM equipment necessary for spent fuel 

cooling. 

Op. Modes: Not Applicable

PD-HU3 Other conditions exist which in the 

judgment of the Emergency Director warrant 

declaration of a (NO)UE. 

Op. Modes: Not Applicable

PD-HA3 Other conditions exist which in the 

judgment of the Emergency Director warrant 

declaration of an Alert. 

Op. Modes: Not Applicable

Table intended for use by 

EAL developers. 

Inclusion in licensee 

documents is not required. 
NEI 99-01 (Revision 6)

November 2012

C-3 

PD-AU1

ECL: Notification of Unusual Event

Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the (sitespecific effluent release controlling document) limits for 60 minutes or longer. 

Operating Mode Applicability: Not Applicable 

Example Emergency Action Levels: (1 or 2)

Notes:

 The Emergency Director should declare the Unusual Event promptly upon 

determining that 60 minutes has been exceeded, or will likely be exceeded. 

 If an ongoing release is detected and the release start time is unknown, assume that the 

release duration has exceeded 60 minutes. 

 If the effluent flow past an effluent monitor is known to have stopped due to actions to 

isolate the release path, then the effluent monitor reading is no longer valid for 

classification purposes. 

(1) Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint 

established by a current radioactivity discharge permit for 60 minutes or longer. 

(2) 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 lowlevel radiological release that exceeds regulatory commitments for an extended period of time 

(e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored 

or un-monitored, including those for which a radioactivity discharge permit is normally prepared. 

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

effluents to the environment. Further, there are administrative controls established to prevent 

unintentional releases, and to control and monitor intentional releases. The occurrence of an 

extended, uncontrolled radioactive release to the environment is indicative of degradation in 

these features and/or controls. 

Radiological effluent EALs are also included to provide a basis for classifying events and 

conditions that cannot be readily or appropriately classified on the basis of plant conditions 

alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses 

the spectrum of possible accident events and conditions. 

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. 
NEI 99-01 (Revision 6)

November 2012

C-4 

Releases should not be prorated or averaged. For example, a release exceeding 4 times release 

limits for 30 minutes does not meet the EAL.

EAL #1 - This EAL addresses radioactivity releases that cause effluent radiation monitor 

readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL 

will typically be associated with planned batch releases from non-continuous release pathways 

(e.g., radwaste, waste gas).

EAL #2 - This EAL addresses uncontrolled gaseous or liquid releases that are detected by 

sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of 

radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.).

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

Developer Notes:

The “site-specific effluent release controlling document” is the Radiological Effluent Technical 

Specifications (RETS) or, for plants that have implemented Generic Letter 89-0111, the Offsite 

Dose Calculation Manual (ODCM). These documents implement regulations related to effluent 

controls (e.g., 10 CFR Part 20 and 10 CFR Part 50, Appendix I). As appropriate, the RETS or 

ODCM methodology should be used for establishing the monitor thresholds for this IC. 

Listed monitors should include the effluent monitors described in the RETS or ODCM. 

Developers may also consider including installed monitors associated with other potential 

effluent pathways that are not described in the RETS or ODCM1213. If included, EAL values for 

these monitors should be determined using the most applicable dose/release limits presented in 

the RETS or ODCM. It is recognized that a calculated EAL value may be below what the 

monitor can read; in that case, the monitor does not need to be included in the list. Also, some 

monitors may not be governed by Technical Specifications or other license-related related 

requirements; therefore, it is important that the associated EAL and basis section clearly identify

any limitations on the use or availability of these monitors. 

Some sites may find it advantageous to address gaseous and liquid releases with separate EALs. 

Radiation monitor readings should reflect values that correspond to a radiological release 

exceeding 2 times a release control limit. The controlling document typically describes 

methodologies for determining effluent radiation monitor setpoints; these methodologies should 

be used to determine EAL values. In cases where a methodology is not adequately defined, 

developers should determine values consistent with effluent control regulations (e.g., 10 CFR 

Part 20 and 10 CFR Part 50 Appendix I) and related guidance. 

 

11 Implementation of Programmatic Controls for Radiological Effluent Technical Specifications in the 

Administrative Controls Section of the Technical Specifications and the Relocation of Procedural Details of RETS to 

the Offsite Dose Calculation Manual or to the Process Control Program

12 This includes consideration of the effluent monitors described in the site emergency plan section(s) which address 

the requirements of 10 CFR 50.47(b)(8) and (9).

13 Developers should keep in mind the requirements of 10 CFR 50.54(q) and the guidance provided by INPO related 

to emergency response equipment when considering the addition of other effluent monitors. 
NEI 99-01 (Revision 6)

November 2012

C-5 

For EAL #1 - Values in this EAL should be 2 times the setpoint established by the radioactivity 

discharge permit to warn of a release that is not in compliance with the specified limits. 

Indexing the value in this manner ensures consistency between the EAL and the setpoint 

established by a specific discharge permit.

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 with a margin sufficient to ensure that an accurate monitor reading 

is available. For example, an EAL monitor reading might be set at 90% to 95% of the highest 

accurate monitor reading. This provision notwithstanding, 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. 

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 EALs. 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 within 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 system; approval will be 

considered on a case-by-case basis. 

ECL Assignment Attributes: 3.1.1.B 
NEI 99-01 (Revision 6)

November 2012

C-6 

PD-AU2

ECL: Notification of Unusual Event

Initiating Condition: UNPLANNED rise in plant radiation levels.

Operating Mode Applicability: Not Applicable 

Example Emergency Action Levels: (1 or 2)

(1) a. UNPLANNED water level drop in the spent fuel pool as indicated by ANY of the 

following: 

(site-specific level indications). 

AND

 b. UNPLANNED rise in area radiation levels as indicated by ANY of the following 

radiation monitors. 

(site-specific list of area radiation monitors). 

(2) Area radiation monitor reading or survey result indicates an UNPLANNED rise of 25 

mR/hr over NORMAL LEVELS. 

Basis:

This IC addresses elevated plant radiation levels caused by a decrease in water level above 

irradiated (spent) fuel or other UNPLANNED events. The increased radiation levels are 

indicative of a minor loss in the ability to control radiation levels within the plant or radioactive 

materials. Either condition is a potential degradation in the level of safety of the plant. 

A water level decrease will be primarily determined by indications from available level 

instrumentation. Other sources of level indications may include reports from plant personnel or 

video camera observations (if available). 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. Note that EAL #1 is applicable only in 

cases where the elevated reading is due to an UNPLANNED water level drop. EAL #2 excludes 

radiation level increases that result from planned activities such as use of radiographic sources 

and movement of radioactive waste materials.

Escalation of the emergency classification level would be via IC PD-AA1 or PD-AA2.
NEI 99-01 (Revision 6)

November 2012

C-7 

Developer Notes:

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 instrumentation, the declaration may also 

be based on indications of water makeup rate and/or decreases in the level of a water storage 

tank. 

For EAL #2 - The specified value of 25 mR/hr may be set to another value for a specific 

application with appropriate justification. 

ECL Assignment Attributes: 3.1.1.B 
NEI 99-01 (Revision 6)

November 2012

C-8 

PD-SU1

ECL: Notification of Unusual Event

Initiating Condition: UNPLANNED spent fuel pool temperature rise.

Operating Mode Applicability: Not Applicable 

Example Emergency Action Levels: 

(1) UNPLANNED spent fuel pool temperature rise to greater than (site-specific °

F).

Basis:

This IC addresses a condition that is a precursor to a more serious event and represents a 

potential degradation in the level of safety of the plant. If uncorrected, boiling in the pool will 

occur, and result in a loss of pool level and increased radiation levels. 

Escalation of the emergency classification level would be via IC PD-AA1 or PD-AA2. 

Developer Notes:

The site-specific temperature should be chosen based on the starting point for fuel damage 

calculations in the SAR. Typically, this temperature is 125o to 150o F. Spent Fuel Pool 

temperature is normally maintained well below this point thus allowing time to correct the 

cooling system malfunction prior to classification.

ECL Assignment Attributes: 3.1.1.A 
NEI 99-01 (Revision 6)

November 2012

C-9 

PD-HU1

ECL: Notification of Unusual Event

Initiating Condition: Confirmed SECURITY CONDITION or threat. 

Operating Mode Applicability: Not Applicable 

Example Emergency Action Levels: (1 or 2 or 3) 

(1) A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by 

the (site-specific security shift supervision). 

(2) Notification of a credible security threat directed at the site. 

(3) A validated notification from the NRC providing information of an aircraft threat. 

Basis:

This IC addresses events that pose a threat to plant personnel or the equipment necessary to 

maintain cooling of spent fuel, 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 IC PD-HA1.

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, Template for 

the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and 

Independent Spent Fuel Storage Installation Security Program]. 

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 10 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 #3 addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters 

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

The status and size of the plane may also be provided by NORAD through the NRC. Validation 

of the threat is performed in accordance with (site-specific procedure).

Emergency plans and implementing procedures are public documents; therefore, EALs should 

not incorporate Security-sensitive information. This includes information that may be 

advantageous to a potential adversary, such as the particulars concerning a specific threat or 

threat location. Security-sensitive information should be contained in non-public documents 

such as the Security Plan.
NEI 99-01 (Revision 6)

November 2012

C-10 

Escalation of the emergency classification level would be via IC PD-HA1. 

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, 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 

example, an EAL may be worded as “Security event #2, #5 or #9 is reported by the (site-specific 

security shift supervision).” 

ECL Assignment Attributes: 3.1.1.A 
NEI 99-01 (Revision 6)

November 2012

C-11 

PD-HU2

ECL: Notification of Unusual Event

Initiating Condition: Hazardous event affecting SAFETY SYSTEM equipment necessary for 

spent fuel cooling. 

Operating Mode Applicability: Not Applicable 

Example Emergency Action Levels: 

(1) a. The occurrence of ANY of the following hazardous events: 

 Seismic event (earthquake)

 Internal or external flooding event 

 High winds or tornado strike 

 FIRE

 EXPLOSION

 (site-specific hazards)

 Other events with similar hazard characteristics as determined by the Shift 

Manager

 AND

b. The event has damaged at least one train of a SAFETY SYSTEM needed for 

spent fuel cooling. 

 AND 

c. The damaged SAFETY SYSTEM train(s) cannot, or potentially cannot, perform 

its design function based on EITHER: 

 Indications of degraded performance 

 VISIBLE DAMAGE

Basis:

This IC addresses a hazardous event that causes damage to at least one train of a SAFETY 

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 condition reduces 

the margin to a loss or potential loss of the fuel clad barrier, and therefore represents a potential 

degradation of the level of safety of the plant. 

For EAL 1.c, the first bullet addresses damage to a SAFETY SYSTEM train that is in 

service/operation since indications for it will be readily available. 

For EAL 1.c, the second bullet addresses damage to a SAFETY SYSTEM train that is not in 

service/operation or readily apparent through indications alone. 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. 
NEI 99-01 (Revision 6)

November 2012

C-12 

Escalation of the emergency classification level could, depending upon the event, be based on 

any of the Alert ICs; PD-AA1, PD-AA2, PD-HA1 or PD-HA3. 

Developer Notes:

For (site-specific hazards), developers should consider including other significant, site-specific 

hazards to the bulleted list contained in EAL 1.a (e.g., a seiche). 

Nuclear power plant SAFETY SYSTEMS are comprised of two or more separate and redundant 

trains of equipment in accordance with site-specific design criteria. 

ECL Assignment Attributes: 3.1.1.A and 3.1.1C 
NEI 99-01 (Revision 6)

November 2012

C-13 

PD-HU3

ECL: Notification of Unusual Event

Initiating Condition: Other conditions exist which in the judgment of the Emergency Director 

warrant declaration of a (NO)UE. 

Operating Mode Applicability: Not Applicable

Example 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 occurs. 

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.
NEI 99-01 (Revision 6)

November 2012

C-14 

PD-AA1

ECL: Alert

Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater 

than 10 mrem TEDE or 50 mrem thyroid CDE. 

Operating Mode Applicability: Not Applicable 

Example Emergency Action Levels: (1 or 2 or 3 or 4) 

Notes:

 The Emergency Director should declare the Alert promptly upon determining that the 

applicable time has been exceeded, or will likely be exceeded. 

 If an ongoing release is detected and the release start time is unknown, assume that the 

release duration has exceeded 15 minutes. 

 If the effluent flow past an effluent monitor is known to have stopped due to actions to 

isolate the release path, then the effluent monitor reading is no longer valid for classification 

purposes. 

 The pre-calculated effluent monitor values presented in EAL #1 should be used for 

emergency classification assessments until the results from a dose assessment using actual 

meteorology are available. 

(1) Reading on ANY of the following radiation monitors greater than the reading shown for 

15 minutes or longer: 

(site-specific monitor list and threshold values) 

(2) Dose assessment using actual meteorology indicates doses greater than 10 mrem TEDE 

or 50 mrem thyroid CDE at or beyond (site-specific dose receptor point). 

(3) Analysis of a liquid effluent sample indicates a concentration or release rate that would 

result in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond (sitespecific dose receptor point) for one hour of exposure. 

(4) Field survey results indicate EITHER of the following at or beyond (site-specific dose 

receptor point): 

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

or longer. 

 Analyses of field survey samples indicate thyroid CDE greater than 50 mrem for one 

hour of inhalation. 

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 
NEI 99-01 (Revision 6)

November 2012

C-15 

radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled 

release). 

Radiological effluent EALs are also included to provide a basis for classifying events and 

conditions that cannot be readily or appropriately classified on the basis of plant conditions 

alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses 

the spectrum of possible accident events and conditions. 

The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE 

was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. 

Classification based on effluent monitor readings assumes that a release path to the environment 

is established. If the effluent flow past an effluent monitor is known to have stopped due to 

actions to isolate the release path, then the effluent monitor reading is no longer valid for 

classification purposes. 

Developer Notes:

While this IC may not be met absent challenges to the cooling of spent fuel, it provides 

classification diversity and may be used to classify events that would not reach the same ECL 

based on plant conditions alone. 

The EPA PAGs are expressed in terms of the sum of the effective dose equivalent (EDE) and the 

committed effective dose equivalent (CEDE), or as the thyroid committed dose equivalent 

(CDE). For the purpose of these IC/EALs, the dose quantity total effective dose equivalent 

(TEDE), as defined in 10 CFR § 20, is used in lieu of “...sum of EDE and CEDE....”. 

The EPA PAG guidance provides for the use adult thyroid dose conversion factors; however, 

some states have decided to base protective actions on child thyroid CDE. Nuclear power plant 

ICs/EALs need to be consistent with the protective action methodologies employed by the States 

within their EPZs. The thyroid CDE dose used in the IC and EALs should be adjusted as 

necessary to align with State protective action decision-making criteria. 

The “site-specific monitor list and threshold values” should be determined with consideration of 

the following: 

 Selection of the appropriate installed gaseous and liquid effluent monitors. 

 The effluent monitor readings should correspond to a dose of 10 mrem TEDE or 50 mrem 

thyroid CDE at the “site-specific dose receptor point” (consistent with the calculation 

methodology employed) for one hour of exposure. 

 Monitor readings will be calculated using a set of assumed meteorological data or 

atmospheric dispersion factors; the data or factors selected for use should be the same as 

those employed to calculate the monitor readings for IC PD-AU1. 

 The calculation of monitor readings will also require use of an assumed release isotopic mix; 

the selected mix should be the same as that employed to calculate monitor readings for IC 

PD-AU1. 

 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 ECL.
NEI 99-01 (Revision 6)

November 2012

C-16 

The “site-specific dose receptor point” is the distance(s) and/or locations used by the licensee to 

distinguish between on-site and offsite doses. The selected distance(s) and/or locations should 

reflect the content of the emergency plan, and the procedural methodology used to determine 

offsite doses and Protective Action Recommendations. 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 with a margin sufficient to ensure that an accurate monitor reading 

is available. For example, an EAL monitor reading might be set at 90% to 95% of the highest 

accurate monitor reading. This provision notwithstanding, 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 survey 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 EALs. 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 within 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 system; approval will be 

considered on a case-by-case basis. 

ECL Assignment Attributes: 3.1.2.C 
NEI 99-01 (Revision 6)

November 2012

C-17 

PD-AA2

ECL: Alert

Initiating Condition: UNPLANNED rise in plant radiation levels that impedes plant access 

required to maintain spent fuel integrity. 

Operating Mode Applicability: Not Applicable 

Example Emergency Action Levels: (1 or 2)

(1) UNPLANNED dose rate greater than 15 mR/hr in ANY of the following areas requiring 

continuous occupancy to maintain control of radioactive material or operation of systems 

needed to maintain spent fuel integrity: 

(site-specific area list)

(2) UNPLANNED Area Radiation Monitor readings or survey results indicate a rise by 100 

mR/hr over NORMAL LEVELS that impedes access to ANY of the following areas 

needed to maintain control of radioactive material or operation of systems needed to 

maintain spent fuel integrity.

(site-specific area list)

Basis:

This IC addresses increased radiation levels that impede necessary access to areas containing 

equipment that must be operated manually or that requires local monitoring, in order to maintain 

systems needed to maintain spent fuel integrity. As used here, ‘impede’ includes hindering or 

interfering, provided that the interference or delay is sufficient to significantly threaten necessary 

plant access. It is this impaired access that results in the actual or potential substantial 

degradation of the level of safety of the plant.

This IC does not apply to anticipated temporary increases due to planned events. 

Developer Notes:

The value of 15mR/hr is derived from the GDC 19 value of 5 rem in 30 days with adjustment for 

expected occupancy times. Although Section III.D.3 of NUREG-0737, Clarification of TMI 

Action Plan Requirements, provides that the 15 mR/hr value can be averaged over the 30 days, 

the value is used here without averaging, as a 30 day duration implies an event potentially more 

significant than an Alert.

The specified value of 100 mR/hr may be set to another value for a specific application with 

appropriate justification. 

ECL Assignment Attributes: 3.1.2.C 
NEI 99-01 (Revision 6)

November 2012

C-18 

PD-HA1

ECL: Alert

Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or 

airborne attack threat within 30 minutes. 

Operating Mode Applicability: Not Applicable 

Example Emergency Action Levels: (1 or 2)

(1) A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED 

AREA as reported by the (site-specific security shift supervision). 

(2) A validated notification from NRC of an aircraft attack threat within 30 minutes of the 

site. 

Basis:

This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED 

AREA or notification of an aircraft attack threat. This event will require rapid response and 

assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the 

need to prepare the plant and staff for a potential aircraft impact. 

Timely and accurate communications between Security Shift Supervision and the Control Room 

is essential for proper classification of a security-related event.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for 

the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and 

Independent Spent Fuel Storage Installation Security Program]. 

As time and conditions allow, these events require a heightened state of readiness by the plant 

staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). 

The Alert declaration will also heighten the awareness of Offsite Response Organizations, 

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 #1 is applicable for any HOSTILE ACTION occurring, or that has occurred, in the 

OWNER CONTROLLED AREA. This includes any action directed against an ISFSI that is 

located within the OWNER CONTROLLED AREA.

EAL #2 addresses the threat from the impact of an aircraft on the plant, and the anticipated 

arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related 

notifications are made in a timely manner so that plant personnel and OROs are in a heightened 
NEI 99-01 (Revision 6)

November 2012

C-19 

state of readiness. This EAL is met when the threat-related information has been validated in 

accordance with (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 

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.

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.

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 

example, an EAL may be worded 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 OWNER CONTROLLED AREA. 

ECL Assignment Attributes: 3.1.2.D 
NEI 99-01 (Revision 6)

November 2012

C-20 

PD-HA3

ECL: Alert

Initiating Condition: Other conditions exist which in the judgment of the Emergency Director 

warrant declaration of an Alert.

Operating Mode Applicability: Not Applicable 

Example 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 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. 

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.