Revision 01/2003

NEI 99-01

Rev. 4

(NUMARC/NESP-007)

Methodology for Development

of Emergency Action Levels

January 2003
Revision 01/2003

ACKNOWLEDGEMENTS

Revision 4 of this report incorporates new Emergency Action Levels (EALs) for Cold Shutdown 

and Refueling modes, Independent Spent Fuel Storage Installations (ISFSI), permanently

Defueled Stations, and Security EAL changes post September 11. The EAL changes are based 

on numerous suggestions provided by utilities and input provided by the staff of the NRC. NEI 

acknowledges the valuable input provided, and the extensive technical support provided by the 

members of the EAL Task Force.

NOTICE

This report was prepared as an account of work sponsored by the Nuclear Energy Institute 

(NEI). Neither NEI nor any of its employees, members, or consultants make any warranty, 

expressed or implied, or assume any legal liability or responsibility for the accuracy,

completeness or usefulness of any information, apparatus, product or process disclosed in this 

report, or represent that its use would not infringe privately-owned rights.

The opinions, conclusions, and recommendations set forth in this report are those of the authors 

and do not necessarily represent the views of NEI, its employees, members or consultants.

Because NEI is supported in part by Federal funds, NEI’s activities are subject to Title VI of the 

Civil Rights Act of 1964, which prohibits discrimination based on race, color, or national origin, 

and other federal laws and regulations, rules, and orders issued thereunder prohibiting

discrimination. Written complaints of exclusion, denial of benefits or other discrimination of 

those bases under this program may be filed with the United States Nuclear Regulatory

Commission, Washington, DC 20555 or any other appropriate federal regulatory agency or, 

among others, the Tennessee Valley Authority (TVA), Office of Equal Employment Opportunity, 

400 West Summit Hill Drive, Knoxville, TN 37902.
Revision 01/2003

FOREWORD

Revision 4 to NUMARC/NESP-007 presents the methodology for development of emergency 

action levels as an alternative to NRC/FEMA guidelines contained in Appendix 1 of NUREG0654/FEMA-REP-1, Rev. 2 "Criteria for Preparation and Evaluation of Radiological Emergency 

Response Plans and Preparedness in Support of Nuclear Power Plants," October 1980 and 

10 CFR 50.47 (a)(4). Revision 4 (NEI 99-01) enhances Revision 3 (NEI 97-03) by consolidating 

the system malfunction initiating conditions and example emergency action levels which address 

conditions that may be postulated to occur at nuclear power plants during plant shutdown 

conditions (Recognition Category C). Also included are initiating conditions and example

emergency action levels that fully address conditions that may be postulated to occur at

permanently Defueled Stations (Recognition Category D) and Independent Spent Fuel Storage 

Installations (Recognition Category E).

Recognition Category C, D, and E initiating conditions and associated emergency action levels 

were written so that they could be implemented by both NUMARC/NESP-007 and NUREG0654/FEMA-REP-1 users. As described in Appendix B, the industry anticipates that the NRC will 

provide written position so that NUREG-0654/FEMA-REP-1 users may implement Recognition 

Category C, D, and E even though they may have chosen to not fully implement the

NUMARC/NESP-007 methodology.

Revision 4 was originally submitted to NRC for review and endorsement in August of 2000. 

Endorsement was delayed due to concerns related to Recognition Category D and E. 

Revision 4 was resubmitted to NRC in September of 2002 with request for endorsement per 

Regulatory Guide 1.101 regardless of Recognition Category D and E. Changes to the

September 2002 submittal of Revision 4 were limited as described below:

• Modifications as agreed upon in the August 8, 2000 letter from NRR to NEI to include:

• Rewording of the Table 5-F-1 GE criteria (Comment 5)

• Deletion of IC E-AU1 (Comment 11)

• Basis wording changes to IC E-HU-1 and Appendix E (Comment 12 and 13)

• Modifies HU4 to incorporate post September 11 security EAL changes based on the 

October 6, 2001 Safeguards Advisory Notice and the NRC’s November 6, 2001

Information Assessment Team Recommended Actions in Response to a Site-specific 

Credible Threat at a Nuclear Power Plant. These security EAL changes were endorsed 

by letter from NRR to NEI dated February 4, 2002. 

• Modifies EAL 1 associated with HU3 and HA3 to more closely describe the IC based on 

operating experience. 

In November 2002, NEI met with the NRC to discuss questions that developed from NRC’s 

review of the September 2002 submittal. Changes to the September 2002 submittal were

requested by NRC based on this meeting and additional reviews that were conducted in

December 2002. Following incorporation of requested NRC changes, NEI 99-01 Revision 4 was 

resubmitted by NEI to NRC in January of 2003 with the request for full Regulatory Guide 1.101 

endorsement. The only substantive changes incorporated into the January 2003 submittal were

• Minor editing as outlined in the November 2002 NEI to NRC meeting handout

• Removal of the September 2002 changes to EAL 1 of HU3 and HA3 described above.
Revision 01/2003

EAL Revision Task Force

1996-2000

Chris Boone Southern Nuclear Operating Company

Rodney Brown Duke Power Company

Gary Cerkas Arizona Public Service Company

John Costello Virginia Power

Dennis M. Emborsky Northeast Utilities

James D. Jones Niagara Mohawk Power Corporation

Ron Jorgensen Washington Public Power Supply System

John Kaminski Niagara Mohawk Power Corporation

Stephen F. LaVie Duquesne Light Company

Walter H. Lee Southern Nuclear Operating Company

Mark Luksic Tennessee Valley Authority

Alan Nelson Nuclear Energy Institute

Kevin Morris Detroit Edison Company

David W. Stobaugh Commonwealth Edison Company

Martin Vonk Commonwealth Edison Company

Additional technical assistance on Shutdown EALs was provided by the following individuals:

Roy K. Brosi Duquesne Light Company

George W. Busch GPU Nuclear

Neil Coddington Pennsylvania Power & Light Company

Mitchell S. Dior Public Service Electric & Gas Company

Albert L. Garrou Carolina Power & Light

Larry E. Gifford GE Nuclear Energy

John Jenkins Duke Power Company

John F. Lohr Baltimore Gas & Electric Company

Robert D. Mandik PECO Energy Company

Frank L. Pavlechko Tennessee Valley Authority

Charles Sawyer Duke Power Company

Liaison to the Task Force

Charles Miller U.S. Nuclear Regulatory Commission

Frederick Hasselberg U.S. Nuclear Regulatory Commission

James O’Brien U.S. Nuclear Regulatory Commission

Randy Sullivan U.S. Nuclear Regulatory Commission

Serge Roudier U.S. Nuclear Regulatory Commission

P. A. Milligan U.S. Nuclear Regulatory Commission
Revision 01/2003

The following persons participated in the development of Revision 2 of the NUMARC/NESP-007 

document.

EAL Revision Task Force

Reginald C. Rodgers, Chairman Northeast Utilities

Thomas B. Blount GPU Nuclear Corporation

Roy K. Brosi Duquesne Light Company

Nick S. Catron Tennessee Valley Authority

David L. Fugere Consumers Power Company

Albert L. Garrou Carolina Power & Light Company

George J. Giangi GPU Nuclear Corporation

Terry Gilman Commonwealth Edison Company

Ronald E. Harris Duke Power Company

Roger Hoyt Wolf Creek Nuclear Operating Corporation

John Jenkins Duke Power Company

Stephen F. LaVie Duquesne Light Company

Theresa M. Lechton Commonwealth Edison Company

James M. Minneman Pennsylvania Power & Light Company

Kevin Moles Wolf Creek Nuclear Operating Corporation

Patrick Taylor Pennsylvania Power & Light Company

Alan P. Nelson, Project Manager NUMARC

Independent Industry Review Group

Craig Adams Public Service Electric & Gas Company

Rod Krieger Indiana & Michigan Electric Company

Milton Stiller Union Electric Company

Liaison to the Task Force

Frederick Hasselberg U.S. Nuclear Regulatory Commission

Michael Jamgochian U.S. Nuclear Regulatory Commission

Aby Mohseni U.S. Nuclear Regulatory Commission

Edward Podolak U.S. Nuclear Regulatory Commission

William Reckley U.S. Nuclear Regulatory Commission

Craig Wingo Federal Emergency Management Agency

Edison Carmack Institute of Nuclear Power Operations

ERC Environmental and Energy Services Company (ERCE) Study 

Team

John S. Fuoto, Study Team Leader
Revision 01/2003 i

TABLE OF CONTENTS

PAGE

EXECUTIVE SUMMARY...................................................................................................................v

ACRONYMS...................................................................................................................................viii

1.0 METHODOLOGY FOR DEVELOPMENT OF EMERGENCY

ACTION LEVELS.............................................................................................................1.1

1.1 Background...........................................................................................................1.1

2.0 CHANGES INCORPORATED WITH REVISION 4 ........................................................2.1

3.0 DEVELOPMENT OF BASIS FOR GENERIC APPROACH...........................................3.1

3.1 Regulatory Context................................................................................................3.1

3.2 Definitions Needed To Develop EAL Methodology................................................3.3

3.3 Differences In Perspective....................................................................................3.4

3.4 Recognition Categories.........................................................................................3.4

3.5 Design Differences ...............................................................................................3.6

3.6 Required Characteristics ......................................................................................3.6

3.7 Emergency Class Descriptions ............................................................................3.7

3.8 Emergency Class Thresholds ..............................................................................3.8

3.9 Emergency Action Levels ...................................................................................3.10

3.10 Treatment Of Multiple Events and Emergency Class Upgrading.......................3.11

3.11 Emergency Class Downgrading.........................................................................3.12

3.12 Classifying Transient Events ..............................................................................3.13

3.13 Interface Between Classification and Activation of Emergency Facilities ..........3.13

3.14 Cold Shutdown/Refueling IC/EALs .....................................................................3.14

3.15 Permanently Defueled Station IC/EALs ..............................................................3.15

3.16 ISFSI IC/EALs......................................................................................................3.15

3.17 Operating Mode Applicability...............................................................................3.15

3.17.1 Mode Applicability Matrix......................................................................................3.16

3.17.2 BWR Operating Modes.......................................................................................3.16

3.17.3 PWR Operating Mode.........................................................................................3.17

4.0 HUMAN FACTORS CONSIDERATIONS.......................................................................4.1

4.1 Level of Integration of EALs with Plant Procedures..............................................4.1

4.2 Method of Presentation .........................................................................................4.2

4.3 Symptom-Based, Event-Based or Barrier-Based EALs ......................................4.2
Revision 01/2003 ii

TABLE OF CONTENTS (continued)

PAGE

5.0 GENERIC EAL GUIDANCE.............................................................................................5.1

5.1 Generic Arrangement............................................................................................5.1

5.2 Generic Bases ......................................................................................................5.2

5.3 Site Specific Implementation.................................................................................5.3

5.4 Definitions..............................................................................................................5.3

INITIATING CONDITIONS MATRICES

Category A Abnormal Rad Levels/Radiological Effluent......................................................5-A-1

Category C Cold Shutdown./ Refueling System Malfunction.............................................. 5-C-1

Category D Defueled Station Malfunction............................................................................ 5-D-1

Category E Events Related to Independent Spent Fuel Storage Installations.....................5-E-1

Category F Fission Product Barrier Degradation ................................................................5-F-1

Category H Hazards and Other Conditions Affecting Plant Safety......................................5-H-1

Category S System Malfunction...........................................................................................5-S-1

APPENDIX A BASIS FOR RADIOLOGICAL EFFLUENT INITIATING CONDITIONS .......A.1

APPENDIX B BASIS FOR IMPLEMENTATION OF CATEGORY C, D, AND E 

INITIATING CONDITIONS BY NUREG-0654/FEMA-REP-1 USERS...........B-1

APPENDIX C BASIS FOR COLD SHUTDOWN/REFUELING INITIATING CONDITION....C-1

APPENDIX D BASIS FOR PERMANENTLY DEFUELED STATION INITIATING 

CONDITIONS..................................................................................D-1

APPENDIX E BASIS FOR ISFSI INITIATING CONDITIONS........................................E-1
Revision 01/2003 iii

TABLE OF CONTENTS (continued)

PAGE

TABLES

TABLE 5-A-1: Recognition Category A Initiating Condition Matrix............................................5-A-1

TABLE 5-C-1 Recognition Category C Initiating Condition Matrix.......................................... 5-C-1

TABLE 5-D-1 Recognition Category D Initiating Condition Matrix.......................................... 5-D-1

TABLE 5-E-1 Recognition Category E Initiating Condition Matrix ...........................................5-E-1

TABLE 5-F-1 Recognition Category F Initiating Condition Matrix............................................5-F-1

TABLE 5-F-2 BWR Fission Product Barrier Reference Table...............................................5-F-2

TABLE 5-F-3 PWR Fission Product Barrier Reference Table...............................................5-F-9

TABLE 5-H-1: Recognition Category H Initiating Condition Matrix ...........................................5-H-1

TABLE 5-S-1 Recognition Category S Initiating Condition Matrix ...........................................5-S-1
Revision 01/2003 iv

EXECUTIVE SUMMARY

Nuclear utilities must respond to a formal set of threshold conditions that require plant personnel to 

take specific actions with regard to notifying state and local governments and the public when certain 

off-normal indicators or events are recognized. Emergency classes are defined in 10 CFR 50. Levels 

of response and the conditions leading to those responses are defined in a joint NRC/FEMA guidelines 

contained in Appendix 1 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," October 1980. 

In 1988, A NUMARC/NESP project was initiated to re-evaluate the emergency action levels (EALs) in 

the context of utility operating experience. At that time, the nuclear utility industry had over ten years of 

experience in adapting the NRC guidelines to specific plant configurations, using them both in

exercises and under actual emergency conditions. As a result, a number of improvements had been 

identified as NUREG-0654, Appendix 1 guidelines had been applied in the development of plant EALs.

The NUMARC/NESP EAL Task Force developed a systematic approach and supporting basis for EAL 

development. This methodology developed a set of generic EAL guidelines, together with the basis for 

each, such that they could be used and adapted by each utility on a consistent basis. The review of 

the industry's experiences with EALs, in conjunction with regulatory considerations, was applied

directly to the development of this generic set of EAL guidelines. The generic guidelines were intended 

to clearly define conditions that represent increasing risk to the public and can give consistent

classifications when applied at different sites. The NUMARC/NESP-007 document resulted from that 

effort. The draft NUMARC/NESP-007 methodology was reviewed by individuals from the industry, 

independent of the task force, was submitted to the entire industry for review, was exercised in a table 

top exercise with the NRC, underwent a regulatory analysis by the NRC, was published for public 

comment in the Federal Register, and was endorsed by the NRC as an acceptable alternative to the 

guidance in NUREG-0654 in Revision 3 to Regulatory Guide 1.101, “Emergency Planning and

Preparedness for Nuclear Power Reactors”. The methodology was presented to the industry in a 

workshop conducted in St Louis in September 1992.

Close to the end of the process described above, concerns developed regarding the classification of 

events which occur during periods of plant shutdowns and refueling. Industry experience had shown 

that plants could be susceptible to a variety of events that could challenge safety during shutdown 

operations. While these events had neither posed nor indicated an undue risk to public health and 

safety, they did indicate the need to consider emergency action levels applicable during shutdown 

modes. Since the issue was still under evaluation, shutdown EALs were not included in Revision 2, but 

were deferred to a later revision of NUMARC/NESP-007. A special task force was formed to address 

this issue and draft shutdown EALs were prepared in conjunction with efforts of the NUMARC

Shutdown Plant Issues Working Group to coordinate industry activities relating to shutdown safety.

As utilities implemented the NUMARC/NESP-007 areas of possible improvement were identified. In 

addition, the staff of the NRC provided suggestions for improvement based on their review of utility 

submittals. A task force was assembled to incorporate the implementation experiences. NEI 97-03, 

Revision 3, was the successor to NUMARC/NESP-007 that incorporated these implementation

experiences

The special task force that was formed to address EALs associated with shutdown plant issues also 

was assigned the task of addressing the need for EALs that relate to permanently defueled stations 

and 10 CFR 72.32 (c) independent spent fuel storage installations. NEI 99-01, Revision 4, is the 

successor to NEI 97-03 that addresses all of these issues.
Revision 01/2003 v

The guidance presented here is not intended to be applied to plants as-is. It is intended to give the 

user the logic for developing site-specific EALs (i.e., instrument readings, etc.) using site-specific EAL 

presentation methods (formats). Basis information is provided to aid station personnel in preparation 

of their own site-specific EALs, to provide necessary information for training, and for explanation to 

state and local officials. In addition, state and local requirements have not been reflected in the generic 

guidance and should be considered on a case-by-case basis with appropriate state and local

emergency response organizations. 

It is important that the NEI EALs be treated as an integrated package. Selecting only portions of this 

guidance for use in developing site-specific EALs could lead to inconsistent or incomplete EALs 

unless explicitly allowed. An example of such an allowance may be found in the NRC’s Branch 

Technical Position Paper dated 7/11/94. As discussed in Appendix B, the industry anticipates that the 

NRC may endorse similar Branch Technical Position guidance for implementation of Recognition 

Category C, D, and E initiating conditions by NUREG-0654/FEMA-REP-1 users who have chosen not 

to implement NEI EALs. Note that the Branch Technical Position was subsequently incorporated into 

EPPOS 1.

Although the basic concerns with barrier integrity and the major safety problems of nuclear power 

plants are similar across plant types, design differences will have a substantial effect on EALs. The 

major differences are found between a BWR and a PWR. In these cases, EAL guidelines unique to 

BWRs and PWRs must be specified. Even among PWRs, however, there are substantial differences 

in design and in types of containment used. There is enough commonality among plants that many 

ICs will be the same or very similar. However, others will have to match plant features and safety 

system designs that are unique to the plant type or even to the specific plant. The EAL Task Force 

believes that there is sufficient information provided in the basis of the EALs to allow the EALs to be 

implemented at plants from all NSSS LWR vendors. However, this generic guidance is not considered 

to be applicable to advanced LWR designs or to away from site radioactive material storage facilities.

The original EAL Task Force identified eight characteristics that were to be incorporated into model 

EALs. Experience to date has shown these considerations to be VALID. These were:

(1) Consistency (i.e., the EALs would lead to similar decisions under similar circumstances at 

different plants);

(2) Human engineering and user friendliness;

(3) Potential for classification upgrade only when there is an increasing threat to public health and 

safety;

(4) Ease of upgrading and downgrading;

(5) Thoroughness in addressing, and disposing of, the issues of completeness and accuracy 

raised regarding NUREG-0654, Appendix 1;

(6) Technical completeness and appropriateness for each classification level;

(7) A logical progression in classification for combinations of multiple events;

(8) Objective, observable values.
Revision 01/2003 vi

Based on the information gathered and reviewed, the Task Force has developed generic EAL

guidance. Because of the wide variety of presentation methods (formats) used at different utilities, the 

Task Force believes that specifying guidance as to what each IC and EAL should address, and 

including sufficient basis information for each EAL will best assure uniformity of approach. The 

information is presented by Recognition Category:

ü A - Abnormal Rad Levels/Radiological Effluent

ü C - Cold Shutdown./ Refueling System Malfunction

ü D – Permanently Defueled Station Malfunction

ü E - Events Related to Independent Spent Fuel Storage Installations (ISFSI)

ü F - Fission Product Barrier Degradation

ü H - Hazards and Other Conditions Affecting Plant Safety

ü S - System Malfunction

Each of the EAL guides in Recognition Categories A, C, D, E, H, and S is structured in the following 

way:

ü Recognition Category - As described above.

ü Emergency Class - NOUE, Alert, Site Area Emergency or General Emergency.

ü Initiating Condition - Symptom- or Event-Based, Generic Identification and Title.

ü Operating Mode Applicability - Power Operation, Hot Standby, Hot Shutdown, Cold Shutdown, 

Refueling, Defueled, All, or Not Applicable.

ü Example Emergency Action Level(s) corresponding to the IC.

ü Basis information for plant-specific readings and factors that may relate to changing the generic IC 

or EAL to a different emergency class, such as for Loss of All AC Power.

For Recognition Category F, the EAL information is presented in a matrix format. The presentation 

method was chosen to clearly show the synergism among the EALs and to support more accurate 

dynamic assessments. For category F, the EALs are arranged by safety function, or fission product 

barrier. Classifications are based on various combinations of function or barrier challenges.

The EAL Guidance has the primary threshold for NOUE as operation outside the safety envelope for 

the plant as defined by plant technical specifications, including LCOs and Action Statement Times. In 

addition, certain precursors of more serious events such as loss of offsite AC power and earthquakes 

are included in NOUE EALs. This provides a clear demarcation between the lowest emergency class 

and "non-emergency" notifications specified by 10 CFR 50.72.
Revision 01/2003 vii

ACRONYMS

AC Alternating Current

AEOD NRC Office for Analysis and Evaluation of Operational Data

ATWS Anticipated Transient Without Scram

B&W Babcock and Wilcox

BWR Boiling Water Reactor

CCW Component Cooling Water

CDE Committed Dose Equivalent

CE Combustion Engineering

CFR Code of Federal Regulations

CMT Containment

CSF Critical Safety Function

CSFST Critical Safety Function Status Tree

DC Direct Current

DHR Decay Heat Removal

DOT Department of Transportation

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

EPRI Electric Power Research Institute

ERG Emergency Response Guideline

ESF Engineered Safeguards Feature

ESW Emergency Service Water

FEMA Federal Emergency Management Agency

FSAR Final Safety Analysis Report

GE General Electric

HPCI High Pressure Coolant Injection

HPSI High Pressure Safety Injection

IC Initiating Condition

IDLH Immediately Dangerous to Life and Health

IPEEE Individual Plant Examination of External Events (Generic Letter 88-20)

ISFSI Independent Spent Fuel Storage Installation 
Revision 01/2003 viii

ACRONYMS (continued)

LCO Limiting Condition of Operation

LER Licensee Event Report

LFL Lower Flammability Limit

LOCA Loss of Coolant Accident

LPSI Low Pressure Safety Injection

LWR Light Water Reactor

MSIV Main Steam Isolation Valve

mR milliRem

Mw Megawatt

NEI Nuclear Energy Institute

NESP National Environmental Studies Project

NRC Nuclear Regulatory Commission

NSSS Nuclear Steam Supply System

NOUE Notification Of Unusual Event

NUMARC Nuclear Management and Resources Council

OBE Operating Basis Earthquake

ODCM Offsite Dose Calculation Manual

PRA/PSA Probabilistic Risk Assessment / Probabilistic Safety Assessment

PWR Pressurized Water Reactor

PSIG Pounds per Square Inch Gauge

R Rem

RCIC Reactor Core Isolation Cooling

RCS Reactor Coolant System

RPS Reactor Protection System

RPV Reactor Pressure Vessel

RVLIS Reactor Vessel Level Indicating System

SBGTS Stand-By Gas Treatment System

SG Steam Generator

SI Safety Injection

SPDS Safety Parameter Display System

SRO Senior Reactor Operator

SSE Safe Shutdown Earthquake

TEDE Total Effective Dose Equivalent

TOAF Top of Active Fuel

TSC Technical Support Center

WE Westinghouse Electric

WOG Westinghouse Owners Group
Revision 01/2003

1.1

1.0 METHODOLOGY FOR DEVELOPMENT OF EMERGENCY ACTION 

LEVELS

1.1 Background

Nuclear utilities must respond to a formal set of threshold conditions that require plant personnel to 

take specific actions with regard to notifying state and local governments and the public when 

certain off-normal indicators or events are recognized. Emergency classes are defined in 10 CFR 

50. Levels of response and the conditions leading to those responses are defined in a joint 

NRC/FEMA guideline contained in Appendix 1 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," October 1980. 

In 1988, A NUMARC/NESP project was initiated to re-evaluate the emergency action levels (EALs) 

in the context of utility operating experience. At that time, the nuclear utility industry had over ten 

years of experience in adapting the NRC guidelines to specific plant configurations, using them both 

in exercises and under actual emergency conditions. As a result, a number of improvements had 

been identified as NUREG-0654, Appendix 1. Guidelines have been applied in the development of 

plant EALs.

The NUMARC/NESP EAL Task Force developed a systematic approach and supporting basis for 

EAL development. This methodology developed a set of generic EAL guidelines, together with the 

basis for each, such that they could be used and adapted by each utility on a consistent basis. The 

review of the industry's experiences with EALs, in conjunction with regulatory considerations, was 

applied directly to the development of this generic set of EAL guidelines. The generic guidelines 

were intended to clearly define conditions that represent increasing risk to the public and can give 

consistent classifications when applied at different sites. The NUMARC/NESP-007 document 

resulted from that effort. The draft NUMARC/NESP-007 methodology was reviewed by individuals 

from the industry, independent of the task force, was submitted to the entire industry for review, 

was exercised in a table top exercise with the NRC, underwent a regulatory analysis by the NRC, 

was published for public comment in the Federal Register, and was endorsed by the NRC as an 

acceptable alternative to the guidance in NUREG-0654 in Revision 3 to Regulatory Guide 1.101, 

“Emergency Planning and Preparedness for Nuclear Power Reactors”. The methodology was 

presented to the industry in a workshop conducted in St Louis in September 1992.

As utilities implemented the NUMARC/NESP-007, areas of possible improvement were identified. In 

addition, the staff of the NRC provided suggestions for improvement based on their review of utility 

submittals. A task force was assembled to incorporate the improvements. NEI-97-03, Revision 3, 

was the successor to NUMARC/NESP-007 that incorporated these improvements.

Close to the end of the process of developing Revision 2 described above, concerns developed 

regarding the classification of events which occur during periods of plant shutdowns and refueling. 

Industry experience had shown that plants could be susceptible to a variety of events that could 

challenge safety during shutdown operations. While these events had neither posed nor indicated 

an undue risk to public health and safety, they did indicate the need to consider emergency action 

levels applicable during shutdown modes. Since the issue was still under evaluation, shutdown 

EALs were not included in Revision 2 or 3 but were deferred. Guidance which addresses cold 

shutdown/refueling, permanently defueled, and independent spent fuel storage EALs have been 

included as part of NEI 99-01. NEI 99-01 addresses both NUMARC/NESP-007 and NUREG-0654 

users for these important issues. 
2. 1

Revision 01/2003

2.0 CHANGES INCORPORATED WITH REVISION 4

This section summarizes the more significant changes made to the EAL methodology with Revision 

4. This is not intended to be a complete tabulation. Minor editorial changes were made in the interest 

of clarity and/or consistent formatting. These changes are not tabulated herein.

2.1 Section 3.0, Development of Basis for Generic Approach

Discussion was added to make recommendations regarding (1) Cold Shutdown/Refueling IC/EALs, 

(2) Permanently Defueled Station IC/EALs, and (3) Independent Spent Fuel Storage Installations 

(ISFSI) IC/EALs. 

2.2 Section 4.0, Human Factors Considerations

No significant changes.

2.3 Section 5.0, Generic EAL Guidance

Discussion was added concerning: (1) Cold Shutdown/Refueling IC/EALs, (2) Permanently

Defueled Station IC/EALs, and (3) ISFSI IC/EALs. 

Additional information regarding site-specific implementation was added in response to numerous 

questions received during utility implementation efforts.

The definitions section was revised to incorporate new terms that relate to Cold

Shutdown/Refueling, Permanently Defueled Station, and ISFSI issues. These words and phrases 

are defined terms having specific meanings as they relate to the EALs. These terms appear in 

capital letters in the IC/EALs, and bases

Some of the Revision 3 Recognition Category S IC/EALs that addressed shutdown events have 

been incorporated into the new Recognition Category C. The EALs affected include SU1, SU4, 

SU5, SU6, SU7, SU8, SA1, SA3, and SS5. EALs SU7, SA1, SA3, and SS5 have been deleted. In 

order to preserve consistency with Revision 3, the IC designations, e.g., AU1, SS1, etc., have not 

been revised. Because of this, there are gaps in the IC designation sequences. The initiating 

condition matrices for each recognition category were re-arranged slightly to align event

progressions where possible. While the individual ICs are presented in sequence by IC designator, 

the IC entries in the initiating condition matrices may not be in sequence.

2.4 Section 5.0, Recognition Category A

No change in the philosophy of classifying abnormal radiological effluent events was incorporated in 

Revision 4. Users should note that all Recognition Category A IC/EALs are applicable for all

operating modes including the cold shutdown and refueling modes. The Category A IC/EALs are 

not applicable for Permanently Defueled stations nor are they applicable for potential releases 

associated with ISFSIs. Separate Radiological effluent IC/EALs have been included in Section D 

and E to address potential effluent releases or radiological concerns. Initiating Conditions D-AU1, 

D-AU2, D-AA1, and D-AA2 were added to Recognition Category D. Initiating Condition E-AU1 was 

added to Recognition Category E.
2. 2

Revision 01/2003

2.5 Section 5.0, Recognition Category C

Recognition Category C is a new category of IC/EALs which completely replaces Recognition 

Category S when in Cold Shutdown and Refueling modes. As discussed previously, some of the 

Revision 3 Recognition Category S IC/EALs that addressed shutdown events have been

incorporated into the Recognition Category C. The following Category S IC/EALs were included in 

Category C: SU1 (CU3), SU4 (CU5), SU5 (CU1), SU6 (CU6), SU7 (CU7), SU8 (CU8), SA1 (CA3), 

SA3 (CA4), and SS5. (CS1 and CS2). In order to adequately address shutdown loss of inventory 

and loss of decay heat removal capability events the following new IC/EALs were added: CU2 

(Unplanned Loss of inventory – Refueling), CU4 (Unplanned Loss of Decay Heat Removal

Capability - Cold Shutdown and Refueling), CA1 (Loss of RCS Inventory – Cold Shutdown), CA2 

(Loss of RPV Inventory – Refueling), and CG1 (Loss of RPV Inventory Affecting Fuel Integrity with 

Containment Challenged – Cold Shutdown and Refueling).

Appendix C was added to provide a common location for describing the basis for the Recognition 

Category C IC/EALs.

2.6 Section 5.0, Recognition Category D

Recognition Category D is a new category that provides IC/EALs for Permanently Defueled

stations. Category D was written to provide a stand alone set of IC/EALs for Permanently Defueled 

Stations. IC/EALs from Recognition Category A, C, F, S, and H were reviewed for applicability and 

where applicable have been included to address all Permanently Defueled station events.

Appendix D was added to provide a common location for describing the basis for the Recognition 

Category D IC/EALs.

2.7 Section 5.0, Recognition Category E

Recognition Category E is a new category that provides IC/EALs for events related to Independent 

Spent Fuel Storage Installations (ISFSI). Category E was written to provide a stand alone set of 

IC/EALs for sites having ISFSI. IC/EALs from Recognition Category A, C, F, S, and H were

reviewed for applicability and where applicable have been included to address all events related to 

the ISFSI.

Appendix E was added to provide a common location for describing the basis for the Recognition 

Category E IC/EALs.

2.8 Section 5.0, Recognition Category F

No significant changes were made.

2.9 Section 5.0, Recognition Category H

No significant changes were made.
2. 3

Revision 01/2003

2.10 Section 5.0, Recognition Category S

Some of the Revision 3 Recognition Category S IC/EALs that addressed shutdown events have 

been incorporated into the new Recognition Category C. The EALs affected include SU1, SU4, 

SU5, SU6, SU7, SU8, SA1, SA3, and SS5. EALs SU7, SA1, SA3, and SS5 have been deleted. In 

order to preserve consistency with Revision 3, the IC designations, e.g., AU1, SS1, etc., have not 

been revised. Because of this, there are gaps in the IC designation sequences. The initiating 

condition matrices for each recognition category were re-arranged slightly to align event

progressions where possible. While the individual ICs are presented in sequence by IC designator, 

the IC entries in the initiating condition matrices may not be in sequence.
Revision 01/2003 3.1

3.0 DEVELOPMENT OF BASIS FOR GENERIC APPROACH

This section addresses several key considerations that were incorporated into the development of 

the original NUMARC/NESP EALs. An understanding of these considerations will facilitate the 

implementation of this generic guidance into site-specific programs. In prior revisions to this 

document, this section described the process by which the Task Force identified and resolved 

these considerations. Since much of this was deemed to be historical in nature, it has been 

removed from this revision.

Literature reviews, review of plant-specific EALs, and on-site utility interviews were performed as 

preparation for the drafting of the generic guidance. The review led to the conclusion that the current 

regulatory structure was not an impediment to the development of the appropriate EALs. Rather, 

the detailed guidance currently in place could be enhanced.

The generic guidance provided in this document is intended to address radiological emergency 

preparedness. Non-radiological events are included in the classification scheme only to the extent 

that these events represent challenges to the continued safety of the reactor plant and its operators. 

There are existing reporting requirements (EPA, OSHA) under which utilities operate. There are 

also requirements for emergency preparedness involving hazardous chemical releases. While the 

proposed classification structure could be expanded to include these non-radiological hazards, 

these events are beyond the scope of this document.

This classification scheme is based on the four classification levels promulgated by the NRC as the 

standard for the United States. This scheme is different from the international severity scale, which 

is not addressed in this generic guidance. The NRC has determined that US nuclear facilities would 

continue to classify events using the four classification levels and that the NRC would re-classify 

the event in any international communication.

3.1 Regulatory Context

Title 10, Code of Federal Regulations, Part 50 provides the regulations that govern emergency 

preparedness at nuclear power plants. Nuclear power reactor licensees are required to have 

NRC-approved "emergency response plans" for dealing with "radiological emergencies." The

requirements call for both onsite and offsite emergency response plans, with the offsite plans being 

those approved by FEMA and used by the State and local authorities. This document deals with the 

utilities' approved onsite plans and procedures for response to radiological emergencies at nuclear 

power plants, and the links they provide to the offsite plans.

Section 50.47 of Title 10 of the Code of Federal Regulations (10 CFR 50.47), entitled "Emergency 

Plans," states the requirement for such plans. Part (a)(1) of this regulation states that "no operating 

license will be issued unless a finding is made by NRC that there is reasonable assurance that 

adequate protective measures can and will be taken in the event of a radiological emergency." 

The major portion of 10 CFR 50.47 lists "standards" that emergency response plans must meet. 

The standards constitute a detailed list of items to be addressed in the plans. Of particular

importance to this project is the fourth standard, which addresses "emergency classification" and 

"action levels." These terms, however, are not defined in the regulation.

10 CFR 50.54, "Conditions of licenses," emphasizes that power reactor licensees must "follow, and 

maintain in effect, emergency plans which meet the standards in Part 50.47(b) and the

requirements in Appendix E to this part." The remainder of this part deals primarily with required 

implementation dates. 
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10 CFR 50.54(q) allows licensees to make changes to emergency plans without prior Commission 

approval only if: (a) the changes do not decrease the effectiveness of the plans and (b) the plans, 

as changed, continue to meet 10 CFR 50.47(b) standards and 10 CFR 50 Appendix E

requirements. The licensee must keep a record of any such changes. Proposed changes that 

decrease the effectiveness of the approved emergency plans may not be implemented without 

application to and approval by the Commission.

10 CFR 50.72 deals with "Immediate notification requirements for operating nuclear power

reactors." The "immediate" notification section actually includes three types of reports: (1)

immediately after notification of State or local agencies (for emergency classification events); (2) 

one-hour reports; and, (3) four-hour reports.

Although 10 CFR 50.72 contains significant detail, it does not define either "Emergency Class" or 

"Emergency Action Level." But one-hour and four-hour reports are listed as "non-emergency 

events," namely, those which are "not reported as a declaration of an Emergency Class." Certain 

10 CFR 50.72 events can also meet the Notification of Unusual Event emergency classification if they 

are precursors of more serious events. These situations also warrant anticipatory notification of 

state and local officials. (See Section 3.7, "Emergency Class Descriptions".)

By footnote, the reader is directed from 10 CFR 50.72 to 10 CFR 50 Appendix E, for information 

concerning "Emergency Classes."

10 CFR 50.73 describes the "Licensee event report system," which requires submittal of follow-up 

written reports within thirty days of required notification of NRC.

10 CFR 50 Appendix E, Section B, "Assessment Actions," mandates that emergency plans must 

contain "emergency action levels." EALs are to be described for: (1) determining the need for 

notification and participation of various agencies, and (2) determining when and what type of 

protective measures should be considered. Appendix E continues by stating that the EALs are to be 

based on: (1) in-plant conditions; (2) in-plant instrumentation; (3) onsite monitoring; and (4) offsite 

monitoring.

10 CFR 50 Appendix E, Section C, "Activation of Emergency Organization," also addresses 

"emergency classes" and "emergency action levels." This section states that EALs are to be based 

on: (1) onsite radiation monitoring information; (2) offsite radiation monitoring information; and, (3) 

readings from a number of plant sensors that indicate a potential emergency, such as containment 

pressure and the response of the Emergency Core Cooling System. This section also states that 

"emergency classes" shall include: (1) Notification of Unusual Events (NOUEs), (2) Alert, (3) Site 

Area Emergency, and (4) General Emergency.

These regulations are supplemented by various regulatory guidance documents. A significant

document that has dealt specifically with EALs is 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. 

Recognition Category D (Permanently Defueled Station) is based on the assumption that the spent 

fuel was generated by an operating nuclear power station under a 10 CFR 50 license that has 

ceased operations and intends to store the spent fuel for some period of time. The emergency 

classifications for Recognition Category D are those provided by NUREG 0654/FEMA Rep.1. The 

Unusual Event classifications are provided as an increased awareness for abnormal conditions. 

The Alert classifications are specific to the actual or potential effects on the spent fuel in storage. 
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In order for Permanently Defueled Stations to relax their existing emergency plan requirements 

these stations must verify that credible events cannot result in significant radiological releases 

beyond the site boundary. It is expected that this verification will confirm that the source term and 

motive force available in the permanently defueled condition is insufficient to warrant classifications 

of Site Area Emergency or General Emergency levels. Analyses for the credible design basis 

accidents are provided in the SAR.

Recognition Category E (Events Related to ISFSI) is applicable to licensees using their 10 CFR 50 

emergency plan to fulfill the requirements of 10 CFR 72.32. Recognition Category E is not

applicable to stand alone ISFSIs, Monitored Retrievable Storage Facilities (MRS), or ISFSIs that 

may process and/or repackage spent fuel. The emergency classifications for Recognition Category 

E are those provided by NUREG 0654/FEMA Rep.1 in accordance with 10 CFR 50.47. The

classification of an ISFSI event under provisions of a 10 CFR 50.47 emergency plan should be 

consistent with the definitions of the emergency classes as used by that plan. A site-specific 

analysis would make this determination, but in most cases it is expected that classification of an 

NOUE would be appropriate. It is expected that the initiating conditions germane to a 10 CFR 72.32 

emergency plan (described in NUREG-1567) are subsumed within 10 CFR 50.47 emergency plan’s 

classification scheme.

3.2 Definitions Used in Developing EAL Methodology

Based on the above review of regulations, review of common utility usage of terms, discussions 

among Task Force members, and existing published information, the following definitions apply to 

the generic EAL methodology:

EMERGENCY CLASS: One of a minimum set of names or titles, established by the 

Nuclear Regulatory Commission (NRC), for grouping off-normal nuclear power plant

conditions according to (1) their relative radiological seriousness, and (2) the time-sensitive 

onsite and off-site radiological emergency preparedness actions necessary to respond to 

such conditions. The existing radiological emergency classes, in ascending order of

seriousness, are called:

• Notification of Unusual Event

• Alert

• Site Area Emergency

• General Emergency

INITIATING CONDITION (IC): One of a predetermined subset of nuclear power plant 

conditions where either the potential exists for a radiological emergency, or such an

emergency has occurred.

Discussion:

In NUREG-0654, the NRC introduced, but does not define, the term "initiating condition." 

Since the term is commonly used in nuclear power plant emergency planning, the definition 

above has been developed and combines both regulatory intent and the greatest degree of 

common usage among utilities.
Revision 01/2003 3.4

Defined in this manner, an IC is an emergency condition which sets it apart from the broad 

class of conditions that may or may not have the potential to escalate into a radiological 

emergency. It can be a continuous, measurable function that is outside technical

specifications, such as elevated RCS temperature or falling reactor coolant level (a 

symptom). It also encompasses occurrences such as FIRE (an event) or reactor coolant 

pipe failure (an event or a barrier breach). 

EMERGENCY ACTION LEVEL (EAL): A pre-determined, site-specific, observable 

threshold for a plant Initiating Condition that places the plant in a given emergency class. An 

EAL can be: an instrument reading; an equipment status indicator; a measurable parameter 

(onsite or offsite); a discrete, observable event; results of analyses; entry into specific 

emergency operating procedures; or another phenomenon which, if it occurs, indicates 

entry into a particular emergency class.

Discussion:

The term "emergency action level" has been defined by example in the regulations, as noted 

in the above discussion concerning regulatory background. The term had not, however, 

been defined operationally in a manner to address all contingencies.

There are times when an EAL will be a threshold point on a measurable continuous function, 

such as a primary system coolant leak that has exceeded technical specifications for a 

specific plant.

At other times, the EAL and the IC will coincide, both identified by a discrete event that 

places the plant in a particular emergency class. For example, "Train Derailment Onsite" is 

an example of an "NOUE" IC in NUREG-0654 that also can be an event-based EAL.

3.3 Differences In Perspective

The purpose of this effort is to define a methodology for EAL development that will better assure a 

consistent emergency classification commensurate with the level of risk. The approach must be 

easily understood and applied by the individuals responsible for onsite and offsite emergency 

preparedness and response. In order to achieve consistent application, this recommended

methodology must be accepted at all levels of application (e.g., licensed operators, health physics 

personnel, facility managers, offsite emergency agencies, NRC and FEMA response organizations, 

etc.).

Commercial nuclear facilities are faced with a range of public service and public acceptance 

pressures. It is of utmost importance that emergency regulations be based on as accurate an 

assessment of the risk as possible. There are evident risks to health and safety in understating the 

potential hazard from an event. However, there are both risks and costs to alerting the public to an 

emergency that exceeds the true threat. This is true at all levels, but particularly if evacuation is 

recommended.

3.4 Recognition Categories

ICs and EALs can be grouped in one of several schemes. This generic classification scheme 

incorporates symptom-based, event-based, and barrier-based ICs and EALs.

The symptom-based category for ICs and EALs refers to those indicators that are measurable over 

some continuous spectrum, such as core temperature, coolant levels, containment pressure, etc. 

When one or more of these indicators begin to show off-normal readings, reactor operators are 

trained to identify the probable causes and potential consequences of these "symptoms" and take 
Revision 01/2003 3.5

corrective action. The level of seriousness indicated by these symptoms depends on the degree to 

which they have exceeded technical specifications, the other symptoms or events that are

occurring contemporaneously, and the capability of the licensed operators to gain control and bring 

the indicator back to safe levels.

Event-based EALs and ICs refer to occurrences with potential safety significance, such as the 

failure of a high-pressure safety injection pump, a safety valve failure, or a loss of electric power to 

some part of the plant. The range of seriousness of these "events" is dependent on the location, 

number of contemporaneous events, remaining plant safety margin, etc.

Barrier-based EALs and ICs refer to the level of challenge to principal barriers used to assure 

containment of radioactive materials contained within a nuclear power plant. For radioactive

materials that are contained within the reactor core, these barriers are: fuel cladding, reactor 

coolant system pressure boundary, and containment. The level of challenge to these barriers 

encompasses the extent of damage (loss or potential loss) and the number of barriers concurrently 

under challenge. In reality, barrier-based EALs are a subset of symptom-based EALs that deal with 

symptoms indicating fission product barrier challenges. These barrier-based EALs are primarily 

derived from Emergency Operating Procedure (EOP) Critical Safety Function (CSF) Status Tree 

Monitoring (or their equivalent). Challenge to one or more barriers generally is initially identified 

through instrument readings and periodic sampling. Under present barrier-based EALs,

deterioration of the reactor coolant system pressure boundary or the fuel clad barrier usually 

indicates an "Alert" condition, two barriers under challenge a Site Area Emergency, and loss of two 

barriers with the third barrier under challenge is a General Emergency. The fission product barrier 

matrix described in Section 5-F is a hybrid approach that recognizes that some events may 

represent a challenge to more than one barrier, and that the containment barrier is weighted less 

than the reactor coolant system pressure boundary and the fuel clad barriers. 

Symptom-based ICs and EALs are most easily identified when the plant is in a normal startup, 

operating or hot shutdown mode of operation, with all of the barriers in place and the plant's 

instrumentation and emergency safeguards features fully operational as required by technical

specifications. It is under these circumstances that the operations staff has the most direct

information of the plant's systems, displayed in the main control room. As the plant moves through 

the decay heat removal process toward cold shutdown and refueling, barriers to fission products 

are reduced (i.e., reactor coolant system pressure boundary may be open) and fewer of the safety 

systems required for power operation are required to be fully operational. Under these plant

operating modes, the identification of an IC in the plant's operating and safety systems becomes 

more event-based, as the instrumentation to detect symptoms of a developing problem may not be 

fully effective; and engineered safeguards systems, such as the Emergency Core Cooling System 

(ECCS), are partially disabled as permitted by the plant's Technical Specifications.

Barrier-based ICs and EALs also are heavily dependent on the ability to monitor instruments that 

indicate the condition of plant operating and safety systems. Fuel cladding integrity and reactor 

coolant levels can be monitored through several indicators when the plant is in a normal operating 

mode, but this capability is much more limited when the plant is in a refueling mode, when many of 

these indicators are disconnected or off-scale. The need for this instrumentation is lessened, 

however, and alternate instrumentation is placed in service when the plant is shut down.

It is important to note that in some operating modes there may not be definitive and unambiguous 

indicators of containment integrity available to control room personnel. For this reason, barrierbased EALs should not place undue reliance on assessments of containment integrity in all

operating modes. Generally, Technical Specifications relax maintaining containment integrity

requirements in modes 5 and 6 in order to provide flexibility in performance of specific tasks during 

shutdown conditions. Containment pressure and temperature indications may not increase if there 
Revision 01/2003 3.6

is a pre-existing breach of containment integrity. At most plants, a large portion of the containment's 

exterior cannot be monitored for leakage by radiation monitors.

Several categories of emergencies have no instrumentation to indicate a developing problem, or the 

event may be identified before any other indications are recognized. A reactor coolant pipe could 

break; FIRE alarms could sound; radioactive materials could be released; and any number of other 

events can occur that would place the plant in an emergency condition with little warning. For 

emergencies related to the reactor system and safety systems, the ICs shift to an event based 

scheme as the plant mode moves toward cold shutdown and refueling modes. For non-radiological 

events, such as FIRE, external floods, wind loads, etc., as described in NUREG-0654 Appendix 1, 

event-based ICs are the norm.

In many cases, a combination of symptom-, event- and barrier-based ICs will be present as an 

emergency develops. In a loss of coolant accident (LOCA), for example:

• Coolant level is dropping; (symptom)

• There is a leak of some magnitude in the system (pipe break, safety valve stuck open) that 

exceeds plant capabilities to make up the loss; (barrier breach or event)

• Core (coolant) temperature is rising; (symptom) and

• At some level, fuel failure begins with indicators such as high off-gas, high coolant activity 

samples, etc. (barrier breach or symptom)

3.5 Design Differences

Although the same basic concerns with barrier integrity and the major safety problems of nuclear 

power plants are similar across plant types, design differences will have a substantial effect on 

EALs. The major differences are found between a BWR and a PWR. In these cases, EAL

guidelines unique to BWRs and PWRs must be specified. Even among PWRs, however, there are 

substantial differences in design and in types of containment used.

There is enough commonality among plants that many ICs will be the same or very similar. 

However, others will have to match plant features and safety system designs that are unique to the 

plant type or even to the specific plant. The basis for each EAL guideline should supply sufficient 

information as to what is required for a site-specific EAL.

3.6 Required Characteristics

Eight characteristics that should be incorporated into model EALs are identified below:

(1) Consistency (i.e., the EALs would lead to similar decisions under similar circumstances at 

different plants);

(2) Human engineering and user friendliness;

(3) Potential for classification upgrade only when there is an increasing threat to public health 

and safety;

(4) Ease of upgrading and downgrading;
Revision 01/2003 3.7

(5) Thoroughness in addressing, and disposing of, the issues of completeness and accuracy 

raised regarding NUREG-0654 Appendix 1;

(6) Technical completeness for each classification level;

(7) A logical progression in classification for multiple events; and

(8) Objective, observable values.

The EAL development procedure pays careful attention to these eight characteristics to assure that 

all are addressed in the proposed EAL methodology. The most pervasive and complex of the eight 

is the first—“consistency." The common denominator that is most appropriate for measuring

consistency among ICs and EALs is relative risk. The approach taken in the development of these 

EALs is based on risk assessment to set the boundaries of the emergency classes and assure that 

all EALs that trigger that emergency class are in the same range of relative risk. Precursor

conditions of more serious emergencies also represent a potential risk to the public and must be 

appropriately classified.

3.7 Emergency Class Descriptions

There are three considerations related to emergency classes. These are:

(1) The potential impact on radiological safety, either as now known or as can be reasonably 

projected;

(2) How far the plant is beyond its predefined design, safety, and operating envelopes; and

(3) Whether or not conditions that threaten health are expected to be confined to within the site 

boundary.

The ICs deal explicitly with radiological safety impact by escalating from levels corresponding to 

releases within regulatory limits to releases beyond EPA Protective Action Guideline (PAG) plume 

exposure levels. In addition, the "Discussion" sections below include offsite dose consequence 

considerations which were not included in NUREG-0654 Appendix 1.

NOTIFICATION OF UNUSUAL EVENT: Events are in process or have occurred which 

indicate a potential degradation of the level of safety of the plant. No releases of radioactive 

material requiring offsite response or monitoring are expected unless further degradation of 

safety systems occurs.

Discussion:

Potential degradation of the level of safety of the plant is indicated primarily by exceeding 

plant technical specification Limiting Condition of Operation (LCO) allowable action

statement time for achieving required mode change. Precursors of more serious events 

should also be included because precursors do represent a potential degradation in the level 

of safety of the plant. Minor releases of radioactive materials are included. In this emergency 

class, however, releases do not require monitoring or offsite response (e.g., dose

consequences of less than 10 millirem).

ALERT: Events are in process or have occurred which involve an actual or potential

substantial degradation of the level of safety of the plant. Any releases are expected to be 

limited to small fractions of the EPA Protective Action Guideline exposure levels.
Revision 01/2003 3.8

Discussion:

Rather than discussing the distinguishing features of "potential degradation" and "potential 

substantial degradation," a comparative approach would be to determine whether increased 

monitoring of plant functions is warranted at the Alert level as a result of safety system 

degradation. This addresses the operations staff's need for help, independent of whether an 

actual decrease in plant safety is determined. This increased monitoring can then be used 

to better determine the actual plant safety state, whether escalation to a higher emergency 

class is warranted, or whether de-escalation or termination of the emergency class 

declaration is warranted. Dose consequences from these events are small fractions of the 

EPA PAG plume exposure levels, i.e., about 10 millirem to 100 millirem TEDE.

SITE AREA EMERGENCY: Events are in process or have occurred which involve actual or 

likely major failures of plant functions needed for 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.

Discussion:

The discriminator (threshold) between Site Area Emergency and General Emergency is 

whether or not the EPA PAG plume exposure levels are expected to be exceeded outside 

the site boundary. This threshold, in addition to dynamic dose assessment considerations 

discussed in the EAL guidelines, clearly addresses NRC and offsite emergency response 

agency concerns as to timely declaration of a General Emergency.

GENERAL EMERGENCY: Events are in process or have occurred which involve actual or 

imminent substantial core degradation or melting with potential for loss of containment 

integrity. Releases can be reasonably expected to exceed EPA Protective Action Guideline 

exposure levels offsite for more than the immediate site area.

Discussion:

The bottom line for the General Emergency is whether evacuation or sheltering of the 

general public is indicated based on EPA PAGs, and therefore should be interpreted to 

include radionuclide release regardless of cause. In addition, it should address concerns as 

to uncertainties in systems or structures (e.g. containment) response, and also events such 

as waste gas tank releases and severe spent fuel pool events postulated to occur at high 

population density sites. To better assure timely notification, EALs in this category must 

primarily be expressed in terms of plant function status, with secondary reliance on dose 

projection. In terms of fission product barriers, loss of two barriers with loss or potential loss 

of the third barrier constitutes a General Emergency.

3.8 Emergency Class Thresholds

The most common bases for establishing these boundaries are the technical specifications and 

setpoints for each plant that have been developed in the design basis calculations and the Final 

Safety Analysis Report (FSAR).

For those conditions that are easily measurable and instrumented, the boundary is likely to be the 

EAL (observable by plant staff, instrument reading, alarm setpoint, etc.) that indicates entry into a 

particular emergency class. For example, the main steam line radiation monitor may detect high 

radiation that triggers an alarm. That radiation level also may be the setpoint that closes the main 
Revision 01/2003 3.9

steam isolation valves (MSIV) and initiates the reactor scram. This same radiation level threshold, 

depending on plant-specific parameters, also may be the appropriate EAL for a direct entry into an 

emergency class.

In addition to the continuously measurable indicators, such as coolant temperature, coolant levels, 

leak rates, containment pressure, etc., the FSAR provides indications of the consequences

associated with design basis events. Examples would include steam pipe breaks, MSIV

malfunctions, and other anticipated events that, upon occurrence, place the plant immediately into 

an emergency class. 

Another approach for defining these boundaries is the use of a plant-specific probabilistic safety 

assessment (PSA - also known as probabilistic risk assessment, PRA). PSAs have been

completed for several individual plants, but this is by no means comprehensive. There are,

however, PSAs that have been completed for representative plant types such as is done in

NUREG-1150, "Severe Accident Risks: An Assessment for Five Nuclear Power Plants," as well as 

several other utility-sponsored PSAs. Existing PSAs can be used as a good first approximation of 

the relevant ICs and risk associated with emergency conditions for existing plants. Generic insights 

from PSAs and related severe accident assessments which apply to EALs and emergency class 

determinations are:

1. Core damage frequency at many BWRs is dominated by sequences involving prolonged loss of 

all AC power. In addition, prolonged loss of all AC power events are extremely important at 

PWRs. This would indicate that should this occur, and AC power is not restored within 15 

minutes, entry into the emergency class at no lower than a Site Area Emergency, when the 

plant was initially at power, would be appropriate. This implies that precursors to loss of all AC 

power events should appropriately be included in the EAL structure.

2. For severe core damage events, uncertainties exist in phenomena important to accident

progressions leading to containment failure. Because of these uncertainties, predicting

containment integrity may be difficult in these conditions. This is why maintaining containment 

integrity alone following sequences leading to severe core damage may be an insufficient basis 

for not escalating to a General Emergency.

3. A review of four full-scope PRAs (3 PWR, 1 BWR) showed that leading contributors to latent 

fatalities were containment bypass, large LOCA with early containment failure, station blackout 

greater than 6 hours (e.g., LOCA consequences of Station Blackout), and reactor coolant pump 

seal failure. This indicates that generic EAL methodology must be sufficiently rigorous to cover 

these sequences in a timely fashion.

Another critical element of the analysis to arrive at these threshold (boundary) conditions is the time 

that the plant might stay in that condition before moving to a higher emergency class. In particular, 

station blackout coping analyses performed in response to 10 CFR 50.63 and Regulatory Guide 

1.155, "Station Blackout," may be used to determine whether a specific plant enters a Site Area 

Emergency or a General Emergency directly, and when escalation to General Emergency is 

indicated. The time dimension is critical to the EAL since the purpose of the emergency class for 

state and local officials is to notify them of the level of mobilization that may be necessary to handle 

the emergency. This is particularly true when a "Site Area Emergency" or "General Emergency" is 

imminent. Establishing EALs for such conditions must take estimated evacuation time into

consideration to minimize the potential for the plume to pass while evacuation is underway.

Regardless of whether or not containment integrity is challenged, it is possible for significant 

radioactive inventory within containment to result in EPA PAG plume exposure levels being

exceeded even assuming containment is within technical specification allowable leakage rates. 
Revision 01/2003 3.10

With or without containment challenge, however, a major release of radioactivity requiring offsite 

protection actions from core damage is not possible unless a major failure of fuel cladding allows 

radioactive material to be released from the core into the reactor coolant. NUREG-1228, "Source 

Estimations During Incident Response to Severe Nuclear Power Plant Accidents," indicates that 

such conditions do not exist when the amount of clad damage is less than 20%.

3.9 Emergency Action Levels

With the emergency classes defined, the thresholds that must be met for each EAL to be placed 

under the emergency class can be determined. There are two basic approaches to determining 

these EALs. EALs and emergency class boundaries coincide for those continuously measurable, 

instrumented ICs, such as radioactivity, core temperature, coolant levels, etc. For these ICs, the 

EAL will be the threshold reading that most closely corresponds to the emergency class description 

using the best available information.

For discrete (discontinuous) events, the approach will have to be somewhat different. Typically, in 

this category are internal and external hazards such as FIRE or earthquake. The purpose for 

including hazards in EALs is to assure that station personnel and offsite emergency response 

organizations are prepared to deal with consequential damage these hazards may cause. If, 

indeed, hazards have caused damage to safety functions or fission product barriers, this should be 

confirmed by symptoms or by observation of such failures. Therefore, it may be appropriate to enter 

an Alert status for events approaching or exceeding design basis limits such as Operating Basis 

Earthquake, design basis wind loads, FIRE within VITAL AREAs, etc. This would give the operating 

staff additional support and improved ability to determine the extent of plant damage. if damage to 

barriers or challenges to Critical Safety Functions (CSFs) have occurred or are identified, then the 

additional support can be used to escalate or terminate the Emergency Class based on what has 

been found. Of course, security events must reflect potential for increasing security threat levels.

Plant emergency operating procedures (EOPs) are designed to maintain and/or restore a set of 

CSFs which are listed in the order of priority for restoration efforts during accident conditions. 

While the actual nomenclature of the CSFs may vary among plants, generally the PWR CSF set 

includes:

ü Subcriticality

ü Core cooling

ü Heat sink

ü Pressure-temperature-stress (RCS integrity)

ü Containment

ü RCS inventory

There are diverse and redundant plant systems to support each CSF. By monitoring the CSFs 

instead of the individual system component status, the impact of multiple events is inherently 

addressed, e.g., the number of operable components available to maintain the critical safety

function.

The EOPs contain detailed instructions regarding the monitoring of these functions and provides a 

scheme for classifying the significance of the challenge to the functions. In providing EALs based 

on these schemes, the emergency classification can flow from the EOP assessment rather than 

being based on a separate EAL assessment. This is desirable as it reduces ambiguity and reduces 

the time necessary to classify the event.

As an example, consider that the Westinghouse Owner's Group (WOG) Emergency Response 

Guidelines (ERGs) classify challenges as YELLOW, ORANGE, and RED paths. If the core exit 
Revision 01/2003 3.11

thermocouples exceed 1200 degrees F or 700 degrees F with low reactor vessel water level, a 

RED path condition exists. The ERG considers a RED path as "... an extreme challenge to a plant 

function necessary for the protection of the public ..." This is almost identical to the present NRC 

NUREG-0654 description of a site area emergency "... actual or likely failures of plant functions 

needed for the protection of the public ..." It reasonably follows that if any CSF enters a RED path, a 

site area emergency exists. A general emergency could be considered to exist if core cooling CSF 

is in a RED path and the EOP function restoration procedures have not been successful in 

restoring core cooling.

Although the majority of the EALs provide very specific thresholds, the Emergency Director must 

remain alert to events or conditions that lead to the conclusion that exceeding the EAL threshold is 

imminent. If, in the judgment of the Emergency Director, an imminent situation is at hand, the 

classification should be made as if the thresholds has been exceeded. While this is particularly 

prudent at the higher emergency classes (as the early classification may provide for more effective 

implementation of protective measures), it is nonetheless applicable to all emergency classes.

3.10 Treatment Of Multiple Events And Emergency Class Upgrading 

The above discussion deals primarily with simpler emergencies and events that may not escalate 

rapidly. However, usable EAL guidance must also consider rapidly evolving and complex events. 

Hence, emergency class upgrading and consideration of multiple events must be addressed.

There are three approaches presently in use for covering multiple events and emergency class 

upgrading. These approaches are:

(U1) Multiple contemporaneous events are counted and are the basis for escalating to a higher 

emergency class. For example, two or more contemporaneous Alerts escalate to a Site 

Area Emergency.

(U2) The emergency class is based on the highest EAL reached. For example, two Alerts remain 

in the Alert category. Or, an Alert and a Site Area Emergency is a Site Area Emergency.

(U3) Emergency Director judgment. Although all emergency classifications require judgment, 

some utilities rely on Emergency Director judgment with little or no additional explicit

guidance.

An additional approach for plants with PRAs is to make use of event tree analysis to define 

combinations of events which lead to equivalent risks. Such event sequences should have an equal 

emergency classification assigned. However, the chief drawback to this approach as well as (U1) 

above, is that multiple events may be masked when they actually occur. Further, for plants using 

symptom-based (and barrier-based) emergency procedures, direct perception of multiple events is 

unnecessary. 

Emergency class upgrading for multi-unit stations with shared safety-related systems and functions 

must also consider the effects of a loss of a common system on more than one unit (e.g. potential 

for radioactive release from more than one core at the same site). For example, many two-unit 

stations have their control panels for both units in close proximity within the same room. Thus, 

control room evacuation most likely would affect both units. There are a number of other systems 

and functions which may be shared at a given multi-unit station. This must be considered in the 

emergency class declaration and in the development of appropriate site-specific ICs and EALs 

based on the generic EAL guidance.
Revision 01/2003 3.12

Although the majority of the EALs provide very specific thresholds, the Emergency Director must 

remain alert to events or conditions that lead to the conclusion that exceeding the EAL threshold is 

imminent. If, in the judgment of the Emergency Director, an imminent situation is at hand, the 

classification should be made as if the thresholds has been exceeded. While this is particularly 

prudent at the higher emergency classes (as the early classification may provide for more effective 

implementation of protective measures), it is nonetheless applicable to all emergency classes.

RECOMMENDATION:

The best approach is (U2) above with appropriate consideration for Emergency

Director judgment EALs. Properly structured EALs on a fission product barrier basis 

and which include equivalent risk, will appropriately escalate multiple events to a 

higher emergency class. For example, common cause failures such as loss of

ultimate heat sink or loss of all AC power, will result in multiple contemporaneous 

symptoms indicating safety system functional failures and increasing challenge to 

fission product barriers. It is the existence of these symptoms (barrier challenges) 

that escalate the emergency class, whether there are one or multiple causes.

3.11 Emergency Class Downgrading

Another important aspect of usable EAL guidance is the consideration of what to do when the risk 

posed by an emergency is clearly decreasing. There are several approaches presently in use for 

emergency class downgrading. These approaches are:

(D1) Terminate the emergency class declaration.

(D2) Recovery from emergency class. 

(D3) Combination of downgrading approaches. Many utilities reviewed include the option to

downgrade to a lower emergency class. This is consistent with actions called for in

NUREG-0654 Appendix 1. However, these utilities state that their experience more closely 

resembles (D1) and (D2) above as practical choices.

Another approach possible with risk-based EALs is a relatively simple approach for upgrading to a 

higher emergency class when the risk increases and downgrading when risk decreases. The 

boundaries for emergency categories are defined in terms of risk in this approach, and discrete 

events fall into these categories based on risk. This means that within each emergency class, there 

is uniformity to the relative levels of risk to human health and safety from radiological accidents. 

However, this option may not be practical when applied to actual emergencies, especially those 

involving General Emergencies.

RECOMMENDATION:

A combination approach involving recovery from General Emergencies and some 

Site Area Emergencies and termination from NOUEs, Alerts, and certain Site Area 

Emergencies causing no long-term plant damage appears to be the best choice. 

Downgrading to lower emergency classes adds notifications but may have merit 

under certain circumstances.
Revision 01/2003 3.13

3.12 Classifying Transient Events

For some events, the condition may be corrected before a declaration has been made. For

example, an emergency classification is warranted when automatic and manual actions taken 

within the control room do not result in a required reactor scram. However, it is likely that actions 

taken outside of the control room will be successful, probably before the Emergency Director 

classifies the event. The key consideration in this situation is to determine whether or not further 

plant damage occurred while the corrective actions were being taken. In some situations, this can 

be readily determined, in other situations, further analyses (e.g., coolant radiochemistry sampling, 

may be necessary). There are several approaches presently in use for handling transient events. 

These approaches are:

(T1) Classify the event as indicated and terminate the emergency once assessment shows that 

there were no consequences from the event and other termination criteria are met.

(T2) No emergency declaration is made, but the event is reported and notifications are made.

RECOMMENDATION

Option (T1) is believed to be appropriate for events at higher emergency

classifications. Option (T2) may be appropriate for events that might have been 

classified as NOUEs, but might not be sufficient for some events (e.g., ATWS). It is 

recommended that the program incorporate aspects of both options with examples 

of when each would be appropriate. Many of the generic event-based IC’s and EAL’s 

have discriminators based on time or magnitude. Generally, if the discriminator is 

exceeded, the event should be classified. In implementing the generic guidance 

into site-specific programs, care should be taken to ensure that the ICs and EALs 

minimize the need for these ad hoc decisions on transient events. 

There may be cases in which a plant condition that exceeded an EAL threshold was not recognized 

at the time of occurrence, but is identified well after the condition has occurred (e.g., as a result of 

routine log or record review) and the condition no longer exists. In these cases, an emergency 

should not be declared.

Reporting requirements of 10 CFR 50.72 are applicable and the guidance of NUREG-1022, Rev. 1, 

Section 3 should be applied.

3.13 Interface Between Classification and Activation of Emergency 

Facilities

Existing regulations call for the activation of various emergency facilities at different levels of

emergency classification. The intent of activating these facilities is to provide needed support to the 

on-shift complement. A question often arises, “If I utilize the TSC as a precautionary measure do I 

have to declare an Alert emergency?” There are two possible situations:

• The Emergency Director is faced with an event or series of events which individually may not 

constitute an Alert emergency, but in combination, is causing the Emergency Director with 

concern over his ability to contend with the situation using his on-shift resources. This should be 

clearly recognized as a case in which the Emergency Director judgment ICs apply, and the 

emergency classification is probably warranted.
Revision 01/2003 3.14

• The site has received warning of severe weather. Site management deems it prudent to utilize 

the onsite emergency facilities to ensure the availability of personnel should the weather cause 

plant damage while personnel travel is hindered. This situation wouldn’t warrant an Alert

classification unless the severe weather warning was such that damage comparable to an Alert 

IC was expected. 

RECOMMENDATION

The key consideration is not the fact that the facilities were utilized, but rather, the 

reason for that use. Facilities may be used for events that may not warrant

classification of an emergency.

3.14 Cold Shutdown/Refueling IC/EALs

Generic Letter 88–17, Loss of Decay Heat Removal, SECY–91–283, Evaluation of Shutdown and 

Low Power Risk Issues, SECY–93–190, Regulatory Approach to Shutdown and Low–power 

Operation, 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, all address nuclear power plant safety issues that are applicable to periods 

when the plant is shutdown. These evaluations identify a number of variables which significantly 

affect the probability and consequences of losing decay heat removal capability during shutdown 

periods. In addition, NUREG–1449 discusses that the need to respond appropriately, including 

emergency classification and notification, still exists during cold–shutdown and refueling conditions. 

Both SECY–93–190 and NUREG–1449 have been reviewed and issues concerning shutdown 

effects on declaring emergencies have been addressed. 

Given the variability of plant configurations (e.g., systems out–of–service for maintenance,

containment open, reduced AC power redundancy, time since shutdown) during these periods, the 

consequences of any given initiating event can vary greatly. For example, a loss of decay heat 

removal capability that occurs at the end of an extended outage has less significance than a similar 

loss occurring during the first week after shutdown. Compounding these events is the likelihood 

that instrumentation necessary for assessment may also be inoperable. The NEI cold shutdown 

and refueling EALs are based on performance capability to the extent possible with consideration 

given to RCS integrity, containment closure, and fuel clad integrity for the applicable modes.

The initiating conditions and example emergency actions levels associated directly with Cold

Shutdown or Refueling safety function are presented in Recognition Category C, Cold

Shutdown/Refueling. The example EALs for both PWR and BWR are consistent with the public 

risk associated with the other events represented in the Fission Product Barrier Matrix and in other 

sections of this document.

Boiling water reactors and pressurized water reactors differ significantly with regard to plant

response to events that occur during shutdowns. There is generally a larger water inventory in a 

BWR than in a PWR. Containment isolation capability is generally better in PWRs than in earlier 

design BWRs. Where differences exist, separate BWR/PWR EALs have been prepared to reflect 

the differences in plant vulnerabilities or mitigation features.

The guidance which addresses cold shutdown/refueling IC/EALs in NEI 99-01 is intended to 

address both NUMARC/NESP-007 and NUREG-0654 users. For NUREG-0654 users, the scope of 

the cold shutdown/refueling initiative is limited to the “new” IC/EALs (CU2, CU4, CU8, CA1, CA2, 

and CG1), CA4 (compare with NUREG-0654 Example Alert 10), and CS1 and CS2 (partially related 

to NUREG-0654 Example Site Area Emergency 10).
Revision 01/2003 3.15

3.15 Permanently Defueled Station IC/EALs

A Permanently Defueled Station is basically a spent fuel storage facility. The spent fuel is stored in 

a pool of water that serves as both the cooling medium for decay heat and shielding from direct 

radiation. The primary functions of this pool configuration become the emphasis of emergency 

classification methodology.

When in the permanently defueled condition, the licensee receives approval for specific emergency 

planning requirements negotiated with the State and local governmental agencies and the NRC. 

The source term and relative risks associated with pool storage are the basis for maintaining only 

an onsite emergency plan. Calculations are provided in the licensing process that quantify

radioactive releases associated with plausible accidents.

The guidance which addresses permanently defueled station IC/EALs in NEI 99-01 is intended to 

address both NUMARC/NESP-007 and NUREG-0654 users. 

3.16 ISFSI IC/EALs

An Independent spent fuel storage installation (ISFSI) is a complex that is designed and constructed 

for the interim storage of spent nuclear fuel and other radioactive materials associated with spent 

fuel storage. The Final Rule governing Emergency Planning Licensing Requirements for

Independent Spent Fuel Storage Facilities (Federal Register Volume 60, Number 120 June 22, 

1995, Pages 32430-32442) indicated that a significant amount of the radioactive material contained 

within a cask must escape its packaging and enter the biosphere for there to be a significant 

environmental impact resulting from an accident involving the dry storage of spent nuclear fuel. 

Formal offsite planning is not required because the postulated worst-case accident involving an 

ISFSI has insignificant consequences to the public health and safety.

The guidance which addresses ISFSI IC/EALs in NEI 99-01 is intended to address both

NUMARC/NESP-007 and NUREG-0654 users. Licensees may choose to present site-specific 

ISFSI IC/EALs separate from other ICs/EALs as presented herein, or integrate them into

Recognition Category A, H, and S IC/EALs.

3.17 Operating Mode Applicability

Emergency action levels have typically been written without regard to the operating mode to which 

they apply. While the applicable operating modes are obvious for some initiating conditions (e.g., 

failure of the reactor protection system), the situation is not as clear for others.

The plant operating mode that existed at the time that the event occurred, prior to any protective 

system or operator action initiated in response to the condition, is compared to the mode

applicability of the EALs. If an event occurs, and a lower or higher plant operating mode is reached 

before the emergency classification can be made, the declaration shall be based on the mode that 

existed at the time the event occurred. 

Note that in Revision 4 the system malfunction matrices have been completely separated such that 

the system ICs that apply to the Hot Shutdown mode and above are located in Category S and the 

system ICs that apply to the Cold Shutdown mode and below are located in Category C. 

For events that occur in Cold Shutdown or Refueling, escalation is via EALs that have Cold 

Shutdown or Refueling for mode applicability, even if Hot Shutdown (or a higher mode) is entered 
Revision 01/2003 3.16

during any subsequent heat–up. In particular, the Fission Product Barrier Matrix EALs are

applicable only to events that initiate in Hot Shutdown or higher.

3.17.1 Mode Applicability Matrix

Recognition Category C completely replaces Recognition Category S when in Cold Shutdown and 

Refueling modes. It should be noted that Recognition Category A and H IC/EALs still apply when in 

Cold Shutdown and Refueling modes. Recognition Category F is not applicable to Cold Shutdown 

and Refueling modes.

MODE APPLICABILITY MATRIX

Recognition Category

Mode A C D E F H S

Operating X X X X

Startup X X X X

Hot Standby X X X X

Hot Shutdown X X X X

Cold Shutdown X X X

Refueling X X X

Defueled X X X

None X X

The modes identified in the IC/EALs were based on the standard technical specifications for BWRs 

and Westinghouse PWRs. To aid in interpreting these modes for PWRs from other NSSSs and for 

plant with non–standard technical specifications, the modes are described below. 

3.17.2 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.

Defueled (None) All reactor fuel removed from reactor pressure vessel

(Full core off load during refueling or extended outage).
Revision 01/2003 3.17

3.17.3 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

Defueled (None) All reactor fuel removed from reactor pressure vessel.

(Full core off load during refueling or extended outage)
Revision 01/2003 4.1

4.0 HUMAN FACTORS CONSIDERATIONS

Some factors that should be considered in determining the method of presentation of EALs:

• Who is the audience (user) for this information? A senior utility executive would likely want 

information presented differently than a licensed operator. Offsite agencies and the NRC may 

have entirely different information needs.

• The conditions under which the information must be read, understood, and acted upon. Since 

the subject matter here is emergency actions, it is highly likely that the user of the EALs will be 

under high stress during the conditions where they are required to be used, particularly under 

conditions corresponding to Site Area Emergency and General Emergency.

• What is the user's perception as to the importance of the EALs compared to other actions and 

decisions that may be needed at the same time? To allow a licensed operator to discharge his 

responsibilities for dealing with the situation and also provide prompt notification to outside 

agencies, the emergency classification and notification process must be rapid and concise.

• Is the EAL consistent with the user's knowledge of what constitutes an emergency situation? 

• How much help does the user receive in deciding which EAL and emergency class is involved? 

An Emergency Director with a staffed TSC and EOF has many more resources immediately at 

his disposal than the licensed operator (typically, the Shift Supervisor) who has to make the 

initial decisions and take first actions.

Based on review of a number of plants' EALs and associated information, interviews with utility 

personnel, and a review of drill experience some recommendations follow.

4.1 Level Of Integration Of EALs With Plant Procedures

A rigorous integration of EALs and emergency class determinations into the plant procedure set, 

although having some benefits, is probably unnecessary. Such a rigorous integration could well 

make it more difficult to keep documentation up-to-date. However, keeping EALs totally separated 

from plant procedures and relying on licensed operator or other utility Emergency Director memory 

during infrequent, high stress periods is insufficient. 

RECOMMENDATION:

Visual cues in the plant procedures that it is appropriate to consult the EALs is a 

method currently used by several utilities. This method can be effective when it is 

tied to appropriate training. Notes in the appropriate plant procedures to consult 

the EALs can also be used. It should be noted that this discussion is not restricted 

to only the emergency procedures; alarm recognition procedures, abnormal

operating procedures, and normal operating procedures that apply to cold shutdown 

and refueling modes should also be included. In addition, EALs can be based on 

entry into particular procedures or existence of particular Critical Safety Function 

conditions.
Revision 01/2003 4.2

4.2 Method Of Presentation

A variety of presentation methods are presently in use. Methods range from directly copying 

NUREG-0654 Appendix 1 language, adding plant-specific indications to clarify NUREG-0654, use of 

procedure language including specific tag numbers for instrument readings and alarms, deliberate 

omission of instrument tag numbers, flow charts, critical safety function status trees, checklists, 

and combinations of the above.

What is clear, however, is that the licensed operator (typically the Shift Supervisor) is the first user 

of this information, has the least amount of help in interpreting the EALs, and also has other 

significant responsibilities to fulfill while dealing with the EALs. Emergency Directors outside the 

control room to whom responsibilities are turned over have other resources and advisors available 

to them that a licensed operator may not have when first faced with an emergency situation. In 

addition, as an emergency situation evolves, the operating staff and the health physics staff are the 

personnel who must first deal with information that is germane to changing the emergency

classification (up, down, or out of the emergency class).

RECOMMENDATION:

The method of presentation should be one with which the operations and health 

physics staff are comfortable. As is the case for emergency procedures, bases for 

steps should be in a separate (or separable) document suitable for training and for 

reference by emergency response personnel and offsite agencies. Each nuclear 

plant should already have presentation and human factors standards as part of its 

procedure writing guidance. EALs that are consistent with those procedure writing 

standards (in particular, emergency operating procedures which most closely

correspond to the conditions under which EALs must be used) should be the norm 

for each utility.

4.3 Symptom-based, Event-based, Or Barrier-based EALs

A review of the emergency class descriptions provided elsewhere in this document shows that 

NOUEs and Alerts deal primarily with sequences that are precursors to more serious emergencies 

or that may have taken a plant outside of its intended operating envelope, but currently pose no 

danger to the public. Observable indications in these classes can be events (e.g. natural

phenomena), symptoms (e.g., high temperature, low water level), or barrier-related (e.g., challenge 

to fission product barrier). As one escalates to Site Area Emergency and General Emergency, 

potential radiological impact to people (both onsite and offsite) increases. However, at this point the 

root cause event(s) leading to the emergency class escalation matter far less than the increased 

(potential for) radiological releases. Thus, EALs for these emergency classes should be primarily 

symptom- and barrier-based. It should be noted again, as stated in Section 3.4, that barrier 

monitoring is a subset of symptom monitoring, i.e., what readings (symptoms) indicate a challenge 

to a fission product barrier.

RECOMMENDATION:

A combination approach that ranges from primarily event-based for NOUEs to

primarily symptom- or barrier-based for General Emergencies is recommended. 

This is to better assure that timely recognition and notification occurs, that events 

occurring during refueling and cold shutdown are appropriately covered, and that 

multiple events can be effectively treated in the EALs.
Revision 01/2003 5.1

5.0 GENERIC EAL GUIDANCE

This section provides generic EAL guidance based on the information gathered and reviewed by the 

Task Force. Because of the wide variety of presentation methods used at different utilities, this 

document specifies guidance as to what each IC and EAL should address, and including sufficient 

basis information for each will best assure uniformity of approach. This approach is analogous to 

reactor vendors' owners groups developing generic emergency procedure guidelines which are 

converted by each utility into plant-specific emergency operating procedures. Each utility is 

reminded, however, to review the "Human Factors Considerations" section of this document as part 

of implementation of the attached Generic EAL Guidance.

5.1 Generic Arrangement

The information is presented by Recognition Categories:

ü A - Abnormal Rad Levels / Radiological Effluent

ü C - Cold Shutdown./ Refueling System Malfunction

ü D – Permanently Defueled Station Malfunction

ü E - Events Related to Independent Spent Fuel Storage Installations

ü F - Fission Product Barrier Degradation

ü H - Hazards and Other Conditions Affecting Plant Safety

ü S - System Malfunction

The Initiating Conditions for each of the above Recognition Categories A, C, D, E, H, and S are in 

the order of NOUE, Alert, Site Area Emergency, and General Emergency. For all Recognition 

Categories, an Initiating Condition matrix versus Emergency Class is first shown. For Recognition 

Category F, the barrier-based EALs are presented in Tables F-1 and F-2 for BWRs and PWRs 

respectively. For all other Recognition Categories separate BWR and PWR Initiating Condition 

matrices are not required. The purpose of the IC matrices is to provide the reader with an overview 

of how the ICs are logically related under each Emergency Class.

Each of the EAL guides in Recognition Categories A, C, D, E, H, and S is structured in the following 

way:

• Recognition Category- As described above.

• Emergency Class - NOUE, Alert, Site Area Emergency or General Emergency. 

• Initiating Condition – Symptom- or Event-Based, Generic Identification and Title.

• Operating Mode Applicability - refers to the operating mode (PWRs) or operating condition 

(BWRs) during which the IC/EAL is applicable - Power Operation (includes Startup Mode in 

PWRs), Hot Standby (includes Hot Standby / Startup Condition in BWRs), Hot Shutdown, Cold 

Shutdown, Refueling, Defueled, All, or None. These modes are defined in each licensee’s 

technical specifications. The mode classifications and terminology appropriate to the specific 
Revision 01/2003 5.2

facility should be used. See also Section 3.15. Note that Permanently Defueled and ISFSI 

IC/EALs have no mode applicability.

If an IC or EAL includes an explicit reference to a technical specification, and the technical 

specification is not applicable because of operating mode, then that particular IC or EAL is also 

not applicable.

• Example Emergency Action Level(s) – these EALs are examples of conditions and

indications that were considered to meet the criteria of the IC. These examples were not 

intended to be all encompassing, and some may not apply to a particular facility. Utilities should 

generally address each example EAL that applies to their site. If an example EAL does not apply 

because of its wording, e.g., specifies instrumentation not available at the site, the utility should 

identify other available means for entry into the IC. Ideally, the example EALs used will be 

unambiguous, expressed in site-specific nomenclature, and be readily discernible from control 

room instrumentation.

• Basis – provides information that explains the IC and example EALs. The bases are written to 

assist the personnel implementing the generic guidance into site-specific procedures. Sitespecific deviations from the IC/EALs should be compared to the Basis for that IC to ensure that 

the fundamental intent of each IC/EAL is met. Some bases provide information intended to 

assist with establishing site-specific instrumentation values. Appendices A, C, D, and E provide 

detailed guidance on implementing their corresponding Recognition Categories.

For Recognition Category F, basis information is presented in a format consistent with Tables 3 

and 4. The presentation method shown for Fission Product Barrier Function Matrix was chosen to 

clearly show the synergism among the EALs and to support more accurate dynamic assessments. 

Other acceptable methods of achieving these goals which are currently in use include flow charts, 

block diagrams, and checklist tables. Utilities selecting these alternative need to ensure that all 

possible EAL combinations in the Fission Product Barrier Function Matrix are addressed in their 

presentation method.

5.2 Generic Bases

The generic guidance has the primary threshold for NOUEs as operation outside the safety

envelope for the plant as defined by plant technical specifications, including LCOs and Action 

Statement Times. In addition, certain precursors of more serious events such as loss of offsite AC 

power and earthquakes are included in NOUE IC/EALs. This provides a clear demarcation

between the lowest emergency class and "non-emergency" notifications specified by 10 CFR 

50.72.

For a number of Alerts, IC/EALs are chosen based on hazards which may cause damage to plant 

safety functions (i.e., tornadoes, hurricanes, FIRE in plant VITAL AREAs) or require additional help 

directly (control room evacuation) and thus increased monitoring of the plant is warranted. The 

symptom-based and barrier-based IC/EALs are sufficiently anticipatory to address the results of 

multiple failures, regardless of whether there is or is not a common cause. Declaration of the Alert 

will already result in the manning of the TSC for assistance and additional monitoring. Thus, direct 

escalation to the Site Area Emergency is unnecessary. Other Alerts, that have been specified, 

correspond to conditions which are consistent with the emergency class description.

The basis for declaring a Site Area Emergency and General Emergency is primarily the extent and 

severity of fission product barrier challenges, based on plant conditions as presently known or as 

can be reasonably projected. 
Revision 01/2003 5.3

With regard to the Hazards Recognition Category, the existence of a hazard that represents a 

potential degradation in the level of safety of the plant is the basis of NOUE classification. If the 

hazard results in VISIBLE DAMAGE to plant structures or equipment associated with safety

systems or if system performance is affected, the event may be escalated to an Alert. The

reference to “duration” or to “damage” to safety systems is intended only to size the event.

Consequential damage from such hazards, if observed, would be the basis for escalation to Site 

Area Emergency or General Emergency, by entry to System Malfunction or Fission Product Barrier 

IC/EALs.

5.3 Site Specific Implementation

The guidance presented here is not intended to be applied to plants as-is. The generic guidance is 

intended to give the logic for developing site-specific IC/EALs using site-specific IC/EAL

presentation methods. Each utility will need to revise the IC/EALs to meet site-specific needs with 

regard to instrumentation, nomenclature, plant arrangement, and method of presentation, etc. Such 

revision is expected and encouraged provided that the intent of the generic guidance is retained. 

Deviations from the intent may be acceptable, but will need to be justified during regulatory review. 

Items associated with presentation, e.g., format, sequencing of IC/EALs, IC numbering, recognition 

categories are at the option of the utility. 

The generic guidance includes both ICs and example EALs. It is the intent of this guidance that both 

be included in the site-specific implementation. Each serves a specific purpose. The IC is intended 

to be the fundamental criteria for the declaration, whereas, the EALs are intended to represent 

unambiguous examples of conditions that may meet the IC. There may be unforeseen events, or 

combinations of events, for which the EALs may not be exceeded, but in the judgment of the 

Emergency Director, the intent of the IC may be met. While the generic guidance does include 

Emergency Director judgment ICs, the additional detail in the individual ICs will facilitate

classifications over the broad guidance of the ED judgment ICs. 

For sites involving more than one reactor unit, consideration needs to be given to how events 

involving shared safety functions may affect more than one unit, and whether or not this may be a 

factor in escalating the event.

State and local requirements have not been reflected in the generic guidance and should be 

considered on a case-by-case basis with appropriate state and local emergency response

organizations.

Although not a requirement, utilities should consider either preparing a basis document or including 

basis information with the IC/EALs. The bases provided for each IC/EAL will provide a starting point 

for developing these site-specific bases. This information may assist the Emergency Director in 

making classifications, particularly those involving judgment or multiple events. The basis

information may be useful in training, for explaining event classifications to offsite officials, and 

would facilitate regulatory review and approval of the classification scheme.

5.4 Definitions

In the IC/EALs, selected words have been set in all capital letters. These words are defined terms 

having specific meanings as they relate to this procedure. Definitions of these terms are provided 

below. 

AFFECTING SAFE SHUTDOWN: Event in progress has adversely affected functions that are 

necessary to bring the plant to and maintain it in the applicable HOT or COLD SHUTDOWN

condition. Plant condition applicability is determined by Technical Specification LCOs in effect.
Revision 01/2003 5.4

Example 1: Event causes damage that results in entry into an LCO that requires the plant 

to be placed in HOT SHUTDOWN. HOT SHUTDOWN is achievable, but COLD

SHUTDOWN is not. This event is not “AFFECTING SAFE SHUTDOWN.”

Example 2: Event causes damage that results in entry into an LCO that requires the plant 

to be placed in COLD SHUTDOWN. HOT SHUTDOWN is achievable, but COLD

SHUTDOWN is not. This event is “AFFECTING SAFE SHUTDOWN.”

BOMB: refers to an explosive device suspected of having sufficient force to damage plant systems 

or structures.

CIVIL DISTURBANCE: is a group of (site-specific #) or more persons violently protesting station 

operations or activities at the site.

CONFINEMENT BOUNDARY: is the barrier(s) between areas containing radioactive substances 

and the environment.

CONTAINMENT CLOSURE: (PWR) is defined by site-specific procedure. (BWR) is considered 

to be Secondary Containment as required by Technical Specifications.

EXPLOSION: is a rapid, violent, unconfined combustion, or catastrophic failure of pressurized 

equipment that imparts energy of sufficient force to potentially damage permanent structures, 

systems, or components.

EXTORTION: is an attempt to cause an action at the station by threat of force.

FAULTED: (PWRs) in a steam generator, the existence of secondary side leakage that results in 

an uncontrolled decrease in steam generator pressure or the steam generator being completely 

depressurized.

FIRE: is 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: is a person(s) held as leverage against the station to ensure that demands will be met 

by the station.

HOSTILE FORCE: one or more individuals who are engaged in a determined assault, overtly or by 

stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing 

destruction.

IMMEDIATELY DANGEROUS TO LIFE AND HEALTH (IDLH): A condition that either poses an 

immediate threat to life and health or an immediate threat of severe exposure to contaminants 

which are likely to have adverse delayed effects on health.

INTRUSION / INTRUDER: is a person(s) present in a specified area without authorization. 

Discovery of a BOMB in a specified area is indication of INTRUSION into that area by a HOSTILE 

FORCE.

LOWER FLAMMABILITY LIMIT (LFL): The minimum concentration of a combustible substance 

that is capable of propagating a flame through a homogenous mixture of the combustible and a 

gaseous oxidizer.
Revision 01/2003 5.5

NORMAL PLANT OPERATIONS: activities at the plant site associated with routine testing,

maintenance, or equipment operations, in accordance with normal operating or administrative

procedures. Entry into abnormal or emergency operating procedures, or deviation from normal 

security or radiological controls posture, is a departure from NORMAL PLANT OPERATIONs.

PROTECTED AREA: is an area which normally encompasses all controlled areas within the 

security protected area fence (site-specific).

RUPTURED: (PWRs) in a steam generator, existence of primary-to-secondary leakage of a 

magnitude sufficient to require or cause a reactor trip and safety injection.

SABOTAGE: is deliberate damage, mis-alignment, or mis-operation of plant equipment with the 

intent to render the equipment inoperable. Equipment found tampered with or damaged due to 

malicious mischief may NOT meet the definition of SABOTAGE until this determination is made by 

security supervision.

SIGNIFICANT TRANSIENT: is an UNPLANNED event involving one or more of the following: (1) 

automatic turbine runback >25% thermal reactor power, (2) electrical load rejection >25% full 

electrical load, (3) Reactor Trip, (4) Safety Injection Activation, or (5) thermal power oscillations 

>10%

STRIKE ACTION: is a work stoppage within the PROTECTED AREA by a body of workers to 

enforce compliance with demands made on (site-specific). The STRIKE ACTION must threaten to 

interrupt NORMAL PLANT OPERATIONs.

UNPLANNED: a parameter change or an event that is not the result of an intended evolution and 

requires corrective or mitigative actions.

VALID: an indication, report, or condition, is considered to be VALID when it is verified by (1) an 

instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct 

observation by plant personnel, such that doubt related to the indicator’s operability, the condition’s 

existence, or the report’s accuracy is removed. Implicit in this definition is the need for timely 

assessment.

VISIBLE DAMAGE: is damage to equipment or structure that is readily observable without

measurements, testing, or analysis. Damage is sufficient to cause concern regarding the continued 

operability or reliability of affected safety structure, system, or component. Example damage

includes: deformation due to heat or impact, denting, penetration, rupture, cracking, paint blistering. 

Surface blemishes (e.g., paint chipping, scratches) should not be included.

VITAL AREA: is any area, normally within the PROTECTED AREA, which contains equipment, 

systems, components, or material, the failure, destruction, or release of which could directly or 

indirectly endanger the public health and safety by exposure to radiation (site-specific).
Revision 01/2003 5-A-1

Table 5-A-1

Recognition Category A

Abnormal Rad Levels / Radiological Effluent

INITIATING CONDITION MATRIX

NOUE ALERT SITE AREA EMERGENCY GENERAL EMERGENCY

AU1 Any UNPLANNED Release of 

Gaseous or Liquid Radioactivity to the Environment 

that Exceeds Two Times the 

Radiological Effluent 

Technical Specifications for 

60 Minutes or Longer.

Op. Modes: All

AA1 Any UNPLANNED Release of 

Gaseous or Liquid Radioactivity 

to the Environment that 

Exceeds 200 Times the 

Radiological Effluent Technical 

Specifications for 15 Minutes 

or Longer.

Op. Modes: All

AS1 Offsite Dose Resulting from an 

Actual or Imminent Release of 

Gaseous Radioactivity Exceeds 

100 mR TEDE or 500 mR 

Thyroid CDE for the Actual or 

Projected Duration of the 

Release.

Op. Modes: All

AG1 Offsite Dose Resulting from an 

Actual or Imminent Release of 

Gaseous Radioactivity Exceeds 

1000 mR TEDE or 5000 mR 

Thyroid CDE for the Actual or 

Projected Duration of the 

Release Using Actual 

Meteorology.

Op. Modes: All

AU2 Unexpected Increase in Plant 

Radiation.

Op. Modes: All

AA3 Release of Radioactive Material 

or Increases in Radiation 

Levels Within the Facility That 

Impedes Operation of Systems 

Required to Maintain Safe 

Operations or to Establish or 

Maintain Cold Shutdown

Op. Modes: All

AA2 Damage to Irradiated Fuel or 

Loss of Water Level that Has or 

Will Result in the Uncovering of 

Irradiated Fuel Outside the 

Reactor Vessel.

Op. Modes: All
Revision 01/2003 5-A-2

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Revision 01/2003 5-A-3

ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT

AU1

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

Any UNPLANNED Release of Gaseous or Liquid Radioactivity to the Environment that 

Exceeds Two Times the Radiological Effluent Technical Specifications for 60 Minutes or 

Longer.

Operating Mode Applicability: All

Example Emergency Action Levels: (1 or 2 or 3 or 4 or 5)

1. VALID reading on any effluent monitor that exceeds two times the alarm setpoint established 

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

2. VALID reading on one or more of the following radiation monitors that exceeds the reading 

shown for 60 minutes or longer: 

 (site-specific list) 

3. Confirmed sample analyses for gaseous or liquid releases indicates concentrations or

release rates, with a release duration of 60 minutes or longer, in excess of two times (sitespecific technical specifications).

4. VALID reading on perimeter radiation monitoring system greater than 0.10 mR/hr above

normal background sustained for 60 minutes or longer [for sites having telemetered perimeter 

monitors].

5. VALID indication on automatic real-time dose assessment capability greater than (sitespecific value) for 60 minutes or longer [for sites having such capability].

Basis:

Refer to Appendix A for a detailed basis of the radiological effluent IC/EALs.

This IC addresses a potential or actual decrease in the level of safety of the plant as indicated by a 

radiological release that exceeds regulatory commitments for an extended period of time. Nuclear 

power plants incorporate features intended to control the release of radioactive effluents to the 

environment. Further, there are administrative controls established to prevent unintentional

releases, or control and monitor intentional releases. These controls are located in the Offsite Dose 

Calculation Manual (ODCM), and for plants that have not implemented Generic Letter 89-01, in the 

Radiological Effluent Technical Specifications (RETS). The occurrence of extended, uncontrolled 

radioactive releases to the environment is indicative of a degradation in these features and/or 

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

separate initiating conditions and EALs.

The RETS multiples are specified in ICs AU1 and AA1 only to distinguish between non-emergency 

conditions, and from each other. While these multiples obviously correspond to an offsite dose or 
Revision 01/2003 5-A-4

dose rate, the emphasis in classifying these events is the degradation in the level of safety of the 

plant, NOT the magnitude of the associated dose or dose rate. Releases should not be prorated or 

averaged. For example, a release exceeding 4x RETS for 30 minutes does not meet the threshold 

for this IC.

UNPLANNED, as used in this context, includes any release for which a radioactivity discharge 

permit was not prepared, or a release that exceeds the conditions (e.g., minimum dilution flow, 

maximum discharge flow, alarm setpoints, etc.) on the applicable permit. The Emergency Director 

should not wait until 60 minutes has elapsed, but should declare the event as soon as it is 

determined that the release duration has or will likely exceed 60 minutes. Also, if an ongoing release 

is detected and the starting time for that release is unknown, the Emergency Director should, in the 

absence of data to the contrary, assume that the release has exceeded 60 minutes.

EAL #1 addresses radioactivity releases, that for whatever reason, cause effluent radiation monitor 

readings to exceed two times the Technical Specification limit and releases are not terminated 

within 60 minutes. This alarm setpoint may be associated with a planned batch release, or a 

continuous release path. In either case, the setpoint is established by the ODCM to warn of a 

release that is not in compliance with the RETS. Indexing the EAL threshold to the ODCM setpoints 

in this manner insures that the EAL threshold will never be less than the setpoint established by a 

specific discharge permit.

EAL #2 is intended for licensees that have established effluent monitoring on non-routine release 

pathways for which a discharge permit would not normally be prepared. The ODCM establishes a 

methodology for determining effluent radiation monitor setpoints. The ODCM specifies default 

source terms and, for gaseous releases, prescribes the use of pre-determined annual average 

meteorology in the most limiting downwind sector for showing compliance with the regulatory 

commitments. These monitor reading EALs should be determined using this methodology.

EAL #3 addresses uncontrolled releases that are detected by sample analyses, particularly on 

unmonitored pathways, e.g., spills of radioactive liquids into storm drains, heat exchanger leakage 

in river water systems, etc. 

The 0.10 mR/hr value in EAL #4 is based on a release rate not exceeding 500 mrem per year, as 

provided in the ODCM / RETS, prorated over 8766 hours, multiplied by two, and rounded. (500 ÷

8766 × 2 = 0.114). This is also the basis of the site specific value in EAL #5. 

EALs #1 and #2 directly correlate with the IC since annual average meteorology is required to be 

used in showing compliance with the RETS and is used in calculating the alarm setpoints. EALs #4 

and #5 are a function of actual meteorology, which will likely be different from the limiting annual 

average value. Thus, there will likely be a numerical inconsistency. However, the fundamental basis 

of this IC is NOT a dose or dose rate, but rather the degradation in the level of safety of the plant 

implied by the uncontrolled release. Exceeding EAL #4 or EAL #5 is an indication of an uncontrolled 

release meeting the fundamental basis for this IC.
Revision 01/2003 5-A-5

ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT

AU2

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

Unexpected Increase in Plant Radiation.

Operating Mode Applicability: All

Example Emergency Action Levels: (1 or 2)

1. a. VALID (site-specific) indication of uncontrolled water level decrease in the reactor refueling 

cavity, spent fuel pool, or fuel transfer canal with all irradiated fuel assemblies remaining 

covered by water.

AND

b. Unplanned VALID (site-specific) Direct Area Radiation Monitor reading increases 

2. Unplanned VALID Direct Area Radiation Monitor readings increases by a factor of 1000 over 

normal* levels.

*Normal levels can be considered as the highest reading in the past twenty-four hours 

excluding the current peak value.

Basis:

This IC addresses increased radiation levels as a result of water level decreases above the RPV 

flange or events that have resulted, or may result, in unexpected increases in radiation dose rates 

within plant buildings. These radiation increases represent a loss of control over radioactive

material and may represent a potential degradation in the level of safety of the plant. 

In light of Reactor Cavity Seal failure incidents at two different PWRs and loss of water in the Spent 

Fuel Pit/Fuel Transfer Canal at a BWR, explicit coverage of these types of events via EAL #1 is 

appropriate given their potential for increased doses to plant staff. Classification as a NOUE is 

warranted as a precursor to a more serious event. Site-specific indications may include

instrumentation such as water level and local area radiation monitors, and personnel (e.g., refueling 

crew) reports. If available, security video cameras may allow remote observation. Depending on 

available level instrumentation, the declaration threshold may need to be based on indications of 

water makeup rate or decrease in refueling water storage tank level. 

While a radiation monitor could detect an increase in dose rate due to a drop in the water level, it 

might not be a reliable indication of whether or not the fuel is covered. For example, the reading on 

an area radiation monitor located on the refueling bridge may increase due to planned evolutions 

such as head lift, or even a fuel assembly being raised in the manipulator mast. Generally,

increased radiation monitor indications will need to combined with another indicator (or personnel 

report) of water loss. For refueling events where the water level drops below the RPV flange 

classification would be via CU2. This event escalates to an Alert per IC AA2 if irradiated fuel outside 

the reactor vessel is uncovered. For events involving irradiated fuel in the reactor vessel, escalation 

would be via the Fission Product Barrier Matrix for events in operating modes 1-4.
Revision 01/2003 5-A-6

EAL #2 addresses UNPLANNED increases in in-plant radiation levels that represent a degradation 

in the control of radioactive material, and represent a potential degradation in the level of safety of 

the plant. This event escalates to an Alert per IC AA3 if the increase in dose rates impedes 

personnel access necessary for safe operation. 
Revision 01/2003 5-A-7

ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT

AA1

Initiating Condition -- ALERT

Any UNPLANNED Release of Gaseous or Liquid Radioactivity to the Environment that 

Exceeds 200 Times the Radiological Effluent Technical Specifications for 15 Minutes or 

Longer.

Operating Mode Applicability: All

Example Emergency Action Levels: (1 or 2 or 3 or 4 or 5)

1. VALID reading on any effluent monitor that exceeds 200 times the alarm setpoint established 

by a current radioactivity discharge permit for 15 minutes or longer.

2. VALID reading on one or more of the following radiation monitors that exceeds the reading 

shown for 15 minutes or longer: 

 (site-specific list) 

3. Confirmed sample analyses for gaseous or liquid releases indicates concentrations or

release rates, with a release duration of 15 minutes or longer, in excess of 200 times (sitespecific technical specifications).

4. VALID reading on perimeter radiation monitoring system greater than 10.0 mR/hr above

normal background sustained for 15 minutes or longer [for sites having telemetered perimeter 

monitors].

5. VALID indication on automatic real-time dose assessment capability greater than (sitespecific value) for 15 minutes or longer [for sites having such capability].

Basis:

Refer to Appendix A for a detailed basis of the radiological effluent IC/EALs.

This IC addresses a potential or actual decrease in the level of safety of the plant as indicated by a 

radiological release that exceeds regulatory commitments for an extended period of time. Nuclear 

power plants incorporate features intended to control the release of radioactive effluents to the 

environment. Further, there are administrative controls established to prevent unintentional

releases, or control and monitor intentional releases. These controls are located in the Offsite Dose 

Calculation Manual (ODCM), and for plants that have not implemented Generic Letter 89-01, in the 

Radiological Effluent Technical Specifications (RETS). The occurrence of extended, uncontrolled 

radioactive releases to the environment is indicative of a degradation in these features and/or 

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

separate initiating conditions and EALs.

The RETS multiples are specified in ICs AU1 and AA1 only to distinguish between non-emergency 

conditions, and from each other. While these multiples obviously correspond to an 
Revision 01/2003 5-A-8

offsite dose or dose rate, the emphasis in classifying these events is the degradation in the level of 

safety of the plant, NOT the magnitude of the associated dose or dose rate. Releases should not be 

prorated or averaged. 

UNPLANNED, as used in this context, includes any release for which a radioactivity discharge 

permit was not prepared, or a release that exceeds the conditions (e.g., minimum dilution flow, 

maximum discharge flow, alarm setpoints, etc.) on the applicable permit. The Emergency Director 

should not wait until 15 minutes has elapsed, but should declare the event as soon as it is 

determined that the release duration has or will likely exceed 15 minutes. Also, if an ongoing release 

is detected and the starting time for that release is unknown, the Emergency Director should, in the 

absence of data to the contrary, assume that the release has exceeded 15 minutes.

EAL #1 addresses radioactivity releases that for whatever reason cause effluent radiation monitor 

readings that exceed two hundred times the alarm setpoint established by the radioactivity

discharge permit. This alarm setpoint may be associated with a planned batch release, or a 

continuous release path. In either case, the setpoint is established by the ODCM to warn of a 

release that is not in compliance with the RETS. Indexing the EAL threshold to the ODCM setpoints 

in this manner insures that the EAL threshold will never be less than the setpoint established by a 

specific discharge permit.

EAL #2 is similar to EAL #1, but is intended to address effluent or accident radiation monitors on 

non-routine release pathways (i.e., for which a discharge permit would not normally be prepared). 

The ODCM establishes a methodology for determining effluent radiation monitor setpoints. The 

ODCM specifies default source terms and, for gaseous releases, prescribes the use of predetermined annual average meteorology in the most limiting downwind sector for showing

compliance with the regulatory commitments. These monitor reading EALs should be determined 

using this methodology.

EAL #3 addresses uncontrolled releases that are detected by sample analyses, particularly on 

unmonitored pathways, e.g., spills of radioactive liquids into storm drains, heat exchanger leakage 

in river water systems, etc. 

The 10.0 mR/hr value in EAL #4 is based on a release rate not exceeding 500 mrem per year, as 

provided in the ODCM / RETS, prorated over 8766 hours, multiplied by 200, and rounded. (500 ÷

8766 × 200 = 11.4). This is also the basis of the site specific value in EAL #5. 

EALs #1 and #2 directly correlate with the IC since annual average meteorology is required to be 

used in showing compliance with the RETS and is used in calculating the alarm setpoints. EALs #4 

and #5 are a function of actual meteorology, which will likely be different from the limiting annual 

average value. Thus, there will likely be a numerical inconsistency. However, the fundamental basis 

of this IC is NOT a dose or dose rate, but rather the degradation in the level of safety of the plant 

implied by the uncontrolled release. Exceeding EAL #4 or EAL #5 is an indication of an uncontrolled 

release meeting the fundamental basis for this IC.

Due to the uncertainty associated with meteorology, emergency implementing procedures should 

call for the timely performance of dose assessments using actual (real-time) meteorology in the 

event of a gaseous radioactivity release of this magnitude. The results of these assessments 

should be compared to the ICs AS1 and AG1 to determine if the event classification should be 

escalated. Contrary to the practices specified in revision 2 of this document, classification should 

not be delayed pending the results of these dose assessments.
Revision 01/2003 5-A-9

ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT

AA2

Initiating Condition -- ALERT

Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the 

Uncovering of Irradiated Fuel Outside the Reactor Vessel.

Operating Mode Applicability: All

Example Emergency Action Levels: (1 or 2)

1. A VALID (site-specific) alarm or reading on one or more of the following radiation monitors: 

(site-specific monitors)

Refuel Floor Area Radiation Monitor

Fuel Handling Building Ventilation Monitor

Refueling Bridge Area Radiation Monitor

2. Water level less than (site-specific) feet for the reactor refueling cavity, spent fuel pool and 

fuel transfer canal that will result in irradiated fuel uncovering.

Basis:

This IC addresses specific events that have resulted, or may result, in unexpected increases in 

radiation dose rates within plant buildings, and may be a precursor to a radioactivity release to the 

environment. These events represent a loss of control over radioactive material and represent a 

degradation in the level of safety of the plant. These events escalate from IC AU2 in that fuel activity 

has been released, or is anticipated due to fuel heatup. This IC applies to spent fuel requiring water 

coverage and is not intended to address spent fuel which is licensed for dry storage, which is 

discussed in IC E-AU1.

EAL #1 addresses radiation monitor indications of fuel uncovery and/or fuel damage. Increased 

readings on ventilation monitors may be indication of a radioactivity release from the fuel, confirming 

that damage has occurred. Increased background at the monitor due to water level decrease may 

mask increased ventilation exhaust airborne activity and needs to be considered. While a radiation 

monitor could detect an increase in dose rate due to a drop in the water level, it might not be a 

reliable indication of whether or not the fuel is covered. For example, the monitor could in fact be 

properly responding to a known event involving transfer or relocation of a source, stored in or near 

the fuel pool or responding to a planned evolution such as removal of the reactor head. Application 

of these Initiating Conditions requires understanding of the actual radiological conditions present in 

the vicinity of the monitor. Information Notice No. 90-08, "KR-85 Hazards from Decayed Fuel"

should be considered in establishing radiation monitor EAL thresholds. 

In EAL #2, site-specific indications may include instrumentation such as water level and local area 

radiation monitors, and personnel (e.g., refueling crew) reports. If available, security video cameras 

may allow remote observation. Depending on available level indication, the declaration threshold 

may need to be based on indications of water makeup rate or decrease in refueling water storage 

tank level. 

Escalation, if appropriate, would occur via IC AS1 or AG1 or Emergency Director judgment. 
Revision 01/2003 5-A-10

ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT

AA3

Initiating Condition -- ALERT

Release of Radioactive Material or Increases in Radiation Levels Within the Facility That 

Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or 

Maintain Cold Shutdown

Operating Mode Applicability: All

Example Emergency Action Levels: (1 or 2)

1. VALID (site-specific) radiation monitor readings GREATER THAN 15 mR/hr in areas requiring 

continuous occupancy to maintain plant safety functions:

(Site-specific) list

2. VALID (site-specific) radiation monitor readings GREATER THAN <site specific> values in 

areas requiring infrequent access to maintain plant safety functions.

(Site-specific) list

Basis:

This IC addresses increased radiation levels that impede necessary access to operating stations, 

or other areas containing equipment that must be operated manually or that requires local

monitoring, in order to maintain safe operation or perform a safe shutdown. It is this impaired ability 

to operate the plant that results in the actual or potential substantial degradation of the level of safety 

of the plant. The cause and/or magnitude of the increase in radiation levels is not a concern of this 

IC. The Emergency Director must consider the source or cause of the increased radiation levels 

and determine if any other IC may be involved. For example, a dose rate of 15 mR/hr in the control 

room may be a problem in itself. However, the increase may also be indicative of high dose rates in 

the containment due to a LOCA. In this latter case, an SAE or GE may be indicated by the fission 

product barrier matrix ICs. 

At multiple-unit sites, the example EALs could result in declaration of an Alert at one unit due to a 

radioactivity release or radiation shine resulting from a major accident at the other unit. This is 

appropriate if the increase impairs operations at the operating unit.

This IC is not meant to apply to increases in the containment dome radiation monitors as these are 

events which are addressed in the fission product barrier matrix ICs. Nor is it intended to apply to 

anticipated temporary increases due to planned events (e.g., incore detector movement, radwaste 

container movement, depleted resin transfers, etc.)

Areas requiring continuous occupancy includes the control room and, as appropriate to the site, any 

other control stations that are manned continuously, such as a radwaste control room or a central 

security alarm station. 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 
Revision 01/2003 5-A-11

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.

For areas requiring infrequent access, the site-specific value(s) should be based on radiation levels 

which result in exposure control measures intended to maintain doses within normal occupational 

exposure guidelines and limits (i.e., 10 CFR 20), and in doing so, will impede necessary access. As 

used here, impede, includes hindering or interfering provided that the interference or delay is 

sufficient to significantly threaten the safe operation of the plant. 

Emergency planners developing the site-specific lists may refer to the site's abnormal operating 

procedures, emergency operating procedures, the 10 CFR 50 Appendix R analysis, and/or, the 

analyses performed in response to Section 2.1.6b of NUREG-0578, "TMI-2 Lessons Learned Task 

Force Status Report and Short-term Recommendations", when identifying areas containing safe 

shutdown equipment. Do not use the dose rates postulated in the NUREG-0578 analyses as a 

basis for the radiation monitor readings for this IC, as the design envelope for the NUREG-0578 

analyses correspond to general emergency conditions.
Revision 01/2003 5-A-12

ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT

AS1

Initiating Condition -- SITE AREA EMERGENCY

Offsite Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity 

Exceeds 100 mR TEDE or 500 mR Thyroid CDE for the Actual or Projected Duration of 

the Release.

Operating Mode Applicability: All

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

Note: If dose assessment results are available at the time of declaration, the classification should 

be based on EAL #2 instead of EAL #1.While necessary declarations should not be delayed 

awaiting results, the dose assessment should be initiated / completed in order to determine 

if the classification should be subsequently escalated.

1. VALID reading on one or more of the following radiation monitors that exceeds or is expected 

to exceed the reading shown for 15 minutes or longer:

 (site-specific list)

2. Dose assessment using actual meteorology indicates doses greater than 100 mR TEDE or 

500 mR thyroid CDE at or beyond the site boundary.

3. A VALID reading sustained for 15 minutes or longer on perimeter radiation monitoring system 

greater than 100 mR/hr. [for sites having telemetered perimeter monitors]

4. Field survey results indicate closed window dose rates exceeding 100 mR/hr expected to 

continue for more than one hour; or analyses of field survey samples indicate thyroid CDE of 

500 mR for one hour of inhalation, at or beyond the site boundary.

Basis:

Refer to Appendix A for a detailed basis of the radiological effluent IC/EALs.

This IC addresses radioactivity releases that result in doses at or beyond the site boundary that 

exceed a small fraction of the EPA Protective Action Guides (PAGs). Releases of this magnitude 

are associated with the failure of plant systems needed for the protection of the public. While these 

failures are addressed by other ICs, this IC provides appropriate diversity and addresses events 

which may not be able to be classified on the basis of plant status alone, e.g., fuel handling accident 

in spent fuel building.

The TEDE dose is set at 10% of the EPA PAG, while the 500 mR thyroid CDE was established in 

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

The Emergency Director should not wait until 15 minutes has elapsed, but should declare the event 

as soon as it is determined that the release duration has or will likely exceed 15 minutes. 
Revision 01/2003 5-A-13

The (site specific) monitor list in EAL #1 should include monitors on all potential release pathways. 

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 

calculate child thyroid CDE. Utility IC/EALs need to be consistent with those of the states involved in 

the facility’s emergency planning zone. 

The monitor reading EALs should be determined using a dose assessment method that back 

calculates from the dose values specified in the IC. The meteorology and source term (noble 

gases, particulates, and halogens) used should be the same as those used for determining the 

monitor reading EALs in ICs AU1 and AA1. This protocol will maintain intervals between the EALs 

for the four classifications. Since doses are generally not monitored in real-time, it is suggested that 

a release duration of one hour be assumed, and that the EALs be based on a site boundary (or 

beyond) dose of 100 mR/hour whole body or 500 mR/hour thyroid, whichever is more limiting (as 

was done for EALs #3 and #4). If individual site analyses indicate a longer or shorter duration for the 

period in which the substantial portion of the activity is released, the longer duration should be used.

Since dose assessment is based on actual meteorology, whereas the monitor reading EALs are 

not, the results from these assessments may indicate that the classification is not warranted, or 

may indicate that a higher classification is warranted. For this reason, emergency implementing 

procedures should call for the timely performance of dose assessments using actual meteorology 

and release information. If the results of these dose assessments are available when the

classification is made (e.g., initiated at a lower classification level), the dose assessment results 

override the monitor reading EALs. Contrary to the practices specified in revision 2 of this 

document, classification should not be delayed pending the results of these dose assessments.
Revision 01/2003 5-A-14

ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT

AG1

Initiating Condition -- GENERAL EMERGENCY

Offsite Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity 

Exceeds 1000 mR TEDE or 5000 mR Thyroid CDE for the Actual or Projected Duration 

of the Release Using Actual Meteorology.

Operating Mode Applicability: All

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

Note: If dose assessment results are available at the time of declaration, the classification should 

be based on EAL #2 instead of EAL #1.While necessary declarations should not be delayed 

awaiting results, the dose assessment should be initiated / completed in order to determine 

if the classification should be subsequently escalated.

1. VALID reading on one or more of the following radiation monitors that exceeds or expected to 

exceed the reading shown for 15 minutes or longer: 

 (site-specific list)

2. Dose assessment using actual meteorology indicates doses greater than 1000 mR TEDE or 

5000 mR thyroid CDE at or beyond the site boundary.

3. A VALID reading sustained for 15 minutes or longer on perimeter radiation monitoring system 

greater than 1000 mR/hr. [for sites having telemetered perimeter monitors]

4. Field survey results indicate closed window dose rates exceeding 1000 mR/hr expected to 

continue for more than one hour; or analyses of field survey samples indicate thyroid CDE of 

5000 mR for one hour of inhalation, at or beyond site boundary.

Basis:

Refer to Appendix A for a detailed basis of the radiological effluent IC/EALs.

This IC addresses radioactivity releases that result in doses at or beyond the site boundary that 

exceed the EPA Protective Action Guides (PAGs). Public protective actions will be necessary. 

Releases of this magnitude are associated with the failure of plant systems needed for the

protection of the public and likely involve fuel damage. While these failures are addressed by other 

ICs, this IC provides appropriate diversity and addresses events which may not be able to be 

classified on the basis of plant status alone. It is important to note that, for the more severe 

accidents, the release may be unmonitored or there may be large uncertainties associated with the 

source term and/or meteorology.

The Emergency Director should not wait until 15 minutes has elapsed, but should declare the event 

as soon as it is determined that the release duration has or will likely exceed 15 minutes. 

The (site specific) monitor list in EAL #1 should include monitors on all potential release pathways. 
Revision 01/2003 5-A-15

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 

calculate child thyroid CDE. Utility IC/EALs need to be consistent with those of the states involved in 

the facilities emergency planning zone. 

The monitor reading EALs should be determined using a dose assessment method that

backcalculates from the dose values specified in the IC. The meteorology and source term (noble 

gases, particulates, and halogens) used should be the same as those used for determining the 

monitor reading EALs in ICs AU1 and AA1. This protocol will maintain intervals between the EALs 

for the four classifications. Since doses are generally not monitored in real-time, it is suggested that 

a release duration of one hour be assumed, and that the EALs be based on a site boundary (or 

beyond) dose of 1000 mR/hour whole body or 5000 mR/hour thyroid, whichever is more limiting (as 

was done for EALs #3 and #4). If individual site analyses indicate a longer or shorter duration for the 

period in which the substantial portion of the activity is released, the longer duration should be used.

Since dose assessment is based on actual meteorology, whereas the monitor reading EALs are 

not, the results from these assessments may indicate that the classification is not warranted, or 

may indicate that a higher classification is warranted. For this reason, emergency implementing 

procedures should call for the timely performance of dose assessments using actual meteorology 

and release information. If the results of these dose assessments are available when the

classification is made (e.g., initiated at a lower classification level), the dose assessment results 

override the monitor reading EALs. Contrary to the practices specified in revision 2 of this 

document, classification should not be delayed pending the results of these dose assessments.
Revision 01/2003 5-C-1

Recognition Category C

Cold Shutdown/Refueling System Malfunction

INITIATING CONDITION MATRIX

NOUE ALERT SITE AREA EMERGENCY GENERAL EMERGENCY

CU1 RCS Leakage.

Op. Mode: Cold Shutdown

CA1 Loss of RCS Inventory.

Op. Modes: Cold Shutdown

CS1 Loss of RPV Inventory Affecting 

Core Decay Heat Removal 

Capability.

Op. Modes: Cold Shutdown

CG1 Loss of RPV Inventory Affecting 

Fuel Clad Integrity with 

Containment Challenged with 

Irradiated Fuel in the RPV.

Op. Modes: Cold Shutdown, 

Refueling

CU2 UNPLANNED Loss of RCS 

Inventory with Irradiated Fuel in 

the RPV

Op. Mode: Refueling

CA2 Loss of RPV Inventory with 

Irradiated Fuel in the RPV.

Op. Modes: Refueling

CS2 Loss of RPV Inventory Affecting 

Core Decay Heat Removal 

Capability with Irradiated Fuel in 

the RPV.

Op. Modes: Refueling

CU3 Loss of All Offsite Power to 

Essential Busses for Greater 

Than 15 Minutes.

Op. Modes: Cold Shutdown, 

Refueling

CA3 Loss of All Offsite Power and 

Loss of All Onsite AC Power to 

Essential Busses.

Op. Modes: Cold Shutdown, 

Refueling, Defueled

CU4 UNPLANNED Loss of Decay 

Heat Removal Capability with 

Irradiated Fuel in the RPV.

OP. Modes: Cold Shutdown, 

Refueling

CA4 Inability to Maintain Plant in Cold 

Shutdown with Irradiated Fuel in 

the RPV. 

Op. Modes: Cold Shutdown, 

Refueling

CU5 Fuel Clad Degradation.

Op. Modes: Cold Shutdown, 

Refueling

CU6 UNPLANNED Loss of All Onsite 

or Offsite Communications 

Capabilities. 

Op. Modes: Cold Shutdown, 

Refueling

CU7 UNPLANNED Loss of Required 

DC Power for Greater than 15 

Minutes.

Op. Modes: Cold Shutdown, 

Refueling
Revision 01/2003 5-C-2

CU8 Inadvertent Criticality.

Op Modes:, Cold Shutdown, 

Refueling
Revision 01/2003 5-C-3

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Revision 01/2003 5-C-5

SYSTEM MALFUNCTION

CU1

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

RCS Leakage.

Operating Mode Applicability: Cold Shutdown

Example Emergency Action Levels: (1 or 2)

1. Unidentified or pressure boundary leakage greater than 10 gpm.

2. Identified leakage greater than 25 gpm.

Basis:

This IC is included as a NOUE because it is considered to be a potential degradation of the level of 

safety of the plant. The 10 gpm value for the unidentified and pressure boundary leakage was 

selected as it is sufficiently large to be observable via normally installed instrumentation (e.g., 

Pressurizer level, RCS loop level instrumentation, etc...) or reduced inventory instrumentation such 

as level hose indication. Lesser values must generally be determined through time-consuming 

surveillance tests (e.g., mass balances). The EAL for identified leakage is set at a higher value due 

to the lesser significance of identified leakage in comparison to unidentified or pressure boundary 

leakage. Prolonged loss of RCS Inventory may result in escalation to the Alert level via either IC 

CA1 (Loss of RCS Inventory with Irradiated Fuel in the RPV) or CA4 (Inability to Maintain Plant in 

Cold Shutdown with Irradiated Fuel in the RPV).

The difference between CU1 and CU2 deals with the RCS conditions that exist between cold 

shutdown and refueling mode applicability. In cold shutdown the RCS will normally be intact and 

RCS inventory and level monitoring means such as Pressurizer level indication and makeup

volume control tank levels are normally available. In the refueling mode the RCS is not intact and 

RPV level and inventory are monitored by different means.

Expanded basis for these assumptions is provided in Appendix C.
Revision 01/2003 5-C-6

SYSTEM MALFUNCTION

CU2

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

UNPLANNED Loss of RCS Inventory with Irradiated Fuel in the RPV.

Operating Mode Applicability: Refueling

Example Emergency Action Levels: (1 or 2)

1. UNPLANNED RCS level decrease below the RPV flange for > 15 minutes

2. a. Loss of RPV inventory as indicated by unexplained {site-specific} sump and tank level 

increase

AND

b. RPV level cannot be monitored

Basis:

This IC is included as a NOUE because it may be a precursor of more serious conditions and, as 

result, is considered to be a potential degradation of the level of safety of the plant. Refueling 

evolutions that decrease RCS water level below the RPV flange are carefully planned and

procedurally controlled. An UNPLANNED event that results in water level decreasing below the 

RPV flange warrants declaration of a NOUE due to the reduced RCS inventory that is available to 

keep the core covered. The allowance of 15 minutes was chosen because it is reasonable to 

assume that level can be restored within this time frame using one or more of the redundant means 

of refill that should be available. If level cannot be restored in this time frame then it may indicate a 

more serious condition exists. Continued loss of RCS Inventory will result in escalation to the Alert 

level via either IC CA2 (Loss of RPV Inventory with Irradiated Fuel in the RPV) or CA4 (Inability to 

Maintain Plant in Cold Shutdown with Irradiated Fuel in the RPV).

The difference between CU1 and CU2 deals with the RCS conditions that exist between cold 

shutdown and refueling modes. In cold shutdown the RCS will normally be intact and standard RCS 

inventory and level monitoring means are available. In the refueling mode the RCS is not intact and 

RPV level and inventory are monitored by different means.

In the refueling mode, normal means of core temperature indication and RCS level indication may 

not be available. Redundant means of RPV level indication will normally be installed (including the 

ability to monitor level visually) to assure that the ability to monitor level will not be interrupted. 

However, if all level indication were to be lost during a loss of RCS inventory event, the operators 

would need to determine that RPV inventory loss was occurring by observing sump and tank level 

changes. Sump and tank level increases must be evaluated against other potential sources of 

leakage such as cooling water sources inside the containment to ensure they are indicative of RCS 

leakage. Escalation to Alert would be via either CA2 or RCS heatup via CA4.

EAL 1 involves a decrease in RCS level below the top of the RPV flange that continues for 15 

minutes due to an UNPLANNED event. This EAL is not applicable to decreases in flooded reactor 

cavity level (covered by AU2 EAL1) until such time as the level decreases to the level of the vessel 
Revision 01/2003 5-C-7

flange. For BWRs, if RPV level continues to decrease and reaches the Low-Low ECCS Actuation 

Setpoint then escalation to CA2 would be appropriate. For PWRs, if RPV level continues to

decrease and reaches the Bottom ID of the RCS Loop then escalation to CA2 would be

appropriate. Note that the Bottom ID of the RCS Loop Setpoint should be the level equal to the 

bottom of the RPV loop penetration (not the low point of the loop).

Expanded basis for these assumptions is provided in Appendix C.
Revision 01/2003 5-C-8

SYSTEM MALFUNCTION

CU3

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes.

Operating Mode Applicability: Cold Shutdown

Refueling

Example Emergency Action Level:

1. a. Loss of power to (site-specific) transformers for greater than 15 minutes.

AND

b. At least (site-specific) emergency generators are supplying power to emergency busses.

Basis:

Prolonged loss of AC power reduces required redundancy and potentially degrades the level of 

safety of the plant by rendering the plant more vulnerable to a complete Loss of AC Power (e.g., 

Station Blackout). Fifteen minutes was selected as a threshold to exclude transient or momentary 

power losses.

Plants that have the capability to cross-tie AC power from a companion unit may take credit for the 

redundant power source in the associated EAL for this IC. Inability to effect the cross-tie within 15 

minutes warrants declaring a NOUE.
Revision 01/2003 5-C-9

SYSTEM MALFUNCTION

CU4

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

UNPLANNED Loss of Decay Heat Removal Capability with Irradiated Fuel in the RPV.

Operating Mode Applicability: Cold Shutdown

Refueling

Example Emergency Action Levels: (1 or 2)

1. An UNPLANNED event results in RCS temperature exceeding the Technical Specification 

cold shutdown temperature limit 

2. Loss of all RCS temperature and RPV level indication for > 15 minutes.

Basis:

This IC is included as a NOUE because it may be a precursor of more serious conditions and, as a 

result, is considered to be a potential degradation of the level of safety of the plant. In cold 

shutdown the ability to remove decay heat relies primarily on forced cooling flow. Operation of the 

systems that provide this forced cooling may be jeopardized due to the unlikely loss of electrical 

power or RCS inventory. Since the RCS usually remains intact in the cold shutdown mode a large 

inventory of water is available to keep the core covered. In cold shutdown the decay heat available 

to raise RCS temperature during a loss of inventory or heat removal event may be significantly 

greater than in the refueling mode. Entry into cold shutdown conditions may be attained within 

hours of operating at power. Entry into the refueling mode procedurally may not occur for typically 

100 hours {site-specific} or longer after the reactor has been shutdown. Thus the heatup threat and 

therefore the threat to damaging the fuel clad may be lower for events that occur in the refueling 

mode with irradiated fuel in the RPV (note that the heatup threat could be lower for cold shutdown 

conditions if the entry into cold shutdown was following a refueling). In addition, the operators 

should be able to monitor RCS temperature and RPV level so that escalation to the alert level via 

CA4 or CA1 will occur if required.

During refueling the level in the RPV will normally be maintained above the RPV flange. Refueling 

evolutions that decrease water level below the RPV flange are carefully planned and procedurally 

controlled. Loss of forced decay heat removal at reduced inventory may result in more rapid 

increases in RCS/RPV temperatures depending on the time since shutdown. Escalation to the Alert 

level via CA4 is provided should an UNPLANNED event result in RCS temperature exceeding the 

Technical Specification cold shutdown temperature limit for greater than 30 minutes with

CONTAINMENT CLOSURE not established.

Unlike the cold shutdown mode, normal means of core temperature indication and RCS level 

indication may not be available in the refueling mode. Redundant means of RPV level indication are 

therefore procedurally installed to assure that the ability to monitor level will not be interrupted. 

However, if all level and temperature indication were to be lost in either the cold shutdown of 

refueling modes, EAL 2 would result in declaration of a NOUE if either temperature or level 

indication cannot be restored within 15 minutes from the loss of both means of indication. 

Escalation to Alert would be via CA2 based on an inventory loss or CA4 based on exceeding its 

temperature criteria. 
Revision 01/2003 5-C-10

The Emergency Director must remain attentive to events or conditions that lead to the conclusion 

that exceeding the EAL threshold is imminent. If, in the judgment of the Emergency Director, an 

imminent situation is at hand, the classification should be made as if the threshold has been 

exceeded.

Expanded basis for these assumptions is provided in Appendix C.
Revision 01/2003 5-C-11

SYSTEM MALFUNCTION

CU5

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

Fuel Clad Degradation.

Operating Mode Applicability: Cold Shutdown

Refueling

Example Emergency Action Levels: (1 or 2)

1. (Site-specific) radiation monitor readings indicating fuel clad degradation greater than

Technical Specification allowable limits.

2. (Site-specific) coolant sample activity value indicating fuel clad degradation greater than

Technical Specification allowable limits.

Basis:

This IC is included as a NOUE because it is considered to be a potential degradation in the level of 

safety of the plant and a potential precursor of more serious problems. EAL #1 addresses sitespecific radiation monitor readings that provide indication of fuel clad integrity. EAL #2 addresses 

coolant samples exceeding coolant technical specifications for iodine spike. 
Revision 01/2003 5-C-12

SYSTEM MALFUNCTION

CU6

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

UNPLANNED Loss of All Onsite or Offsite Communications Capabilities. 

Operating Mode Applicability: Cold Shutdown

Refueling

Example Emergency Action Levels: (1 or 2)

1. Loss of all (site-specific list) onsite communications capability affecting the ability to perform 

routine operations.

2. Loss of all (site-specific list) offsite communications capability. 

Basis:

The purpose of this IC and its associated EALs is to recognize a loss of communications capability 

that either defeats the plant operations staff ability to perform routine tasks necessary for plant 

operations or the ability to communicate problems with offsite authorities. The loss of offsite

communications ability is expected to be significantly more comprehensive than the condition 

addressed by 10 CFR 50.72.

The availability of one method of ordinary offsite communications is sufficient to inform state and 

local authorities of plant problems. This EAL is intended to be used only when extraordinary means 

(e.g., relaying of information from radio transmissions, individuals being sent to offsite locations, 

etc.) are being utilized to make communications possible. 

Site-specific list for onsite communications loss must encompass the loss of all means of routine 

communications (e.g., commercial telephones, sound powered phone systems, page party system 

and radios / walkie talkies).

Site-specific list for offsite communications loss must encompass the loss of all means of

communications with offsite authorities. This should include the ENS, commercial telephone lines,

telecopy transmissions, and dedicated phone systems. 
Revision 01/2003 5-C-13

SYSTEM MALFUNCTION

CU7

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

UNPLANNED Loss of Required DC Power for Greater than 15 Minutes.

Operating Mode Applicability: Cold Shutdown

Refueling

Example Emergency Action Level:

1. a. UNPLANNED Loss of Vital DC power to required DC busses based on (site-specific) bus 

voltage indications.

AND

b. Failure to restore power to at least one required DC bus within 15 minutes from the time of 

loss.

Basis: 

The purpose of this IC and its associated EALs is to recognize a loss of DC power compromising 

the ability to monitor and control the removal of decay heat during Cold Shutdown or Refueling 

operations. This EAL is intended to be anticipatory in as much as the operating crew may not have 

necessary indication and control of equipment needed to respond to the loss.

UNPLANNED is included in this IC and EAL to preclude the declaration of an emergency as a result 

of planned maintenance activities. Routinely plants will perform maintenance on a Train related 

basis during shutdown periods. It is intended that the loss of the operating (operable) train is to be 

considered. If this loss results in the inability to maintain cold shutdown, the escalation to an Alert 

will be per CA4 "Inability to Maintain Plant in Cold Shutdown with Irradiated Fuel in the RPV."

(Site-specific) bus voltage should be based on the minimum bus voltage necessary for the

operation of safety related 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. Typically the value for the entire 

battery set is approximately 105 VDC. For a 60 cell string of batteries the cell voltage 1.75 Volts per 

cell. For a 58 string battery set the minimum voltage is typically 1.81 Volts per cell.
Revision 01/2003 5-C-14

SYSTEM MALFUNCTION

CU8

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

Inadvertent Criticality.

Operating Mode Applicability: Cold Shutdown

Refueling

Example Emergency Action Levels: (1 or 2)

1. An UNPLANNED extended positive period observed on nuclear instrumentation.

2. An UNPLANNED sustained positive startup rate observed on nuclear instrumentation.

Basis:

This IC addresses criticality events that occur in Cold Shutdown or Refueling modes (NUREG 

1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United 

States) such as fuel mis-loading events and inadvertent dilution events. This IC indicates a potential 

degradation of the level of safety of the plant, warranting a NOUE classification. This IC excludes 

inadvertent criticalities that occur during planned reactivity changes associated with reactor

startups (e.g., criticality earlier than estimated) which are addressed in the companion IC SU8.

This condition can be identified using period monitors/startup rate monitor. The terms “extended“ 

and “sustained” are used in order to allow exclusion of expected short term positive periods/startup 

rates from planned fuel bundle or control rod movements during core alteration for PWRs and 

BWRs. These short term positive periods/startup rates are the result of the increase in neutron 

population due to subcritical multiplication.

Escalation would be by Emergency Director Judgment.
Revision 01/2003 5-C-15

SYSTEM MALFUNCTION

CA1

Initiating Condition -- ALERT

Loss of RCS Inventory.

Operating Mode Applicability: Cold Shutdown

Example Emergency Action Levels: (1 or 2)

1. Loss of RCS inventory as indicated by RPV level less than {site-specific level}. 

(low-low ECCS actuation setpoint) (BWR) 

(bottom ID of the RCS loop) (PWR)

2. a. Loss of RCS inventory as indicated by unexplained {site-specific} sump and tank level 

increase

AND

b. RCS level cannot be monitored for > 15 minutes

Basis:

These example EALs serve as precursors to a loss of ability to adequately cool the fuel. The 

magnitude of this loss of water indicates that makeup systems have not been effective and may not 

be capable of preventing further RPV level decrease and potential core uncovery. This condition will 

result in a minimum classification of Alert. The BWR Low-Low ECCS Actuation Setpoint was 

chosen because it is a standard setpoint at which all available injection systems automatically start. 

The PWR Bottom ID of the RCS Loop Setpoint was chosen because at this level remote RCS level 

indication may be lost and loss of suction to decay heat removal systems has occurred. The 

Bottom ID of the RCS Loop Setpoint should be the level equal to the bottom of the RPV loop 

penetration (not the low point of the loop). The inability to restore and maintain level after reaching 

this setpoint would therefore be indicative of a failure of the RCS barrier.

In cold shutdown the decay heat available to raise RCS temperature during a loss of inventory or 

heat removal event may be significantly greater than in the refueling mode. Entry into cold shutdown 

conditions may be attained within hours of operating at power or hours after refueling is completed. 

Entry into the refueling mode procedurally may not occur for typically 100 hours {site-specific} or 

longer after the reactor has been shutdown. Thus the heatup threat and therefore the threat to 

damaging the fuel clad may be lower for events that occur in the refueling mode with irradiated fuel 

in the RPV (note that the heatup threat could be lower for cold shutdown conditions if the entry into 

cold shutdown was following a refueling). The above forms the basis for needing both a cold 

shutdown specific IC (CA1) and a refueling specific IC (CA2).

In the cold shutdown mode, normal RCS level and RPV level instrumentation systems will normally 

be available. However, if all level indication were to be lost during a loss of RCS inventory event, the 

operators would need to determine that RPV inventory loss was occurring by observing sump and 

tank level changes. Sump and tank level increases must be evaluated against other potential 

sources of leakage such as cooling water sources inside the containment to ensure they are 

indicative of RCS leakage. The 15-minute duration for the loss of level indication was chosen 
Revision 01/2003 5-C-16

because it is half of the CS1 Site Area Emergency EAL duration. The 15-minute duration allows 

CA1 to be an effective precursor to CS1. Significant fuel damage is not expected to occur until the 

core has been uncovered for greater than 1 hour per the analysis referenced in the CS1 basis. 

Therefore this EAL meets the definition for an Alert emergency.

The difference between CA1 and CA2 deals with the RCS conditions that exist between cold 

shutdown and refueling mode applicability. In cold shutdown the RCS will normally be intact and 

standard RCS inventory and level monitoring means are available. In the refueling mode the RCS 

is not intact and RPV level and inventory are monitored by different means.

If RPV level continues to decrease then escalation to Site Area will be via CS1 (Loss of Inventory 

Affecting Core Decay Heat Removal Capability with Irradiated Fuel in the RPV).

Expanded basis for these assumptions is provided in Appendix C.
Revision 01/2003 5-C-17

SYSTEM MALFUNCTION

CA2

Initiating Condition -- ALERT

Loss of RPV Inventory with Irradiated Fuel in the RPV.

Operating Mode Applicability: Refueling

Example Emergency Action Levels: (1 or 2)

1. Loss of RPV inventory as indicated by RPV level less than {site-specific level}. 

(low-low ECCS actuation setpoint) (BWR) 

(bottom ID of the RCS loop) (PWR)

2. a. Loss of RPV inventory as indicated by unexplained {site-specific} sump and tank level 

increase

AND

b. RPV level cannot be monitored for > 15 minutes

Basis:

These example EALs serve as precursors to a loss of heat removal. The magnitude of this loss of 

water indicates that makeup systems have not been effective and may not be capable of preventing 

further RPV level decrease and potential core uncovery. This condition will result in a minimum 

classification of Alert. The BWR Low-Low ECCS Actuation Setpoint was chosen because it is a 

standard setpoint at which all available injection systems automatically start. The Bottom ID of the 

RCS Loop Setpoint was chosen because at this level remote RCS level indication may be lost and 

loss of suction to decay heat removal systems may occur. The Bottom ID of the RCS Loop 

Setpoint should be the level equal to the bottom of the RPV loop penetration (not the low point of the 

loop). The inability to restore and maintain level after reaching this setpoint would therefore be 

indicative of a failure of the RCS barrier.

In cold shutdown the decay heat available to raise RCS temperature during a loss of inventory or 

heat removal event may be significantly greater than in the refueling mode. Entry into cold shutdown 

conditions may be attained within hours of operating at power or hours after refueling is completed. 

Entry into the refueling mode procedurally may not occur for typically 100 hours {site-specific} or 

longer after the reactor has been shutdown. Thus the heatup threat and therefore the threat to 

damaging the fuel clad may be lower for events that occur in the refueling mode with irradiated fuel 

in the RPV (note that the heatup threat could be lower for cold shutdown conditions if the entry into 

cold shutdown was following a refueling). The above forms the basis for needing both a cold 

shutdown specific IC (CA1) and a refueling specific IC (CA2).

In the refueling mode, normal means of RPV level indication may not be available. Redundant 

means of RPV level indication will be normally installed (including the ability to monitor level visually) 

to assure that the ability to monitor level will not be interrupted. However, if all level indication were 

to be lost during a loss of RCS inventory event, the operators would need to determine that RPV 

inventory loss was occurring by observing sump and tank level changes. Sump and tank level 

increases must be evaluated against other potential sources of leakage such as cooling water 
Revision 01/2003 5-C-18

sources inside the containment to ensure they are indicative of RCS leakage. The 15-minute 

duration for the loss of level indication was chosen because it is half of the CS2 Site Area

Emergency EAL duration. The 15-minute duration allows CA2 to be an effective precursor to CS2. 

Significant fuel damage is not expected to occur until the core has been uncovered for greater than 

1 hour per the analysis referenced in the CS2 basis. Therefore this EAL meets the definition for an 

Alert.

The difference between CA1 and CA2 deals with the RCS conditions that exist between cold 

shutdown and refueling mode applicability. In cold shutdown the RCS will normally be intact and 

standard RCS inventory and level monitoring means are available. In the refueling mode the RCS 

is not intact and RPV level and inventory are monitored by different means.

If RPV level continues to decrease then escalation to Site Area will be via CS1 (Loss of Inventory 

Affecting Core Decay Heat Removal Capability with Irradiated Fuel in the RPV).

Expanded basis for these assumptions is provided in Appendix C.
Revision 01/2003 5-C-19

SYSTEM MALFUNCTION

CA3

Initiating Condition -- ALERT

Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses.

Operating Mode Applicability: Cold Shutdown

Refueling

Defueled

Example Emergency Action Level:

1. a. Loss of power to (site-specific) transformers.

AND

b. Failure of (site-specific) emergency generators to supply power to emergency busses.

AND

c. Failure to restore power to at least one emergency bus within 15 minutes from the time of 

loss of both offsite and onsite AC power.

Basis:

Loss of all AC power compromises all plant safety systems requiring electric power including RHR, 

ECCS, Containment Heat Removal, Spent Fuel Heat Removal and the Ultimate Heat Sink. When in 

cold shutdown, refueling, or defueled mode the event can be classified as an Alert, because of the 

significantly reduced decay heat, lower temperature and pressure, increasing the time to restore 

one of the emergency busses, relative to that specified for the Site Area Emergency EAL.

Escalating to Site Area Emergency IC SS1, if appropriate, is by Abnormal Rad Levels / Radiological 

Effluent, or Emergency Director Judgment ICs. Fifteen minutes was selected as a threshold to 

exclude transient or momentary power losses.

Consideration should be given to operable loads necessary to remove decay heat or provide 

Reactor Vessel makeup capability when evaluating loss of AC power to essential busses. Even 

though an essential bus may be energized, if necessary loads (i.e., loads that if lost would inhibit 

decay heat removal capability or Reactor Vessel makeup capability) are not operable on the

energized bus then the bus should not be considered operable.
Revision 01/2003 5-C-20

SYSTEM MALFUNCTION

CA4

Initiating Condition -- ALERT

Inability to Maintain Plant in Cold Shutdown with Irradiated Fuel in the RPV.

Operating Mode Applicability: Cold Shutdown

Refueling

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

1. With CONTAINMENT CLOSURE and RCS integrity not established an UNPLANNED event 

results in RCS temperature exceeding the Technical Specification cold shutdown temperature 

limit.

2. With CONTAINMENT CLOSURE established and RCS integrity not established or RCS 

inventory reduced an UNPLANNED event results in RCS temperature exceeding the

Technical Specification cold shutdown temperature limit for greater than 20 minutes1. 

3. An UNPLANNED event results in RCS temperature exceeding the Technical Specification 

cold shutdown temperature limit for greater than 60 minutes1or results in an RCS pressure 

increase of greater than {site specific} psig.

Basis:

EAL 1 addresses complete loss of functions required for core cooling during refueling and cold 

shutdown modes when neither CONTAINMENT CLOSURE nor RCS integrity are established. RCS 

integrity is in place when the RCS pressure boundary is in its normal condition for the cold 

shutdown mode of operation (e.g., no freeze seals or nozzle dams). No delay time is allowed for 

EAL1 because the evaporated reactor coolant that may be released into the Containment during 

this heatup condition could also be directly released to the environment. 

EAL 2 addresses the complete loss of functions required for core cooling for > 20 minutes during 

refueling and cold shutdown modes when CONTAINMENT CLOSURE is established but RCS 

integrity is not established or RCS inventory is reduced (e.g., mid loop operation in PWRs). As in 

EAL 1, RCS integrity should be assumed to be in place when the RCS pressure boundary is in its 

normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams). 

The allowed 20 minute time frame was included to allow operator action to restore the heat removal 

function, if possible. The allowed time frame is consistent with the guidance provided by Generic 

Letter 88-17, "Loss of Decay Heat Removal" (discussed later in this basis) and is believed to be 

conservative given that a low pressure Containment barrier to fission product release is

established. Note 1 indicates that EAL 2 is not applicable if actions are successful in restoring an 

RCS heat removal system to operation and RCS temperature is being reduced within the 20 minute 

time frame.

 

1Note: if an RCS heat removal system is in operation within this time frame and RCS 

temperature is being reduced then this EAL is not applicable.
Revision 01/2003 5-C-21

EAL 3 addresses complete loss of functions required for core cooling for > 60 minutes during 

refueling and cold shutdown modes when RCS integrity is established. As in EAL 1 and 2, RCS 

integrity should be considered to be in place when the RCS pressure boundary is in its normal 

condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams). The 

status of CONTAINMENT CLOSURE in this EAL is immaterial given that the RCS is providing a 

high pressure barrier to fission product release to the environment. The 60 minute time frame 

should allow sufficient time to restore cooling without there being a substantial degradation in plant 

safety. The {site specific} pressure increase covers situations where, due to high decay heat 

loads, the time provided to restore temperature control, should be less than 60 minutes. The RCS 

pressure setpoint chosen should be 10 psig or the lowest pressure that the site can read on 

installed Control Board instrumentation that is equal to or greater than 10 psig. Note 1 indicates that 

EAL 3 is not applicable if actions are successful in restoring an RCS heat removal system to 

operation and RCS temperature is being reduced within the 60 minute time frame assuming that 

the RCS pressure increase has remained less than the site specific pressure value. 

Escalation to Site Area would be via CS1 or CS2 should boiling result in significant RPV level loss 

leading to core uncovery.

For PWRs, this IC and its associated EALs are based on 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 sequences that 

can cause core uncovery in 15 to 20 minutes and severe core damage within an hour after decay 

heat removal is lost. 

A loss of Technical Specification components alone is not intended to constitute an Alert. The 

same is true of a momentary UNPLANNED excursion above 200_F when the heat removal function 

is available. 

The Emergency Director must remain alert to events or conditions that lead to the conclusion that 

exceeding the EAL threshold is imminent. If, in the judgment of the Emergency Director, an

imminent situation is at hand, the classification should be made as if the threshold has been 

exceeded.

Expanded basis for these assumptions is provided in Appendix C.
Revision 01/2003 5-C-22

SYSTEM MALFUNCTION

CS1

Initiating Condition -- SITE AREA EMERGENCY

Loss of RPV Inventory Affecting Core Decay Heat Removal Capability.

Operating Mode Applicability: Cold Shutdown

Example Emergency Action Levels: (1 or 2)

1. With CONTAINMENT CLOSURE not established: 

a. RPV inventory as indicated by RPV level less than {site-specific level}

(6" below the low-low ECCS actuation setpoint) (BWR) 

(6" below the bottom ID of the RCS loop) (PWR)

OR

b. RPV level cannot be monitored for > 30 minutes with a loss of RPV inventory as 

indicated by unexplained {site-specific} sump and tank level increase

2. With CONTAINMENT CLOSURE established 

a. RPV inventory as indicated by RPV level less than TOAF 

OR

b. RPV level cannot be monitored for > 30 minutes with a loss of RPV inventory as 

indicated by either: 

• Unexplained {site-specific} sump and tank level increase

• Erratic Source Range Monitor Indication

Basis:

Under the conditions specified by this IC, continued decrease in RPV level is indicative of a loss of 

inventory control. Inventory loss may be due to an RPV breach, pressure boundary leakage, or 

continued boiling in the RPV.

In cold shutdown the decay heat available to raise RCS temperature during a loss of inventory or 

heat removal event may be significantly greater than in the refueling mode. Entry into cold shutdown 

conditions may be attained within hours of operating at power or hours after refueling is completed. 

Entry into the refueling mode procedurally may not occur for typically 100 hours {site-specific} or 

longer after the reactor has been shutdown. Thus the heatup threat and therefore the threat to 

damaging the fuel clad may be lower for events that occur in the refueling mode with irradiated fuel 

in the RPV (note that the heatup threat could be lower for cold shutdown conditions if the entry into 

cold shutdown was following a refueling). The above forms the basis for needing both a cold 

shutdown specific IC (CS1) and a refueling specific IC (CS2).
Revision 01/2003 5-C-23

In the cold shutdown mode, normal RCS level and reactor vessel level indication systems (RVLIS) 

will normally be available. However, if all level indication were to be lost during a loss of RCS 

inventory event, the operators would need to determine that RPV inventory loss was occurring by 

observing sump and tank level changes. Sump and tank level increases must be evaluated against 

other potential sources of leakage such as cooling water sources inside the containment to ensure 

they are indicative of RCS leakage.

These example EALs are based on 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. A

number of variables, (BWRs - e.g., such as initial vessel level, or shutdown heat removal system 

design) (PWRs - e.g., mid-loop, reduced level/flange level, head in place, or cavity flooded, RCS 

venting strategy, decay heat removal system design, vortexing pre-disposition, steam generator Utube draining) can have a significant impact on heat removal capability challenging the fuel clad 

barrier. Analysis in the above references indicates that core damage may occur within an hour 

following continued core uncovery therefore, conservatively, 30-minutes was chosen. 

If a PWRs RVLIS is unable to distinguish 6” below the bottom ID of the RCS loop penetration, then 

the first observable point below the bottom ID of the loop should be chosen as the setpoint. If a 

RVLIS is not available such that the PWR EAL setpoint cannot be determined, then EAL 1.b should 

be used to determine if the IC has been met.

The 30-minute duration allowed when CONTAINMENT CLOSURE is established allows sufficient 

time for actions to be performed to recover needed cooling equipment and is considered to be 

conservative given that level is being monitored via CS1 and CS2. For PWRs the effluent release is 

not expected with closure established. For BWRs releases would be monitored and escalation 

would be via Category A ICs if required. 

Thus, for both PWR and BWR declaration of a Site Area Emergency is warranted under the 

conditions specified by the IC. Escalation to a General Emergency is via CG1 (Loss of RPV

Inventory Affecting Fuel Clad Integrity with Containment Challenged with Irradiated Fuel in the RPV) 

or radiological effluent IC AG1 (Offsite Dose Resulting from an Actual or Imminent Release of 

Gaseous Radioactivity Exceeds 1000 mR TEDE or 5000 mR Thyroid CDE for the Actual or 

Projected Duration of the Release Using Actual Meteorology).

Expanded basis for these assumptions is provided in Appendix C.
Revision 01/2003 5-C-24

SYSTEM MALFUNCTION

CS2

Initiating Condition -- SITE AREA EMERGENCY

Loss of RPV Inventory Affecting Core Decay Heat Removal Capability with Irradiated 

Fuel in the RPV.

Operating Mode Applicability: Refueling 

Example Emergency Action Levels: (1 or 2)

1. With CONTAINMENT CLOSURE not established: 

a. RPV inventory as indicated by RPV level less than {site-specific level}

(6" below the low-low ECCS actuation setpoint) (BWR) 

(6" below the bottom ID of the RCS loop) (PWR)

OR

b. RPV level cannot be monitored with Indication of core uncovery as evidenced by one or 

more of the following:

• Containment High Range Radiation Monitor reading > {site-specific} setpoint 

• Erratic Source Range Monitor Indication

• Other {site-specific} indications

2. With CONTAINMENT CLOSURE established 

a. RPV inventory as indicated by RPV level less than TOAF 

OR

b. RPV level cannot be monitored with Indication of core uncovery as evidenced by one or 

more of the following:

• Containment High Range Radiation Monitor reading > {site-specific} setpoint 

• Erratic Source Range Monitor Indication

• Other {site-specific} indications

Basis:

Under the conditions specified by this IC, continued decrease in RPV level is indicative of a loss of 

inventory control. Inventory loss may be due to an RPV breach or continued boiling in the RPV. 

Since BWRs have RCS penetrations below the setpoint, continued level decrease may be

indicative of pressure boundary leakage.

In cold shutdown the decay heat available to raise RCS temperature during a loss of inventory or 

heat removal event may be significantly greater than in the refueling mode Entry into cold shutdown 

conditions may be attained within hours of operating at power or hours after refueling is completed. 

Entry into the refueling mode procedurally may not occur for typically 100 hours {site-specific} or 

longer after the reactor has been shutdown. Thus the heatup threat and therefore the threat to 
Revision 01/2003 5-C-25

damaging the fuel clad may be lower for events that occur in the refueling mode with irradiated fuel 

in the RPV (note that the heatup threat could be lower for cold shutdown conditions if the entry into 

cold shutdown was following a refueling). The above forms the basis for needing both a cold 

shutdown specific IC (CS1) and a refueling specific IC (CS2).

These example EALs are based on 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. A

number of variables, (BWRs - e.g., such as initial vessel level, or shutdown heat removal system 

design) (PWRs - e.g., mid-loop, reduced level/flange level, head in place, or cavity flooded, RCS 

venting strategy, decay heat removal system design, vortexing pre-disposition, steam generator Utube draining) can have a significant impact on heat removal capability challenging the fuel clad 

barrier. Analysis in the above references indicates that core damage may occur within an hour 

following continued core uncovery therefore, conservatively, 30 minutes was chosen. 

If a PWRs RVLIS is unable to distinguish 6” below the bottom ID of the RCS loop penetration, then 

the first observable point below the bottom ID of the loop should be chosen as the setpoint. If a 

RVLIS is not available such that the PWR EAL setpoint cannot be determined, then EAL 1.b should 

be used to determine if the IC has been met.

As water level in the RPV lowers, the dose rate above the core will increase. The dose rate due to 

this core shine should result in up-scaled Containment High Range Monitor indication and possible 

alarm. EAL 1.b and EAL 2.b calculations should be performed to conservatively estimate a sitespecific dose rate setpoint indicative of core uncovery (ie., level at TOAF). Additionally, post-TMI 

studies indicated that the installed 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 2 in the refueling mode, normal means of RPV level indication may not be available. 

Redundant means of RPV level indication will be normally installed (including the ability to monitor 

level visually) to assure that the ability to monitor level will not be interrupted. 

For PWRs the effluent release is not expected with closure established. For BWRs releases would 

be monitored and escalation would be via Category A ICs if required.

Thus, for both PWR and BWR declaration of a Site Area Emergency is warranted under the 

conditions specified by the IC. Escalation to a General Emergency is via CG1 (Loss of RPV

Inventory Affecting Fuel Clad Integrity with Containment Challenged with Irradiated Fuel in the RPV) 

or radiological effluent IC AG1 (Offsite Dose Resulting from an Actual or Imminent Release of 

Gaseous Radioactivity Exceeds 1000 mR TEDE or 5000 mR Thyroid CDE for the Actual or 

Projected Duration of the Release Using Actual Meteorology).

Expanded basis for these assumptions is provided in Appendix C.
Revision 01/2003 5-C-26

SYSTEM MALFUNCTION

CG1

Initiating Condition -- GENERAL EMERGENCY

Loss of RPV Inventory Affecting Fuel Clad Integrity with Containment Challenged with 

Irradiated Fuel in the RPV.

Operating Mode Applicability: Cold Shutdown

Refueling

Example Emergency Action Level: (1 and 2 and 3)

1. Loss of RPV inventory as indicated by unexplained {site-specific} sump and tank level 

increase

2. RPV Level:

a. less than TOAF for > 30 minutes

OR

b. cannot be monitored with Indication of core uncovery for > 30 minutes as evidenced by 

one or more of the following:

• Containment High Range Radiation Monitor reading > {site-specific} setpoint 

• Erratic Source Range Monitor Indication

• Other {site-specific} indications

3. {Site specific} indication of CONTAINMENT challenged as indicated by one or more of the 

following:

• Explosive mixture inside containment 

• Pressure above {site specific} value

• CONTAINMENT CLOSURE not established

• Secondary Containment radiation monitors above {site specific} value (BWR only)

Basis:

For EAL 1 in the cold shutdown mode, normal RCS level and RPV level instrumentation systems 

will normally be available. However, if all level indication were to be lost during a loss of RCS 

inventory event, the operators would need to determine that RPV inventory loss was occurring by 

observing sump and tank level changes. Sump and tank level increases must be evaluated against 

other potential sources of leakage such as cooling water sources inside the containment to ensure 

they are indicative of RCS leakage.

For EAL 1 in the refueling mode, normal means of RPV level indication may not be available. 

Redundant means of RPV level indication will be normally installed (including the ability to monitor 

level visually) to assure that the ability to monitor level will not be interrupted. However, if all level 

indication were to be lost during a loss of RCS inventory event, the operators would need to 

determine that RPV inventory loss was occurring by observing sump and tank level changes. 
Revision 01/2003 5-C-27

For both cold shutdown and refueling modes sump and tank level increases must be evaluated 

against other potential sources of leakage such as cooling water sources inside the containment to 

ensure they are indicative of RCS leakage.

EAL 2 represents the inability to restore and maintain RPV level to above the top of active fuel. Fuel 

damage is probable if RPV level cannot be restored, as available decay heat will cause boiling, 

further reducing the RPV level.

These example EALs are based on 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. A

number of variables, (BWRs - e.g., such as initial vessel level, or shutdown heat removal system 

design) (PWRs - e.g., mid-loop, reduced level/flange level, head in place, or cavity flooded, RCS 

venting strategy, decay heat removal system design, vortexing pre-disposition, steam generator Utube draining) can have a significant impact on heat removal capability challenging the fuel clad 

barrier. Analysis in the above references indicates that core damage may occur within an hour 

following continued core uncovery therefore, conservatively, 30 minutes was chosen. 

As water level in the RPV lowers, the dose rate above the core will increase. For most designs the 

dose rate due to this core shine should result in up-scaled Containment High Range Monitor 

indication and possible alarm. Calculations should be performed to conservatively estimate a sitespecific dose rate setpoint indicative of core uncovery (ie...level at TOAF). Additionally, post-TMI 

studies indicated that the installed nuclear instrumentation will operate erratically when the core is 

uncovered and that this should be used as a tool for making such determinations.

The GE is declared on the occurrence of the loss or imminent loss of function of all three barriers. 

Based on the above discussion, RCS barrier failure resulting in core uncovery for 30 minutes or 

more may cause fuel clad failure. With the CONTAINMENT breached or challenged then the 

potential for unmonitored fission product release to the environment is high. This represents a direct 

path for radioactive inventory to be released to the environment. This is consistent with the definition 

of a GE.

In the context of EAL 3, CONTAINMENT CLOSURE is the action taken to secure containment and 

its associated structures, systems, and components as a functional barrier to fission product 

release under existing plant conditions. CONTAINMENT CLOSURE should not be confused with 

refueling containment integrity as defined in technical specifications. Site shutdown contingency 

plans typically provide for re-establishing CONTAINMENT CLOSURE following a loss of heat 

removal or RCS inventory functions. If the closure is re-established prior to exceeding the 

temperature or level thresholds of the RCS Barrier and Fuel Clad Barrier EALs, escalation to GE 

would not occur.

The site-specific pressure at which CONTAINMENT is considered challenged may change based 

on the condition of the CONTAINMENT. If the Unit is in the cold shutdown mode and the

CONTAINMENT is fully intact then the site-specific setpoint should be equivalent to the

CONTAINMENT design pressure. This is consistent with typical owner’s groups Emergency

Response Procedures. If CONTAINMENT CLOSURE is established intentionally by the plant staff 

in preparations for inspection, maintenance, or refueling then the site-specific setpoint should be 

based on the site-specific pressure assumed for CONTAINMENT CLOSURE.

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 site-
Revision 01/2003 5-C-28

specific radiation monitor values should be based on the EOP “maximum safe values” because 

these values are easily recognizable and have an emergency basis.

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 mixture of dissolved gasses in CONTAINMENT. However, 

CONTAINMENT monitoring and/or sampling should be performed to verify this assumption and a 

General Emergency declared if it is determined that an explosive mixture exists.

Expanded basis for these assumptions is provided in Appendix C.
Revision 01/2003 5-D-1

Recognition Category D

Permanently Defueled Station Malfunction

INITIATING CONDITION MATRIX

NOUE ALERT

D-AU1 UNPLANNED release of gaseous or liquid radioactivity to the 

environment > 2 times the Technical Specification Release 

Limit for > 60 Minutes.

Op. Mode: Not Applicable

D-AA1 UNPLANNED release of gaseous or liquid radioactivity to 

the environment > 200 times the Technical Specification 

Release Limit for > 15 Minutes.

Op. Mode: Not Applicable

D-AU2 UNCONTROLLED increase in plant radiation levels.

Op. Mode: Not Applicable

D-AA2 UNCONTROLLED increase in plant radiation levels that 

impedes operations

Op. Mode: Not Applicable

D-SU1 Decrease in Spent Fuel Pool level OR temperature increase 

that is not the result of a planned evolution.

Op. Mode: Not Applicable

D-HU1 Confirmed security event with potential loss of level of safety of 

the plant.

Op. Mode: Not Applicable

D-HA1 Confirmed security event in the Fuel Building or Control 

Room

Op. Mode: Not Applicable

D-HU2 Other conditions judged warranting declaration of an 

UNUSUAL EVENT.

Op. Mode: Not Applicable

D-HA2 Other conditions judged warranting declaration of ALERT.

Op. Mode: Not Applicable

D-HU3 Natural OR destructive phenomena inside the Protected Area 

affecting the ability to maintain spent fuel integrity.

Op. Mode: Not Applicable
Revision 01/2003 5-D-2

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Revision 01/2003 5-D-3

PERMANENTLY DEFUELED STATION MALFUNCTION

D-AU1

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

UNPLANNED release of gaseous or liquid radioactivity to the environment > 2 times the 

Technical Specification Release Limit for > 60 Minutes.

Operating Mode Applicability: Not Applicable

Example Emergency Action Levels: (1 or 2)

1. UNPLANNED VALID reading on any effluent monitor that exceeds two times the Technical 

Specification Release Limit for > 60 Minutes.

2. Grab sample results indicate UNPLANNED gaseous release rates or liquid concentrations > 2 

times the Technical Specification Release Limit for > 60 Minutes.

Basis:

An UNPLANNED release that cannot be terminated in 60 minutes represents an uncontrolled 

situation that is a potential degradation of the level of safety of the plant. The degradation in plant 

control implied by the fact that the release can not be terminated in 60 minutes is the primary 

concern. The Emergency Director should not wait until 60 minutes has elapsed, but should declare 

an UNUSUAL EVENT as soon as the release is determined to be uncontrolled or projected to be 

unisolable within 60 minutes. 

The EAL 1 limit ensures compliance with 10CFR20.1301 dose limits to the public. This limit also 

ensures the concentration of liquid effluents released is < 2 times the value specified in 10CFR20, 

Appendix B. 

The EAL 2 grab samples are used to determine gaseous release rates or liquid concentrations to 

confirm monitor readings or when the effluent monitors are not in service.
Revision 01/2003 5-D-4

PERMANENTLY DEFUELED STATION MALFUNCTION

D-AU2

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

UNCONTROLLED increase in plant radiation levels.

Operating Mode Applicability: Not Applicable

Example Emergency Action Level: 

1. Area Radiation Monitor readings or survey results indicate an uncontrolled increase in radiation 

level by 25 mR/hr that is not the result of a planned evolution. 

Basis:

UNCONTROLLED means an increase in < 12 hours of monitored radiation level that is not the 

result of a planned evolution and the source of the increased is not immediately recognized and 

controlled. 

Classification of an UNUSUAL EVENT is warranted as a precursor to more serious events. The 

concern of this EAL is the loss of control of radioactive material representing a potential degradation 

of the level of safety of the plant.
Revision 01/2003 5-D-5

PERMANENTLY DEFUELED STATION MALFUNCTION

D-SU1

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

Decrease in Spent Fuel Pool Level OR temperature increase that is not the result of a 

planned evolution.

Operating Mode Applicability: Not Applicable

Example Emergency Action Levels: (1 or 2)

1. a. VALID (site-specific) indication of uncontrolled water level decrease in spent fuel pool with 

all irradiated fuel assemblies remaining covered by water.

AND

b. UNPLANNED VALID (site-specific) Direct Area Radiation Monitor reading increases 

2. Spent Fuel Pool temperature increase to > [site-specific] 0F that is not the result of a planned 

evolution.

Basis:

Classification of an NOUE for the EAL threshold value is warranted as a precursor to more serious 

events and a potential degradation in the level of safety of the plant. Since loss of level or continued 

pool boiling would result in increased radiation levels exceeding the criteria of D-AA2, continued 

system related loss of level type events are bounded by D-AA2.

The EAL1 site-specific value for level should be based on a calculated level that will result in 

prohibitive radiation levels in the Fuel Building. The site-specific radiation monitors should be 

chosen so that indication of decreasing pool levels is provided.

The EAL2 site-specific temperature should be chosen based on the initial temperature starting point 

for fuel damage calculations (typically 125 to 150¢F) in the Safety Analysis Report (SAR). 
Revision 01/2003 5-D-6

PERMANENTLY DEFUELED STATION MALFUNCTION

D-HU1

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

Confirmed Security Event with potential loss of level of safety of the plant.

Operating Mode Applicability: Not Applicable

Example Emergency Action Levels: 

1. Security Event as determined from (site-specific) Safeguards Contingency Plan and reported 

by the (site-specific) security shift supervision.

Basis:

This EAL is based on (site-specific) Site Security Plans. Security events which do not represent a 

potential degradation in the level of safety of the plant, are reported under 10 CFR 73.71 or in some 

cases under 10 CFR 50.72. 

INTRUSION into the Fuel Building or Control Room by a HOSTILE FORCE would result in EAL 

escalation to an ALERT.

Reference is made to (site-specific) security shift supervision because these individuals are the 

designated personnel on-site qualified and trained to confirm that a security event is occurring or 

has occurred. Training on security event classification confirmation is closely controlled due to the 

strict secrecy controls placed on the plant Security Plan.
Revision 01/2003 5-D-7

PERMANENTLY DEFUELED STATION MALFUNCTION

D-HU2

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

Other conditions judged warranting declaration of an UNUSUAL EVENT

Operating Mode Applicability: Not Applicable

Example Emergency Action Levels: 

1. Other conditions exist which in the judgment of the Shift Supervisor /Emergency Director 

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

Basis:

Any condition not explicitly detailed as an EAL threshold value, which, in the judgment of the 

Emergency Director, is a potential degradation in the level of safety of the plant. Emergency 

Director judgment is to be based on known conditions and the expected response to mitigating 

activities within a short time period.
Revision 01/2003 5-D-8

PERMANENTLY DEFUELED STATION MALFUNCTION

D-HU3

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

Natural or destructive phenomena inside the PROTECTED AREA affecting the ability to 

maintain spent fuel integrity

Operating Mode Applicability: Not Applicable

Example Emergency Action Levels: (1 or 2 or 3 or 4 or 5 or 6 or 7 or 8)

1. (Site-Specific) method indicates felt earthquake.

2. Report by plant personnel of tornado or high winds greater than (site-specific) mph striking 

within the PROTECTED AREA that has the potential to affect equipment needed to maintain 

spent fuel integrity.

3. Vehicle crash into plant structures or systems within PROTECTED AREA boundary that has 

the potential to affect equipment needed to maintain spent fuel integrity.

4. Report by plant personnel of an unanticipated EXPLOSION within PROTECTED AREA

boundary resulting in VISIBLE DAMAGE that has the potential to affect equipment needed to 

maintain spent fuel integrity.

5. Uncontrolled flooding in (site-specific) areas of the plant that has the potential to affect

equipment needed to maintain spent fuel integrity.

6. FIRE in the following (Site-Specific) buildings or areas not extinguished within 15 minutes of 

Control Room notification or verification of a control room alarm that has the potential to affect 

equipment needed to maintain spent fuel integrity.

7. Toxic or flammable gas within the PROTECTED AREA that has the potential to affect the 

operation of equipment needed to maintain spent fuel integrity.

8. (Site-Specific) occurrences affecting the PROTECTED AREA that has the potential to affect 

equipment needed to maintain spent fuel integrity.

Basis:

NOUE in this IC are categorized on the basis of the occurrence of an event of sufficient magnitude 

to be of concern to plant operators. Areas identified in the EALs define the location of the event 

based on the potential for damage to equipment contained therein. 

EAL #1 should be developed on site-specific basis. Damage may be caused to some portions of 

the site, but should not affect ability to operate spent fuel pool equipment. Method of detection can 

be based on instrumentation, validated by a reliable source, or operator assessment. As defined in 

the EPRI-sponsored "Guidelines for Nuclear Plant Response to an Earthquake", dated October 

1989, a "felt earthquake" is:
Revision 01/2003 5-D-9

An earthquake of sufficient intensity such that: (a) the vibratory ground motion is felt at the 

nuclear plant site and recognized as an earthquake based on a consensus of control room 

operators on duty at the time, and (b) for plants with operable seismic instrumentation, the 

seismic switches of the plant are activated. For most plants with seismic instrumentation, the 

seismic switches are set at an acceleration of about 0.01g.

EAL #2 is based on the assumption that a tornado striking (touching down) or high winds within the 

protected area may have potentially damaged plant structures containing functions or systems 

required to maintain spent fuel integrity. The high wind site specific value in EAL#2 should be based 

on site-specific FSAR design basis. 

EAL #3 is intended to address crashes of vehicles that cause significant damage to plant structures 

containing functions and systems necessary to maintain spent fuel integrity.

EAL #4 addresses only those EXPLOSIONs of sufficient force to damage equipment needed to 

maintain spent fuel integrity. No attempt is made in this EAL to assess the actual magnitude of the 

damage. The occurrence of the EXPLOSION with reports of evidence of damage is sufficient for 

declaration. The Emergency Director also needs to consider any security aspects of the

EXPLOSION, if applicable. 

EAL #5 addresses the effect of flooding caused by internal events such as component failures or 

equipment misalignment that has the potential to affect equipment needed to maintain spent fuel 

integrity. The site-specific areas include those areas that contain systems required to maintain fuel 

integrity, that are not designed to be wetted or submerged. 

EAL #6 addresses FIREs that may have the potential to affect the ability to maintain spent fuel 

integrity. As used here, Detection is visual observation and report by plant personnel or sensor 

alarm indication. The 15 minute time period begins within a credible notification that a FIRE is 

occurring, or indication of a VALID fire detection system alarm. Verification of a fire detection 

system alarm includes actions that can be taken with the control room or other nearby site-specific 

location to ensure that the alarm is not spurious. A verified alarm is assumed to be an indication of 

a FIRE unless it is disproved within the 15 minute period by personnel dispatched to the scene. In 

other words, a personnel report from the scene may be used to disprove a sensor alarm if received 

within 15 minutes of the alarm, but shall not be required to verify the alarm. 

The intent of this 15 minute duration is to size the FIRE and to discriminate against small FIREs 

that are readily extinguished (e.g., smoldering waste paper basket). The site-specific list should be 

limited and applies to buildings and areas containing equipment important to maintaining spent fuel 

integrity. This excludes FIREs within administration buildings, waste-basket FIREs, and other small 

FIREs of no safety consequence.

EAL #7 addresses toxic or flammable gas in the protected area that has the potential to affect the 

ability to maintain spent fuel integrity due to the potential damage to equipment or the evacuation of 

personnel preventing operation or maintenance of spent fuel pool equipment.

EAL #8 covers other site-specific phenomena such as hurricane, flood, or seiche that have the 

potential to result loss of spent fuel integrity. 

Escalation to the ALERT level will be via D-AA2 if any of the above events has caused damage that 

results in radiation levels increasing by 100 mr/hr and impedes operation of systems needed to 

maintain spent fuel integrity.
Revision 01/2003 5-D-10

PERMANENTLY DEFUELED STATION MALFUNCTION

D-AA1

Initiating Condition -- ALERT

UNPLANNED release of gaseous or liquid radioactivity to the environment > 200 times 

the Technical Specification Release Limit for > 15 Minutes.

Operating Mode Applicability: Not Applicable

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

1. UNPLANNED VALID reading on any effluent monitor that exceeds 200 times the Technical 

Specification Release Limit for > 15 Minutes.

2. Confirmed sample analyses for gaseous or liquid releases indicates concentrations or

release rates, with a duration of 15 minutes or longer, in excess of 200 times (site –specific 

Technical Specifications.

Basis:

An UNPLANNED release of this magnitude that cannot be terminated in 15 minutes represents an 

uncontrolled situation that is an actual or potential substantial degradation of the level of safety of 

the plant. The degradation in plant control implied by the fact that the release can not be terminated 

in 15 minutes is the primary concern. The Emergency Director should not wait until 15 minutes has 

elapsed, but should declare an ALERT as soon as the release is determined to be uncontrolled or 

projected to be unisolable within 15 minutes. 

The EAL1 release rate limit ensures compliance with 10CFR20.1301 dose limits to the public. This 

limit also ensures the concentration of liquid effluents is < 200 times the value specified in

10CFR20, Appendix B. 

The EAL2 grab samples are used to determine gaseous release rates or liquid concentrations to 

confirm monitor readings or when the effluent monitors are not in service.
Revision 01/2003 5-D-11

PERMANENTLY DEFUELED STATION MALFUNCTION

D-AA2

Initiating Condition -- ALERT

UNCONTROLLED increase in plant radiation levels that impede operations

Operating Mode Applicability: Not Applicable

Example Emergency Action Levels: (1 or 2)

1. Area Radiation Monitor readings or survey results indicate an UNCONTROLLED increase in 

radiation level by 100 mR/hr that is not the result of a planned evolution and impedes access 

to areas needed to maintain control of radioactive material or operation of systems needed to 

maintain spent fuel integrity.

(Site-specific) list

2. VALID (site-specific) radiation monitor readings GREATER THAN 15 mR/hr in areas requiring 

continuous occupancy:

(Site-specific) list

Basis:

The site specific list for EAL1 will include available Fuel Handling building radiation monitors.

An increase in radiation levels that is not the result of a planned evolution that impedes operations 

necessary to allow maintenance of spent fuel integrity warrants the classification of an ALERT.

Damage to spent fuel represents a substantial degradation in the level of safety of the plant and 

therefore warrants an ALERT classification. 

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.
Revision 01/2003 5-D-12

PERMANENTLY DEFUELED STATION MALFUNCTION

D-HA1

Initiating Condition -- ALERT

Confirmed Security Event in the Fuel Building or Control Room.

Operating Mode Applicability: Not Applicable

Example Emergency Action Levels: 

1. INTRUSION into the Fuel Building or Control Room by a HOSTILE FORCE.

Basis:

This class of security events represents an escalated threat to plant safety above that contained in 

the NOUE. A confirmed INTRUSION report is satisfied if physical evidence indicates the presence 

of a HOSTILE FORCE within the Fuel Handling Building or Control Room. 
Revision 01/2003 5-D-13

PERMANENTLY DEFUELED STATION MALFUNCTION

D-HA2

Initiating Condition -- ALERT

Other conditions judged warranting declaration of ALERT.

Operating Mode Applicability: Not Applicable

Example Emergency Action Levels: 

1. Other conditions exist which in the judgment of the Emergency Director indicate that plant 

systems may be substantially degraded and that increased monitoring of plant functions is 

warranted. Any releases are expected to be limited to small fractions of the EPA Protective 

Action Guideline exposure levels.

Basis:

A condition exists which, in the judgement of the Emergency Director, presents an actual or 

potential substantial degradation in the level of safety of the plant. Emergency Director judgement 

is to be based on known conditions and the expected response to mitigating activities.
Revision 01/2003 5-E-1

Recognition Category E

Events Related to ISFSI Malfunction

INITIATING CONDITION MATRIX

NOUE

E-HU1 Damage to a loaded cask CONFINEMENT BOUNDARY.

Op. Mode: Not Applicable

E-HU2 Confirmed security event with potential loss of level of safety of the ISFSI

Op. Mode: Not Applicable
Revision 01/2003 5-E-2

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Revision 01/2003 5-E-3

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Revision 01/2003 5-E-4

EVENTS RELATED TO ISFSI

E-HU1

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

Damage to a loaded cask CONFINEMENT BOUNDARY.

Operating Mode Applicability: Not applicable

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

1. Natural phenomena events affecting a loaded cask CONFINEMENT BOUNDARY.

(site-specific list)

2. Accident conditions affecting a loaded cask CONFINEMENT BOUNDARY.

(site-specific list)

3. Any condition in the opinion of the Emergency Director that indicates loss of loaded fuel 

storage cask CONFINEMENT BOUNDARY.

Basis:

A NOUE in this IC is categorized on the basis of the occurrence of an event of sufficient magnitude 

that a loaded cask CONFINEMENT BOUNDARY is damaged or violated. This includes

classification based on a loaded fuel storage cask CONFINEMENT BOUNDARY loss leading to the 

degradation of the fuel during storage or posing an operational safety problem with respect to its 

removal from storage.

For EAL #1 and EAL #2, the results of the ISFSI Safety Analysis Report (SAR) per NUREG 1536 or 

SAR referenced in the cask(‘s) Certificate of Compliance and the related NRC Safety Evaluation 

Report should be used to develop the site-specific list of natural phenomena events and accident 

conditions. These EALs would address responses to a dropped cask, a tipped over cask,

explosion, missile damage, fire damage or natural phenomena affecting a cask (e.g., seismic 

event, tornado, etc.).

For EAL #3, any condition not explicitly detailed as an EAL threshold value, which, in the judgment 

of the Emergency Director, is a potential degradation in the level of safety of the ISFSI. Emergency 

Director judgment is to be based on known conditions and the expected response to mitigating 

activities within a short time period.
Revision 01/2003 5-E-5

EVENTS RELATED TO ISFSI

E-HU2

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

Confirmed Security Event with potential loss of level of safety of the ISFSI.

Operating Mode Applicability: Not applicable

Example Emergency Action Levels: 

1. Security Event as determined from (site-specific) Security Plan and reported by the (sitespecific) security shift supervision.

Basis:

This EAL is based on (site-specific) Security Plans. Security events which do not represent a 

potential degradation in the level of safety of the ISFSI, are reported under 10 CFR 73.71 or in some 

cases under 10 CFR 50.72. 

Reference is made to (site-specific) security shift supervision because these individuals are the 

designated personnel qualified and trained to confirm that a security event is occurring or has 

occurred. Training on security event classification confirmation is closely controlled due to the strict 

secrecy controls placed on the Security Plan.
Revision 01/2003 5-F-1

Table 5-F-1

Recognition Category F

Fission Product Barrier Degradation

INITIATING CONDITION MATRIX

See Table 3 for BWR Example EALs

See Table 4 for PWR Example EALs

NOUE ALERT SITE AREA EMERGENCY GENERAL EMERGENCY

FU1 ANY Loss or ANY Potential Loss 

of Containment

Op. Modes: Power Operation, 

Hot Standby, Startup, Hot 

Shutdown

FA1 ANY Loss or ANY Potential Loss 

of EITHER Fuel Clad OR RCS

Op. Modes: Power Operation, 

Hot Standby, Startup, Hot 

Shutdown

FS1 Loss or Potential Loss of ANY 

Two Barriers

Op. Modes: Power Operation, 

Hot Standby, Startup, Hot 

Shutdown

FG1 Loss of ANY Two Barriers AND 

Loss or Potential Loss of Third 

Barrier

Op. Modes: Power Operation, 

Hot Standby, Startup, Hot 

Shutdown

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 (See Sections 3.4 and 3.8). NOUE ICs associated 

with RCS and Fuel Clad Barriers are addressed under System Malfunction ICs.

• At the Site Area Emergency level, there must be some ability to dynamically assess how far present conditions are from the threshold for a General 

Emergency. For example, if Fuel Clad and RCS Barrier “Loss” EALs existed, that, in addition to offsite dose assessments, would require continual 

assessments of radioactive inventory and containment integrity. Alternatively, if both Fuel Clad and RCS Barrier “Potential Loss” EALs existed, the 

Emergency Director would have more assurance that there was no immediate need to escalate to a General Emergency.

• The ability to escalate to higher emergency classes as an event deteriorates must be maintained. For example, RCS leakage steadily increasing would 

represent an increasing risk to public health and safety.

2. Fission Product Barrier ICs must be capable of addressing event dynamics. Thus, the EAL Reference Table 3 and 4 state that imminent (i.e., within 2 

hours) Loss or Potential Loss should result in a classification as if the affected threshold(s) are already exceeded, particularly for the higher emergency 

classes.
TABLE 5-F-2

BWR Emergency Action Level

Fission Product Barrier Reference Table

Thresholds For LOSS or POTENTIAL LOSS of Barriers*

*Determine which combination of the three barriers are lost or have a potential loss and use the following key to classify the event. Also, multiple events could occur which result in the 

conclusion that exceeding the loss or Potential loss thresholds is imminent (i.e., within 1 to 2 hours). In this imminent loss situation use judgment and classify as if the thresholds are exceeded.

UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY

ANY loss or ANY Potential Loss of 

Containment

ANY loss or ANY Potential Loss of EITHER 

Fuel Clad or RCS

Loss or Potential Loss of ANY two Barriers Loss of ANY two Barriers AND

Loss or Potential Loss of Third Barrier

Fuel Clad Barrier Example EALS RCS Barrier Example EALS Containment Barrier Example EALS

LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS

Revision 01/2003 5-F-2

1. Primary Coolant Activity Level 1. Drywell Pressure 1. Drywell Pressure

Coolant Activity GREATER 

THAN (site-specific) Value

Not Applicable Pressure GREATER THAN 

(site-specific) PSIG

Not Applicable Rapid unexplained 

decrease following initial 

increase

OR

Drywell pressure 

response not consistent 

with LOCA conditions

(Site-specific) PSIG and 

increasing

OR

Explosive mixture exists

OR OR OR

2. Reactor Vessel Water Level 2. Reactor Vessel Water Level 2. Reactor Vessel Water Level

Level LESS THAN (sitespecific value)Level LESS THAN (sitespecific value)Level LESS THAN (sitespecific value)Not Applicable Not Applicable Primary containment 

flooding required

OR OR

3. RCS Leak Rate 3. CNMT Isolation Failure or Bypass
TABLE 5-F-2

BWR Emergency Action Level

Fission Product Barrier Reference Table

Thresholds For LOSS or POTENTIAL LOSS of Barriers*

*Determine which combination of the three barriers are lost or have a potential loss and use the following key to classify the event. Also, multiple events could occur which result in the 

conclusion that exceeding the loss or Potential loss thresholds is imminent (i.e., within 1 to 2 hours). In this imminent loss situation use judgment and classify as if the thresholds are exceeded.

UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY

ANY loss or ANY Potential Loss of 

Containment

ANY loss or ANY Potential Loss of EITHER 

Fuel Clad or RCS

Loss or Potential Loss of ANY two Barriers Loss of ANY two Barriers AND

Loss or Potential Loss of Third Barrier

Fuel Clad Barrier Example EALS RCS Barrier Example EALS Containment Barrier Example EALS

LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS

Revision 01/2003 5-F-3

(Site-specific) Indication of 

an unisolable Main 

Steamline Break

RCS leakage GREATER 

THAN 50 gpm inside the 

drywell

OR

Unisolable primary system 

leakage outside drywell as 

indicated by area 

temperature or area 

radiation alarm

Failure of both valves in 

any one line to close AND 

downstream pathway to 

the environment exists

OR

Intentional venting per 

EOPs

OR

Unisolable primary system 

leakage outside drywell as 

indicated by area 

temperature or area 

radiation alarm

Not applicable

OR OR OR
TABLE 5-F-2

BWR Emergency Action Level

Fission Product Barrier Reference Table

Thresholds For LOSS or POTENTIAL LOSS of Barriers*

*Determine which combination of the three barriers are lost or have a potential loss and use the following key to classify the event. Also, multiple events could occur which result in the 

conclusion that exceeding the loss or Potential loss thresholds is imminent (i.e., within 1 to 2 hours). In this imminent loss situation use judgment and classify as if the thresholds are exceeded.

UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY

ANY loss or ANY Potential Loss of 

Containment

ANY loss or ANY Potential Loss of EITHER 

Fuel Clad or RCS

Loss or Potential Loss of ANY two Barriers Loss of ANY two Barriers AND

Loss or Potential Loss of Third Barrier

Fuel Clad Barrier Example EALS RCS Barrier Example EALS Containment Barrier Example EALS

LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS

Revision 01/2003 5-F-4

3. Drywell Radiation Monitoring 4. Drywell Radiation Monitoring 4. Significant Radioactive Inventory in Containment

Drywell Radiation monitor 

reading GREATER THAN 

(site-specific) R/hr

Not Applicable Drywell Radiation monitor 

reading GREATER THAN 

(site-specific) R/hr

Not Applicable Not applicable Drywell Radiation monitor 

reading GREATER THAN 

(site-specific) R/hr

OR OR OR

4. Other (Site-Specific) Indications 5. Other (Site-Specific) Indications 5. Other (site-specific) Indications

(Site specific ) as 

applicable

(Site specific) as 

applicable

(Site-specific) as 

applicable

(Site-specific) as 

applicable

(Site specific) as 

applicable

(Site specific) as 

applicable

OR OR OR

5. Emergency Director Judgment 6. Emergency Director Judgment 6. Emergency Director Judgment

Any condition in the opinion of the Emergency Director that

indicates Loss or Potential Loss of the Fuel Clad Barrier

Any condition in the opinion of the Emergency Director that 

indicates Loss or Potential Loss of the RCS Barrier

Any condition in the opinion of the Emergency Director

that indicates Loss or Potential Loss of the Containment 

barrier
Revision 01/2003 5-F-5

Basis Information For Table 5-F-2

BWR Emergency Action Level

Fission Product Barrier Reference Table

FUEL CLAD BARRIER EXAMPLE EALs: (1 or 2 or 3 or 4 or 5)

The Fuel Clad barrier is the zircalloy or stainless steel tubes that contain the fuel pellets.

1. Primary Coolant Activity Level

This (site-specific) value corresponds to 300 μCi/gm I131 equivalent. Assessment by the NUMARC 

EAL Task Force indicates that this amount of coolant activity is well above that expected for iodine 

spikes and corresponds to less than 5% fuel clad damage. This amount of radioactivity indicates 

significant clad damage and thus the Fuel Clad Barrier is considered lost. The value expressed can 

be either in mR/hr observed on the sample or as uCi/gm results from analysis. 

There is no equivalent "Potential Loss" EAL for this item.

2. Reactor Vessel Water Level

The "Loss" EAL (site-specific) value corresponds to the level which is used in EOPs to indicate 

challenge of core cooling. Depending on the plant this may be top of active fuel or 2/3 coverage of 

active fuel. This is the minimum value to assure core cooling without further degradation of the clad. 

The "Potential Loss" EAL is the same as the RCS barrier "Loss" EAL #2 below and corresponds to 

the (site-specific) water level at the top of the active fuel. Thus, this EAL indicates a "Loss" of RCS 

barrier and a "Potential Loss" of the Fuel Clad Barrier. This EAL appropriately escalates the

emergency class to a Site Area Emergency. If the "Loss" value is also the Top of Active Fuel, the 

"Potential Loss" value must be a value indicating a higher level also corresponding to a higher level 

indicated in the RCS barrier "Loss" EAL #2.

3. Drywell Radiation Monitoring

The (site-specific) reading is a value which indicates the release of reactor coolant, with elevated 

activity indicative of fuel damage, into the drywell. The reading should be calculated assuming the 

instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory

associated with a concentration of 300 μCi/gm dose equivalent I-131 or the calculated

concentration equivalent to the clad damage used in EAL #1 into the drywell atmosphere. Reactor 

coolant concentrations of this magnitude are several times larger than the maximum

concentrations (including iodine spiking) allowed within technical specifications and are therefore 

indicative of fuel damage. This value is higher than that specified for RCS barrier Loss EAL #4. 

Thus, this EAL indicates a loss of both Fuel Clad barrier and RCS barrier.

Caution: it is important to recognize that in the event the radiation monitor is sensitive to shine from 

the reactor vessel or piping, spurious readings will be present and another indicator of fuel clad 

damage is necessary or compensated for in the threshold value.

There is no "Potential Loss" EAL associated with this item. 
Revision 01/2003 5-F-6

4. Other (Site-Specific) Indications

This EAL is to cover other (site-specific) indications that may indicate loss or potential loss of the 

Fuel Clad barrier, including indications from containment air monitors or any other (site-specific) 

instrumentation.

5. Emergency Director Judgment

This EAL addresses any other factors that are to be used by the Emergency Director in determining 

whether the Fuel Clad barrier is lost or potentially lost. In addition, the inability to monitor the barrier 

should also be incorporated in this EAL as a factor in Emergency Director judgment that the barrier 

may be considered lost or potentially lost. (See also IC SG1, "Prolonged Loss of All Offsite Power 

and Prolonged Loss of All Onsite AC Power", for additional information.)

RCS BARRIER EXAMPLE EALs: (1 or 2 or 3 or 4 or 5 or 6)

The RCS Barrier is the reactor coolant system pressure boundary and includes the reactor vessel 

and all reactor coolant system piping up to the isolation valves.

1. Drywell Pressure

The (site-specific) drywell pressure is based on the drywell high pressure set point which indicates 

a LOCA by automatically initiating the ECCS or equivalent makeup system. 

There is no "Potential Loss" EAL corresponding to this item.

2. Reactor Vessel Water Level

This "Loss" EAL is the same as "Potential Loss" Fuel Clad Barrier EAL #2. The (site-specific) water 

level corresponds to the level which is used in EOPs to indicate challenge of core cooling.

Depending on the plant this may be top of active fuel or 2/3 coverage of active fuel. This EAL 

appropriately escalates the emergency class to a Site Area Emergency. Thus, this EAL indicates a 

loss of the RCS barrier and a Potential Loss of the Fuel Clad Barrier.

There is no "Potential Loss" EAL corresponding to this item.

3. RCS Leak Rate

An unisolable MSL break is a breach of the RCS barrier. Thus, this EAL is included for consistency 

with the Alert emergency classification. The potential loss of RCS based on leakage is set at a level 

indicative of a small breach of the RCS but which is well within the makeup capability of normal and 

emergency high pressure systems. Core uncovery is not a significant concern for a 50 gpm leak, 

however, break propagation leading to significantly larger loss of inventory is possible. Many BWRs 

may be unable to measure an RCS leak of this size because the leak would likely increase drywell 

pressure above the drywell isolation set point. The system normally used to monitor leakage is 

typically isolated as part of the drywell isolation and is therefore unavailable. If primary system leak 

rate information is unavailable, other indicators of RCS leakage should be used. 

Potential loss of RCS based on primary system leakage outside the drywell is determined from 

site-specific temperature or area radiation alarms low setpoint in the areas of the main steam line 

tunnel, main turbine generator, RCIC, HPCI, etc., which indicate a direct path from the RCS to 

areas outside primary containment. The indicators should be confirmed to be caused by RCS 

leakage. The area temperature or radiation low alarm setpoints are indicated for this example to 
Revision 01/2003 5-F-7

enable an Alert classification. An unisolable leak which is indicated by a high alarm setpoint 

escalates to a Site Area Emergency when combined with Containment Barrier EAL 3 (after a 

containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also

exceeded.

4. Drywell Radiation Monitoring

The (site-specific) reading is a value which indicates the release of reactor coolant to the drywell. 

The reading should be calculated assuming the instantaneous release and dispersal of the reactor 

coolant noble gas and iodine inventory associated with normal operating concentrations (i.e., within 

T/S) into the drywell atmosphere. This reading will be less than that specified for Fuel Clad Barrier 

EAL #3. Thus, this EAL would be indicative of a RCS leak only. If the radiation monitor reading 

increased to that value specified by Fuel Clad Barrier EAL #3, fuel damage would also be indicated.

However, if the site specific physical location of the drywell radiation monitor is such that radiation 

from a cloud of released RCS gases could not be distinguished from radiation from adjacent piping 

and components containing elevated reactor coolant activity, this EAL should be omitted and other 

site specific indications of RCS leakage substituted.

There is no "Potential Loss" EAL associated with this item.

5. Other (Site-Specific) Indications

This EAL is to cover other (site-specific) indications that may indicate loss or potential loss of the 

RCS barrier.

6. Emergency Director Judgment

This EAL addresses any other factors that are to be used by the Emergency Director in determining 

whether the RCS barrier is lost or potentially lost. In addition, the inability to monitor the barrier 

should also be incorporated in this EAL as a factor in Emergency Director judgment that the barrier 

may be considered lost or potentially lost. (See also IC SG1, "Prolonged Loss of Offsite Power and 

Prolonged Loss of All Onsite AC Power", for additional information.)

PRIMARY CONTAINMENT BARRIER EXAMPLE EALs: (1 or 2 or 3 or 4 or 5 or 6)

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 EALs are used primarily as discriminators for escalation from an Alert to a Site 

Area Emergency or a General Emergency.

1. Drywell Pressure

Rapid unexplained loss of pressure (i.e., not attributable to drywell spray or condensation effects) 

following an initial pressure increase indicates a loss of containment integrity. Drywell pressure 

should increase as a result of mass and energy release into containment from a LOCA. Thus, 

drywell pressure not increasing under these conditions indicates a loss of containment integrity. 

This indicator relies on the operators recognition of an unexpected response for the condition and 

therefore does not have a specific value associated. The unexpected response is important 

because it is the indicator for a containment bypass condition. The (site-specific) PSIG for potential 

loss of containment is based on the containment drywell design pressure. Existence of an

explosive mixture means a hydrogen and oxygen concentration of at least the lower deflagration 
Revision 01/2003 5-F-8

limit curve exists. This applies to BWRs with Mark III containments, as well as Mark I and II

containment designs when they are de-inerted.

2. Reactor Vessel Water Level

The entry into the Primary Containment Flooding emergency procedure indicates reactor vessel 

water level can not be restored and that a core melt sequence is in progress. EOPs direct the 

operators to enter Containment Flooding when Reactor Vessel Level cannot be restored to greater 

than a Site Specific value (generally 2/3 core height) or is unknown. Entry into Containment 

Flooding procedures is a logical escalation in response to the inability to maintain reactor vessel 

level.

The conditions in this potential loss EAL represent imminent core melt sequences which, if not 

corrected, could lead to vessel failure and increased potential for containment failure. In conjunction 

with and an escalation of the level EALs in the Fuel and RCS barrier columns, this EAL will result in 

the declaration of a General Emergency -- loss of two barriers and the potential loss of a third. If the 

emergency operating procedures have been ineffective in restoring reactor vessel level above the 

RCS and Fuel Clad Barrier Threshold Values, there is not a "success" path and a core melt 

sequence is in progress. Entry into Containment flooding procedures is a logical escalation in 

response to the inability to maintain reactor vessel level.

Severe accident analysis (e.g., NUREG-1150) have concluded that function restoration procedures 

can arrest core degradation with the reactor vessel in a significant fraction of the core damage 

scenarios, and the likelihood of containment failure is very small in these events. Given this, it is 

appropriate to provide a reasonable period to allow emergency operating procedures to arrest the 

core melt sequence. Whether or not the procedures will be effective should be apparent within the 

time provided. The Emergency Director should make the declaration as soon as it is determined 

that the procedures have been, or will be, ineffective. There is no "loss" EAL associated with this 

item.

3. Containment Isolation Failure or Bypass

This EAL is intended to cover the inability to isolate the containment when containment isolation is 

required. In addition, the presence of area radiation or temperature alarms high setpoint indicating 

unisolable primary system leakage outside the drywell are covered after a containment isolation. 

The indicators should be confirmed to be caused by RCS leakage. Also, an intentional venting of 

primary containment for pressure control per EOPs to the secondary containment and/or the 

environment is considered a loss of containment. Containment venting for temperature or pressure 

when not in an accident situation should not be considered. 

There is no "Potential Loss" EAL associated with this item.

4. Significant Radioactive Inventory in Containment

The (site-specific) reading is a value which indicates significant fuel damage well in excess of that 

required for loss of RCS and Fuel Clad. As stated in Section 3.8, a major release of radioactivity 

requiring offsite protective actions from core damage is not possible unless a major failure of fuel 

cladding allows radioactive material to be released from the core into the reactor coolant.

Regardless of whether containment is challenged, this amount of activity in containment, if

released, could have such severe consequences that it is prudent to treat this as a potential loss of 

containment, such that a General Emergency declaration is warranted. NUREG-1228, "Source 

Estimations During Incident Response to Severe Nuclear Power Plant Accidents," indicates that 

such conditions do not exist when the amount of clad damage is less than 20%. Unless there is a 
Revision 01/2003 5-F-9

(site-specific) analysis justifying a higher value, it is recommended that a radiation monitor reading 

corresponding to 20% fuel clad damage be specified here.

There is no "Loss" EAL associated with this item.

5. Other (Site-Specific) Indications

This EAL is to cover other (site-specific) indications that may indicate loss or potential loss of the 

containment barrier.

6. Emergency Director Judgment

This EAL addresses any other factors that are to be used by the Emergency Director in determining 

whether the Containment barrier is lost or potentially lost. In addition, the inability to monitor the 

barrier should also be incorporated in this EAL as a factor in Emergency Director judgment that the 

barrier may be considered lost or potentially lost. (See also IC SG1, "Prolonged Loss of All Offsite 

Power and Prolonged Loss of All Onsite AC Power", for additional information.)
TABLE 5-F-4

PWR Emergency Action Level

Fission Product Barrier Reference Table

Thresholds For LOSS or POTENTIAL LOSS of Barriers*

*Determine which combination of the three barriers are lost or have a potential loss and use the following key to classify the event. Also an event for multiple events could occur which result 

in the conclusion that exceeding the loss or potential loss thresholds is imminent (i.e., within 1 to 2 hours). In this imminent loss situation use judgment and classify as if the thresholds are 

exceeded.

UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY

ANY loss or ANY Potential Loss of 

Containment

ANY loss or ANY Potential Loss of EITHER 

Fuel Clad or RCS

Loss or Potential Loss of ANY two Barriers Loss of ANY two Barriers AND

Loss or Potential Loss of Third Barrier

Fuel Clad Barrier Example EALS RCS Barrier Example EALS Containment Barrier Example EALS

LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS

Revision 01/2003 5-F-10

1. Critical Safety Function Status 1. Critical Safety Function Status 1. Critical Safety Function Status

Core-Cooling Red Core Cooling-Orange OR 

Heat Sink-Red

Not Applicable RCS Integrity-Red OR Heat 

Sink-Red

Not Applicable Containment-Red

OR OR OR

2. Primary Coolant Activity Level 2. RCS Leak Rate 2. Containment Pressure

Coolant Activity GREATER 

THAN (site-specific) Value

Not Applicable GREATER THAN available 

makeup capacity as 

indicated by a loss of RCS 

subcooling

Unisolable leak exceeding 

the capacity of one 

charging pump in the 

normal charging mode

Rapid unexplained 

decrease following initial 

increase

OR

Containment pressure or 

sump level response not 

consistent with LOCA 

conditions

(Site-specific) PSIG and 

increasing 

OR

Explosive mixture exists

OR

Pressure greater than 

containment depressurization actuation setpoint with 

less than one full train of 

depressurization 

equipment operating

OR OR

3. Core Exit Thermocouple Readings 3. Core Exit Themocouple Reading
TABLE 5-F-4

PWR Emergency Action Level

Fission Product Barrier Reference Table

Thresholds For LOSS or POTENTIAL LOSS of Barriers*

*Determine which combination of the three barriers are lost or have a potential loss and use the following key to classify the event. Also an event for multiple events could occur which result 

in the conclusion that exceeding the loss or potential loss thresholds is imminent (i.e., within 1 to 2 hours). In this imminent loss situation use judgment and classify as if the thresholds are 

exceeded.

UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY

ANY loss or ANY Potential Loss of 

Containment

ANY loss or ANY Potential Loss of EITHER 

Fuel Clad or RCS

Loss or Potential Loss of ANY two Barriers Loss of ANY two Barriers AND

Loss or Potential Loss of Third Barrier

Fuel Clad Barrier Example EALS RCS Barrier Example EALS Containment Barrier Example EALS

LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS

Revision 01/2003 5-F-11

GREATER THAN (sitespecific) degree FGREATER THAN (sitespecific) degree FNot applicable Core exit thermocouples in 

excess of 1200 degrees 

and restoration procedures 

not effective within 15 

minutes; or, core exit 

thermocouples in excess 

of 700 degrees with 

reactor vessel level below 

top of active fuel and 

restoration procedures not 

effective within 15 minutes
TABLE 5-F-4

PWR Emergency Action Level

Fission Product Barrier Reference Table

Thresholds For LOSS or POTENTIAL LOSS of Barriers*

*Determine which combination of the three barriers are lost or have a potential loss and use the following key to classify the event. Also an event for multiple events could occur which result 

in the conclusion that exceeding the loss or potential loss thresholds is imminent (i.e., within 1 to 2 hours). In this imminent loss situation use judgment and classify as if the thresholds are 

exceeded.

UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY

ANY loss or ANY Potential Loss of 

Containment

ANY loss or ANY Potential Loss of EITHER 

Fuel Clad or RCS

Loss or Potential Loss of ANY two Barriers Loss of ANY two Barriers AND

Loss or Potential Loss of Third Barrier

Fuel Clad Barrier Example EALS RCS Barrier Example EALS Containment Barrier Example EALS

LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS

Revision 01/2003 5-F-12

OR OR OR

4. Reactor Vessel Water Level 3. SG Tube Rupture 4. SG Secondary Side Release with P-to-S Leakage

Not Applicable Level LESS than (sitespecific) valueSGTR that results in an 

ECCS (SI) Actuation

Not Applicable RUPTURED S/G is also 

FAULTED outside of 

containment 

OR

Primary-to-Secondary

leakrate greater than 10 

gpm with nonisolable 

steam release from 

affected S/G to the 

environment

Not applicable

OR

5. CNMT Isolation Valves Status After CNMT 

Isolation

Valve(s) not closed AND 

downstream pathway to 

the environment exists 

Not Applicable

OR OR OR

5. Containment Radiation Monitoring 4. Containment Radiation Monitoring 6. Significant Radioactive Inventory in Containment

Containment rad monitor 

reading GREATER THAN 

(site-specific) R/hr

Not Applicable Containment rad monitor 

reading GREATER THAN 

(site-specific) R/hr

Not Applicable Not Applicable Containment rad monitor 

reading GREATER THAN 

(site-specific) R/hr
TABLE 5-F-4

PWR Emergency Action Level

Fission Product Barrier Reference Table

Thresholds For LOSS or POTENTIAL LOSS of Barriers*

*Determine which combination of the three barriers are lost or have a potential loss and use the following key to classify the event. Also an event for multiple events could occur which result 

in the conclusion that exceeding the loss or potential loss thresholds is imminent (i.e., within 1 to 2 hours). In this imminent loss situation use judgment and classify as if the thresholds are 

exceeded.

UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY

ANY loss or ANY Potential Loss of 

Containment

ANY loss or ANY Potential Loss of EITHER 

Fuel Clad or RCS

Loss or Potential Loss of ANY two Barriers Loss of ANY two Barriers AND

Loss or Potential Loss of Third Barrier

Fuel Clad Barrier Example EALS RCS Barrier Example EALS Containment Barrier Example EALS

LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS

Revision 01/2003 5-F-13

OR OR OR

6. Other (Site-Specific) Indications 5. Other (Site-Specific) Indications 7. Other (site-specific) Indications

(Site specific ) as applicable (Site specific) as 

applicable

(Site-specific) as 

applicable

(Site-specific) as 

applicable

(Site specific) as 

applicable

(Site specific) as 

applicable

OR OR OR

7. Emergency Director Judgment 6. Emergency Director Judgment 8. Emergency Director Judgment

Any condition in the opinion of the Emergency Director that

indicates Loss or Potential Loss of the Fuel Clad Barrier

Any condition in the opinion of the Emergency Director that 

indicate Loss or Potential Loss of the RCS Barrier

Any condition in the opinion of the Emergency Director

that indicates Loss or Potential Loss of the Containment 

barrier
Revision 01/2003 5-F-14

Basis Information For Table 5-F-4

PWR Emergency Action Level

Fission Product Barrier Reference Table

FUEL CLAD BARRIER EXAMPLE EALs: (1 or 2 or 3 or 4 or 5 or 6)

The Fuel Clad Barrier is the zircalloy or stainless steel tubes that contain the fuel pellets.

1. Critical Safety Function Status

This EAL is for PWRs using Critical Safety Function Status Tree (CSFST) monitoring and

functional restoration procedures. For more information, please refer to Section 3.9 of this report. 

RED path indicates an extreme challenge to the safety function. ORANGE path indicates a severe 

challenge to the safety function.

Core Cooling - ORANGE indicates subcooling has been lost and that some clad damage may 

occur. Heat Sink - RED indicates the ultimate heat sink function is under extreme challenge and 

thus these two items indicate potential loss of the Fuel Clad Barrier.

Core Cooling - RED indicates significant superheating and core uncovery and is considered to 

indicate loss of the Fuel Clad Barrier.

2. Primary Coolant Activity Level

This (site-specific) value corresponds to 300 μCi/gm I131 equivalent. Assessment by the NUMARC 

EAL Task Force indicates that this amount of coolant activity is well above that expected for iodine 

spikes and corresponds to less than 5% fuel clad damage. This amount of radioactivity indicates 

significant clad damage and thus the Fuel Clad Barrier is considered lost. The value expressed can 

be either in mR/hr observed on the sample or as μCi/gm results from analysis. 

There is no equivalent "Potential Loss" EAL for this item.

3. Core Exit Thermocouple Readings

Core Exit Thermocouple Readings are included in addition to the Critical Safety Functions to 

include conditions when the CSFs may not be in use (initiation after SI is blocked) or plants which 

do not have a CSF scheme.

The "Loss" EAL (site-specific) reading should correspond to significant superheating of the coolant. 

This value typically corresponds to the temperature reading that indicates core cooling - RED in 

Fuel Clad Barrier EAL #1 which is usually about 1200 degrees F.

The "Potential Loss" EAL (site-specific) reading should correspond to loss of subcooling. This value 

typically corresponds to the temperature reading that indicates core cooling - ORANGE in Fuel Clad 

Barrier EAL #1 which is usually about 700 to 900 degrees F.

4. Reactor Vessel Water Level

There is no "Loss" EAL corresponding to this item because it is better covered by the other Fuel 

Clad Barrier "Loss" EALs.
Revision 01/2003 5-F-15

The (site-specific) value for the "Potential Loss" EAL corresponds to the top of the active fuel. For 

sites using CSFSTs, the "Potential Loss" EAL is defined by the Core Cooling - ORANGE path. The 

(site-specific) value in this EAL should be consistent with the CSFST value.

5. Containment Radiation Monitoring

The (site-specific) reading is a value which indicates the release of reactor coolant, with elevated 

activity indicative of fuel damage, into the containment. The reading should be calculated assuming 

the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory 

associated with a concentration of 300 μCi/gm dose equivalent I-131 into the containment

atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the 

maximum concentrations (including iodine spiking) allowed within technical specifications and are 

therefore indicative of fuel damage. This value is higher than that specified for RCS barrier Loss 

EAL #4. Thus, this EAL indicates a loss of both the fuel clad barrier and a loss of RCS barrier.

There is no "Potential Loss" EAL associated with this item.

6. Other (Site-Specific) Indications

This EAL is to cover other (site-specific) indications that may indicate loss or potential loss of the 

Fuel Clad barrier, including indications from containment air monitors or any other (site-specific) 

instrumentation.

7. Emergency Director Judgment

This EAL addresses any other factors that are to be used by the Emergency Director in determining 

whether the Fuel Clad barrier is lost or potentially lost. In addition, the inability to monitor the barrier 

should also be incorporated in this EAL as a factor in Emergency Director judgment that the barrier 

may be considered lost or potentially lost. (See also IC SG1, "Prolonged Loss or All Offsite Power 

and Prolonged Loss of All Onsite AC Power", for additional information.)

RCS BARRIER EXAMPLE EALs: (1 or 2 or 3 or 4 or 5 or 6)

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. Critical Safety Function Status

This EAL is for PWRs using Critical Safety Function Status Tree (CSFST) monitoring and

functional restoration procedures. For more information, please refer to Section 3.9 of this report. 

RED path indicates an extreme challenge to the safety function derived from appropriate instrument 

readings, and these CSFs indicate a potential loss of RCS barrier.

There is no "Loss" EAL associated with this item.

2. RCS Leak Rate

The "Loss" EAL addresses conditions where leakage from the RCS is greater than available 

inventory control capacity such that a loss of subcooling has occurred. The loss of subcooling is 

the fundamental indication that the inventory control systems are inadequate in maintaining RCS 

pressure and inventory against the mass loss through the leak.
Revision 01/2003 5-F-16

The "Potential Loss" EAL is based on the inability to maintain normal liquid inventory within the 

Reactor Coolant System (RCS) by normal operation of the Chemical and Volume Control System 

which is considered as one centrifugal charging pump discharging to the charging header. A

second charging pump being required is indicative of a substantial RCS leak. For plants with low 

capacity charging pumps, a 50 gpm leak rate value may be used to indicate the Potential Loss. 

3. SG Tube Rupture

This EAL is intended to address the full spectrum of Steam Generator (SG) tube rupture events in 

conjunction with Containment Barrier "Loss" EAL #4 and Fuel Clad Barrier EALs. The "Loss" EAL 

addresses RUPTURED SG(s) for which the leakage is large enough to cause actuation of ECCS 

(SI). This is consistent to the RCS Barrier “Potential Loss” EAL #2. For plants that have

implemented W.O.G. emergency response guides, this condition is described by “entry into E-3 

required by EOPs”. By itself, this EAL will result in the declaration of an Alert. However, if the SG is 

also FAULTED (i.e., two barriers failed), the declaration escalates to a Site Area Emergency per 

Containment Barrier "Loss" EAL #4. 

There is no "Potential Loss" EAL.

4. Containment Radiation Monitoring

The (site-specific) reading is a value which indicates the release of reactor coolant to the

containment. The reading should be calculated assuming the instantaneous release and dispersal 

of the reactor coolant noble gas and iodine inventory associated with normal operating

concentrations (i.e., within T/S) into the containment atmosphere. This reading will be less than that 

specified for Fuel Clad Barrier EAL #5. Thus, this EAL would be indicative of a RCS leak only. If the 

radiation monitor reading increased to that specified by Fuel Clad Barrier EAL #5, fuel damage 

would also be indicated.

However, if the site specific physical location of the containment radiation monitor is such that 

radiation from a cloud of released RCS gases could not be distinguished from radiation from nearby 

piping and components containing elevated reactor coolant activity, this EAL should be omitted and 

other site specific indications of RCS leakage substituted.

There is no "Potential Loss" EAL associated with this item.

5. Other (Site-Specific) Indications

This EAL is to cover other (site-specific) indications that may indicate loss or potential loss of the 

RCS barrier, including indications from containment air monitors or any other (site-specific) 

instrumentation.

6. Emergency Director Judgment

This EAL addresses any other factors that are to be used by the Emergency Director in determining 

whether the RCS barrier is lost or potentially lost. In addition, the inability to monitor the barrier 

should also be incorporated in this EAL as a factor in Emergency Director judgment that the barrier 

may be considered lost or potentially lost. (See also IC SG1, "Prolonged Loss of All Offsite Power 

and Prolonged Loss of All Onsite AC Power", for additional information.)
Revision 01/2003 5-F-17

CONTAINMENT BARRIER EXAMPLE EALs: (1 or 2 or 3 or 4 or 5 or 6 or 7 or 8)

The Containment Barrier includes the containment building, its 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.

1. Critical Safety Function Status

This EAL is for PWRs using Critical Safety Function Status Tree (CSFST) monitoring and

functional restoration procedures. For more information, please refer to Section 3.9 of this report. 

RED path indicates an extreme challenge to the safety function derived from appropriate instrument 

readings and/or sampling results, and thus represents a potential loss of containment. Conditions 

leading to a containment RED path result from RCS barrier and/or Fuel Clad Barrier Loss. Thus, 

this EAL is primarily a discriminator between Site Area Emergency and General Emergency

representing a potential loss of the third barrier.

There is no "Loss" EAL associated with this item.

2. Containment Pressure

Rapid unexplained loss of pressure (i.e., not attributable to containment spray or condensation 

effects) following an initial pressure increase indicates a loss of containment integrity. Containment 

pressure and sump levels should increase as a result of the mass and energy release into

containment from a LOCA. Thus, sump level or pressure not increasing indicates containment 

bypass and a loss of containment integrity. The (site-specific) PSIG for potential loss of

containment is based on the containment design pressure. Existence of an explosive mixture 

means a hydrogen and oxygen concentration of at least the lower deflagration limit curve exists. 

The indications of potential loss under this EAL corresponds to some of those leading to the RED 

path in EAL #1 above and may be declared by those sites using CSFSTs. As described above, this 

EAL is primarily a discriminator between Site Area Emergency and General Emergency

representing a potential loss of the third barrier.

The second potential loss EAL represents a potential loss of containment in that the containment 

heat removal/depressurization system (e.g., containment sprays, ice condenser fans, etc., but not 

including containment venting strategies) are either lost or performing in a degraded manner, as 

indicated by containment pressure greater than the setpoint at which the equipment was supposed 

to have actuated.

3. Core Exit Thermocouples

In this EAL, the function restoration procedures are those emergency operating procedures that 

address the recovery of the core cooling critical safety functions. The procedure is considered 

effective if the temperature is decreasing or if the vessel water level is increasing. For units using 

the CSF status trees a direct correlation to those status trees can be made if the effectiveness of 

the restoration procedures is also evaluated as stated below. 

Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures 

can arrest core degradation within the reactor vessel in a significant fraction of the core damage 

scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is 

appropriate to provide a reasonable period to allow function restoration procedures to arrest the 

core melt sequence. Whether or not the procedures will be effective should be apparent within 15 

minutes. The Emergency Director should make the declaration as soon as it is determined that the 
Revision 01/2003 5-F-18

procedures have been, or will be ineffective. The reactor vessel level chosen should be consistent 

with the emergency response guides applicable to the facility. 

The conditions in this potential loss EAL represent an imminent core melt sequence which, if not 

corrected, could lead to vessel failure and an increased potential for containment failure. In

conjunction with the Core Cooling and Heat Sink criteria in the Fuel and RCS barrier columns, this 

EAL would result in the declaration of a General Emergency -- loss of two barriers and the potential 

loss of a third. If the function restoration procedures are ineffective, there is no "success" path.

There is no "Loss" EAL associated with this item.

4. SG Secondary Side Release With Primary To Secondary Leakage

This “loss” EAL recognizes that SG tube leakage can represent a bypass of the containment barrier 

as well as a loss of the RCS barrier. The first “loss” EAL addresses the condition in which a 

RUPTURED steam generator is also FAULTED. This condition represents a bypass of the RCS 

and containment barriers. In conjunction with RCS Barrier “loss” EAL #3, this would always result in 

the declaration of a Site Area Emergency. 

The second “loss” EAL addresses SG tube leaks that exceed 10 gpm in conjunction with a 

nonisolable release path to the environment from the affected steam generator. The threshold for 

establishing the nonisolable secondary side release is intended to be a prolonged release of

radioactivity from the RUPTURED steam generator directly to the environment. This could be 

expected to occur when the main condenser is unavailable to accept the contaminated steam (i.e., 

SGTR with concurrent loss of offsite power and the RUPTURED steam generator is required for 

plant cooldown or a stuck open relief valve). If the main condenser is available, there may be 

releases via air ejectors, gland seal exhausters, and other similar controlled, and often monitored, 

pathways. These pathways do not meet the intent of a nonisolable release path to the environment. 

These minor releases are assessed using Abnormal Rad Levels / Radiological Effluent ICs. 

Users should realize that the two “loss” EALs described above could be considered redundant. 

This was recognized during the development process. The inclusion of an EAL that uses 

Emegency Procedure commonly used terms like “ruptured and faulted” adds to the ease of the 

classification process and has been included based on this human factor concern.

The leakage threshold for this EAL has been increased with Revision 3. In the earlier revision, the 

threshold was leakage greater than T/S allowable. Since the prior revision, many plants have 

implemented reduced steam generator T/S limits (e.g., 150 gpd) as a defense in depth associated 

with alternate steam generator plugging criteria. The 150 gpd threshold is deemed too low for use 

as an emergency threshold. A pressure boundary leakage of 10 gpm was used as the threshold in 

IC SU5, RCS Leakage, and is deemed appropriate for this EAL. For smaller breaks, not exceeding 

the normal charging capacity threshold in RCS Barrier "Potential Loss" EAL #2 (RCS Leak Rate) or 

not resulting in ECCS actuation in EAL #3 (SG Tube Rupture), this EAL results in a NOUE. For 

larger breaks, RCS barrier EALs #2 and #3 would result in an Alert. For SG tube ruptures which 

may involve multiple steam generators or unisolable secondary line breaks, this EAL would exist in 

conjunction with RCS barrier "Loss" EAL #3 and would result in a Site Area Emergency. Escalation 

to General Emergency would be based on "Potential Loss" of the Fuel Clad Barrier.
Revision 01/2003 5-F-19

5. Containment Isolation Valve Status After Containment Isolation

This EAL is intended to address incomplete containment isolation that allows direct release to the 

environment. It represents a loss of the containment barrier.

The use of the modifier “direct” in defining the release path discriminates against release paths 

through interfacing liquid systems. The existence of an in–line charcoal filter does not make a 

release path indirect since the filter is not effective at removing fission noble gases. Typical filters 

have an efficiency of 95-99% removal of iodine. Given the magnitude of the core inventory of iodine, 

significant releases could still occur. In addition, since the fission product release would be driven 

by boiling in the reactor vessel, the high humidity in the release stream can be expected to render 

the filters ineffective in a short period.

There is no "Potential Loss" EAL associated with this item.

6. Significant Radioactive Inventory in Containment

The (site-specific) reading is a value which indicates significant fuel damage well in excess of the 

EALs associated with both loss of Fuel Clad and loss of RCS Barriers. As stated in Section 3.8, a 

major release of radioactivity requiring offsite protective actions from core damage is not possible 

unless a major failure of fuel cladding allows radioactive material to be released from the core into 

the reactor coolant. 

Regardless of whether containment is challenged, this amount of activity in containment, if

released, could have such severe consequences that it is prudent to treat this as a potential loss of 

containment, such that a General Emergency declaration is warranted. NUREG-1228, "Source 

Estimations During Incident Response to Severe Nuclear Power Plant Accidents," indicates that 

such conditions do not exist when the amount of clad damage is less than 20%. Unless there is a 

(site-specific) analysis justifying a higher value, it is recommended that a radiation monitor reading 

corresponding to 20% fuel clad damage be specified here.

There is no "Loss" EAL associated with this item.

7. Other (Site-Specific) Indications

This EAL should cover other (site-specific) indications that may unambiguously indicate loss or 

potential loss of the containment barrier, including indications from area or ventilation monitors in 

containment annulus or other contiguous buildings. If site emergency operating procedures provide 

for venting of the containment during an emergency as a means of preventing catastrophic failure, a 

Loss EAL should be included for the containment barrier. This EAL should be declared as soon as 

such venting is imminent. Containment venting as part of recovery actions is classified in

accordance with the radiological effluent ICs.

8. Emergency Director Judgment

This EAL addresses any other factors that are to be used by the Emergency Director in determining 

whether the Containment barrier is lost or potentially lost. In addition, the inability to monitor the 

barrier should also be incorporated in this EAL as a factor in Emergency Director judgment that the 

barrier may be considered lost or potentially lost. (See also IC SG1, "Prolonged Loss of All Offsite 

Power and Prolonged Loss of All Onsite AC Power", for additional information.)
Revision 01/2003 5-H-1

TABLE 5-H-1

Recognition Category H

Hazards and Other Conditions Affecting Plant Safety

INITIATING CONDITION MATRIX

NOUE ALERT SITE AREA EMERGENCY GENERAL EMERGENCY

HU1 Natural and Destructive 

Phenomena Affecting the 

PROTECTED AREA.

Op. Modes: All

HA1 Natural and Destructive 

Phenomena Affecting the Plant 

VITAL AREA.

Op. Modes: All

HU2 FIRE Within PROTECTED AREA 

Boundary Not Extinguished 

Within 15 Minutes of Detection.

Op. Modes: All

HA2 FIRE or EXPLOSION Affecting 

the Operability of Plant Safety 

Systems Required to Establish 

or Maintain Safe Shutdown.

Op. Modes: All

HU3 Release of Toxic or Flammable 

Gases Deemed Detrimental to 

Safe Operation of the Plant.

Op. Modes: All

HA3 Release of Toxic or Flammable 

Gases Within or Contiguous to a 

VITAL AREA Which Jeopardizes 

Operation of Safety Systems 

Required to Establish or 

Maintain Safe Shutdown.

Op. Modes: All

HU4 Confirmed Security Event Which 

Indicates a Potential 

Degradation in the Level of 

Safety of the Plant.

Op. Modes: All

HA4 Confirmed Security Event in a 

Plant PROTECTED AREA.

Op. Modes: All

HS1 Confirmed Security Event in a 

Plant VITAL AREA.

Op. Modes: All

HG1 Security Event Resulting in Loss 

Of Physical Control of the 

Facility.

Op. Modes: All

HU5 Other Conditions Existing Which 

in the Judgment of the 

Emergency Director Warrant 

Declaration of a NOUE.

Op. Modes: All

HA6 Other Conditions Existing Which 

in the Judgment of the 

Emergency Director Warrant 

Declaration of an Alert.

Op. Modes: All

HS3 Other Conditions Existing Which 

in the Judgment of the 

Emergency Director Warrant 

Declaration of Site Area 

Emergency.

Op. Modes: All

HG2 Other Conditions Existing Which 

in the Judgment of the 

Emergency Director Warrant 

Declaration of General 

Emergency.

Op. Modes: All

HA5 Control Room Evacuation Has 

Been Initiated.

Op. Modes: All

HS2 Control Room Evacuation Has 

Been Initiated and Plant Control 

Cannot Be Established.

Op. Modes: All
Revision 01/2003 5-H-2

This page intentionally blank.
Revision 01/2003 5-H-3

HAZARDS AND OTHER CONDITIONS

AFFECTING PLANT SAFETY

HU1

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

Natural and Destructive Phenomena Affecting the PROTECTED AREA.

Operating Mode Applicability: All

Example Emergency Action Level: (1 or 2 or 3 or 4 or 5 or 6 or 7)

1. (Site-Specific) method indicates felt earthquake.

2. Report by plant personnel of tornado or high winds greater than (site-specific) mph striking 

within PROTECTED AREA boundary.

3. Vehicle crash into plant structures or systems within PROTECTED AREA boundary.

4. Report by plant personnel of an unanticipated EXPLOSION within PROTECTED AREA

boundary resulting in VISIBLE DAMAGE to permanent structure or equipment.

5. Report of turbine failure resulting in casing penetration or damage to turbine or generator 

seals.

6. Uncontrolled flooding in (site-specific) areas of the plant that has the potential to affect safety 

related equipment needed for the current operating mode.

7. (Site-Specific) occurrences affecting the PROTECTED AREA.

Basis:

NOUE in this IC are categorized on the basis of the occurrence of an event of sufficient magnitude 

to be of concern to plant operators. Areas identified in the EALs define the location of the event 

based on the potential for damage to equipment contained therein. Escalation of the event to an 

Alert occurs when the magnitude of the event is sufficient to result in damage to equipment 

contained in the specified location.

EAL #1 should be developed on site-specific basis. Damage may be caused to some portions of 

the site, but should not affect ability of safety functions to operate. Method of detection can be based 

on instrumentation, validated by a reliable source, or operator assessment. As defined in the EPRIsponsored "Guidelines for Nuclear Plant Response to an Earthquake", dated October 1989, a "felt 

earthquake" is:

An earthquake of sufficient intensity such that: (a) the vibratory ground motion is felt at the 

nuclear plant site and recognized as an earthquake based on a consensus of control room 

operators on duty at the time, and (b) for plants with operable seismic instrumentation, the 

seismic switches of the plant are activated. For most plants with seismic instrumentation, the 

seismic switches are set at an acceleration of about 0.01g.
Revision 01/2003 5-H-4

EAL #2 is based on the assumption that a tornado striking (touching down) or high winds within the 

PROTECTED AREA may have potentially damaged plant structures containing functions or

systems required for safe shutdown of the plant. The high wind site specific value in EAL#2 should 

be based on site-specific FSAR design basis. If such damage is confirmed visually or by other inplant indications, the event may be escalated to Alert.

EAL #3 is intended to address crashes of vehicle types large enough to cause significant damage 

to plant structures containing functions and systems required for safe shutdown of the plant. If the 

crash is confirmed to affect a plant VITAL AREA, the event may be escalated to Alert. 

For EAL #4 only those EXPLOSIONs of sufficient force to damage permanent structures or

equipment within the PROTECTED AREA should be considered. No attempt is made in this EAL to 

assess the actual magnitude of the damage. The occurrence of the EXPLOSION with reports of 

evidence of damage is sufficient for declaration. The Emergency director also needs to consider 

any security aspects of the EXPLOSION, if applicable. 

EAL #5 is intended to address main turbine rotating component failures of sufficient magnitude to 

cause observable damage to the turbine casing or to the seals of the turbine generator. Of major 

concern is the potential for leakage of combustible fluids (lubricating oils) and gases (hydrogen 

cooling) to the plant environs. Actual FIREs and flammable gas build up are appropriately classified 

via HU2 and HU3. Generator seal damage observed after generator purge does not meet the intent 

of this EAL because it did not impact normal operation of the plant. This EAL is consistent with the 

definition of a NOUE while maintaining the anticipatory nature desired and recognizing the risk to 

non-safety related equipment. Escalation of the emergency classification is based on potential 

damage done by missiles generated by the failure or by the radiological releases for a BWR, or in 

conjunction with a steam generator tube rupture, for a PWR. These latter events would be

classified by the radiological ICs or Fission Product Barrier ICs.

EAL #6 addresses the effect of flooding caused by internal events such as component failures, 

equipment misalignment, or outage activity mishaps. The site-specific areas include those areas

that contain systems required for safe shutdown of the plant, that are not designed to be wetted or 

submerged. Escalation of the emergency classification is based on the damage caused or by 

access restrictions that prevent necessary plant operations or systems monitoring. The plant’s 

IPEEE may provide insight into areas to be considered when developing this EAL.

EAL #7 covers other site-specific phenomena such as hurricane, flood, or seiche. These EALs can 

also be precursors of more serious events. In particular, sites subject to severe weather as defined 

in the NUMARC station blackout initiatives, should include an EAL based on activation of the severe 

weather mitigation procedures (e.g., precautionary shutdowns, diesel testing, staff call-outs, etc.).
Revision 01/2003 5-H-5

HAZARDS AND OTHER CONDITIONS

AFFECTING PLANT SAFETY

HU2

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

FIRE Within PROTECTED AREA Boundary Not Extinguished Within 15 Minutes of 

Detection.

Operating Mode Applicability: All

Example Emergency Action Level:

1. FIRE in buildings or areas contiguous to any of the following (site-specific) areas not

extinguished within 15 minutes of control room notification or verification of a control room 

alarm:

(Site-specific) list

Basis:

The purpose of this IC is to address the magnitude and extent of FIREs that may be potentially 

significant precursors to damage to safety systems. As used here, Detection is visual observation 

and report by plant personnel or sensor alarm indication. The 15 minute time period begins with a 

credible notification that a FIRE is occurring, or indication of a VALID fire detection system alarm. 

Verification of a fire detection system alarm includes actions that can be taken with the control room 

or other nearby site-specific location to ensure that the alarm is not spurious. A verified alarm is 

assumed to be an indication of a FIRE unless it is disproved within the 15 minute period by 

personnel dispatched to the scene. In other words, a personnel report from the scene may be used 

to disprove a sensor alarm if received within 15 minutes of the alarm, but shall not be required to 

verify the alarm. 

The intent of this 15 minute duration is to size the FIRE and to discriminate against small FIREs 

that are readily extinguished (e.g., smoldering waste paper basket). The site-specific list should be 

limited and applies to buildings and areas contiguous (in actual contact with or immediately

adjacent) to plant VITAL AREAs or other significant buildings or areas. The intent of this IC is not to 

include buildings (i.e., warehouses) or areas that are not contiguous (in actual contact with or 

immediately adjacent) to plant VITAL AREAs. This excludes FIREs within administration buildings, 

waste-basket FIREs, and other small FIREs of no safety consequence.

Escalation to a higher emergency class is by IC HA4, "FIRE Affecting the Operability of Plant Safety 

Systems Required for the Current Operating Mode".
Revision 01/2003 5-H-6

HAZARDS AND OTHER CONDITIONS

AFFECTING PLANT SAFETY

HU3

Initiating Condition – NOTIFICATION OF UNUSUAL EVENT

Release of Toxic or Flammable Gases Deemed Detrimental to Normal Operation of the 

Plant.

Operating Mode Applicability: All

Example Emergency Action Levels: (1 or 2)

1. Report or detection of toxic or flammable gases that has or could enter the site area boundary 

in amounts that can affect NORMAL PLANT OPERATIONS.

2. Report by Local, County or State Officials for evacuation or sheltering of site personnel based 

on an offsite event.

Basis:

This IC is based on the existence of uncontrolled releases of toxic or flammable gas that may enter 

the site boundary and affect normal plant operations. It is intended that releases of toxic or

flammable gases are of sufficient quantity, and the release point of such gases is such that normal 

plant operations would be affected. This would preclude small or incidental releases, or releases 

that do not impact structures needed for plant operation. The EALs are intended to not require 

significant assessment or quantification. The IC assumes an uncontrolled process that has the 

potential to affect plant operations, or personnel safety.

Escalation of this EAL is via HA3, which involves a quantified release of toxic or flammable gas 

affecting VITAL AREAs.
Revision 01/2003 5-H-7

HAZARDS AND OTHER CONDITIONS

AFFECTING PLANT SAFETY

HU4

Initiating Condition – NOTIFICATION OF UNUSUAL EVENT

Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety 

of the Plant.

Operating Mode Applicability: All

Example Emergency Action Levels:

1. Security events as determined from (site-specific) Safeguards Contingency Plan 

and reported by the (site-specific) security shift supervision

2. A credible site specific security threat notification.

Basis:

Reference is made to (site-specific) security shift supervision because these individuals are the 

designated personnel on-site qualified and trained to confirm that a security event is occurring or 

has occurred. Training on security event classification confirmation is closely controlled due to the 

strict secrecy controls placed on the plant Safeguards Contingency Plan.

This EAL 1 is based on (site-specific) Site Security Plans. Security events which do not represent a 

potential degradation in the level of safety of the plant, are reported under 10 CFR 73.71 or in some 

cases under 10 CFR 50.72. Examples of security events that indicate Potential Degradation in the 

Level of Safety of the Plant are provided below for consideration. 

Consideration should be given to the following types of events when evaluating an event against the 

criteria of the site specific Security Contingency Plan: SABOTAGE, HOSTAGE / EXTORTION, 

CIVIL DISTURBANCE, and STRIKE ACTION. 

INTRUSION into the plant PROTECTED AREA by a HOSTILE FORCE would result in EAL

escalation to an ALERT.

The intent of EAL 2 is to ensure that appropriate notifications for the security threat are made in a 

timely manner. Only the plant to which the specific threat is made need declare the Notification of 

an Unusual Event.

The determination of “credible” is made through use of information found in the (site-specific) 

Safeguards Contingency Plan.

A higher initial classification could be made based upon the nature and timing of the threat and 

potential consequences. The licensee shall consider upgrading the emergency response status 

and emergency classification in accordance with the [site security specific] Safeguards 

Contingency Plan and Emergency Plans.
Revision 01/2003 5-H-8

HAZARDS AND OTHER CONDITIONS

AFFECTING PLANT SAFETY

HU5

Initiating Condition – NOTIFICATION OF UNUSUAL EVENT

Other Conditions Existing Which in the Judgment of the Emergency Director Warrant 

Declaration of a NOUE.

Operating Mode Applicability: All

Example Emergency Action Level:

1. Other conditions exist which in the judgment of the Emergency Director indicate that events 

are in process or have occurred which indicate a potential degradation of the level of safety of 

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

expected unless further degradation of safety systems occurs.

Basis:

This EAL is intended to address 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 NOUE emergency class.

From a broad perspective, one area that may warrant Emergency Director judgment is related to 

likely or actual breakdown of site-specific event mitigating actions. Examples to consider include 

inadequate emergency response procedures, transient response either unexpected or not

understood, failure or unavailability of emergency systems during an accident in excess of that 

assumed in accident analysis, or insufficient availability of equipment and/or support personnel.
Revision 01/2003 5-H-9

 HAZARDS AND OTHER CONDITIONS

AFFECTING PLANT SAFETY

HA1

Initiating Condition -- ALERT

Natural and Destructive Phenomena Affecting the Plant VITAL AREA.

Operating Mode Applicability: All

Example Emergency Action Levels: (1 or 2 or 3 or 4 or 5 or 6)

1. (Site-Specific) method indicates Seismic Event greater than Operating Basis Earthquake 

(OBE).

2. Tornado or high winds greater than (site-specific) mph within PROTECTED AREA boundary 

and resulting in VISIBLE DAMAGE to any of the following plant structures / equipment or 

Control Room indication of degraded performance of those systems.

• Reactor Building

• Intake Building

• Ultimate Heat Sink

• Refueling Water Storage Tank

• Diesel Generator Building

• Turbine Building

• Condensate Storage Tank

• Control Room

• Other (Site-Specific) Structures.

3. Vehicle crash within PROTECTED AREA boundary and resulting in VISIBLE DAMAGE to any 

of the following plant structures or equipment therein or control indication of degraded

performance of those systems:

• Reactor Building

• Intake Building

• Ultimate Heat Sink

• Refueling Water Storage Tank

• Diesel Generator Building

• Turbine Building

• Condensate Storage Tank

• Control Room

• Other (Site-Specific) Structures.

4. Turbine failure-generated missiles result in any VISIBLE DAMAGE to or penetration of any of 

the following plant areas: (site-specific) list.

5. Uncontrolled flooding in (site-specific) areas of the plant that results in degraded safety

system performance as indicated in the control room or that creates industrial safety hazards 

(e.g., electric shock) that precludes access necessary to operate or monitor safety

equipment.
Revision 01/2003 5-H-10

6. (Site-Specific) occurrences within PROTECTED AREA boundary and resulting in VISIBLE 

DAMAGE to plant structures containing equipment necessary for safe shutdown, or has 

caused damage as evidenced by control room indication of degraded performance of those 

systems.

Basis:

The EALs in this IC escalate from the NOUE EALs in HU1 in that the occurrence of the event has 

resulted in VISIBLE DAMAGE to plant structures or areas containing equipment necessary for a 

safe shutdown, or has caused damage to the safety systems in those structures evidenced by 

control indications of degraded system response or performance. The occurrence of VISIBLE 

DAMAGE and/or degraded system response is intended to discriminate against lesser events. The 

initial "report" should not be interpreted as mandating a lengthy damage assessment prior to 

classification. No attempt is made in this EAL to assess the actual magnitude of the damage. The 

significance here is not that a particular system or structure was damaged, but rather, that the 

event was of sufficient magnitude to cause this degradation. Escalation to higher classifications 

occur on the basis of other ICs (e.g., System Malfunction).

EAL #1 should be based on site-specific FSAR design basis. Seismic events of this magnitude can 

result in a plant VITAL AREA being subjected to forces beyond design limits, and thus damage may 

be assumed to have occurred to plant safety systems. See EPRI-sponsored "Guidelines for 

Nuclear Plant Response to an Earthquake", dated October 1989, for information on seismic event 

categories.

EAL #2 should be based on site-specific FSAR design basis. Wind loads of this magnitude can 

cause damage to safety functions.

EAL #s 2, 3, 4, 5 should specify site-specific structures or areas containing systems and functions 

required for safe shutdown of the plant.

EAL #3 is intended to address crashes of vehicle types large enough to cause significant damage 

to plant structures containing functions and systems required for safe shutdown of the plant.

EAL #4 is intended to address the threat to safety related equipment imposed by missiles 

generated by main turbine rotating component failures. This site-specific list of areas should include 

all areas containing safety-related equipment, their controls, and their power supplies. This EAL is, 

therefore, consistent with the definition of an ALERT in that if missiles have damaged or penetrated 

areas containing safety-related equipment the potential exists for substantial degradation of the level 

of safety of the plant.

EAL #5 addresses the effect of internal flooding that has resulted in degraded performance of 

systems affected by the flooding, or has created industrial safety hazards (e.g., electrical shock) 

that preclude necessary access to operate or monitor safety equipment. The inability to operate or 

monitor safety equipment represents a potential for substantial degradation of the level of safety of 

the plant. This flooding may have been caused by internal events such as component failures, 

equipment misalignment, or outage activity mishaps. The site-specific areas includes those areas 

that contain systems required for safe shutdown of the plant, that are not designed to be wetted or 

submerged. The plant’s IPEEE may provide insight into areas to be considered when developing 

this EAL.

EAL #6 covers other site-specific phenomena such as hurricane, flood, or seiche. These EALs can 

also be precursors of more serious events. 
Revision 01/2003 5-H-11

HAZARDS AND OTHER CONDITIONS

AFFECTING PLANT SAFETY

HA2

Initiating Condition -- ALERT

FIRE or EXPLOSION Affecting the Operability of Plant Safety Systems Required to 

Establish or Maintain Safe Shutdown.

Operating Mode Applicability: All

Example Emergency Action Level:

1. FIRE or EXPLOSION in any of the following (site-specific) areas:

(Site-specific) list

AND

Affected system parameter indications show degraded performance or plant personnel report 

VISIBLE DAMAGE to permanent structures or equipment within the specified area.

Basis:

Site-specific areas containing functions and systems required for the safe shutdown of the plant 

should be specified. Site-Specific Safe Shutdown Analysis should be consulted for equipment and 

plant areas required to establish or maintain safe shutdown. This will make it easier to determine if 

the FIRE or EXPLOSION is potentially affecting one or more redundant trains of safety systems. 

Escalation to a higher emergency class, if appropriate, will be based on System Malfunction, 

Fission Product Barrier Degradation, Abnormal Rad Levels / Radiological Effluent, or Emergency 

Director Judgment ICs. 

This EAL addresses a FIRE / EXPLOSION and not the degradation in performance of affected 

systems. System degradation is addressed in the System Malfunction EALs. The reference to 

damage of systems is used to identify the magnitude of the FIRE / EXPLOSION and to discriminate 

against minor FIREs / EXPLOSIONs. The reference to safety systems is included to discriminate 

against FIREs / EXPLOSIONs in areas having a low probability of affecting safe operation. The 

significance here is not that a safety system was degraded but the fact that the FIRE / EXPLOSION 

was large enough to cause damage to these systems. Thus, the designation of a single train was 

intentional and is appropriate when the FIRE / EXPLOSION is large enough to affect more than one 

component.

This situation is not the same as removing equipment for maintenance that is covered by a plant’s 

Technical Specifications. Removal of equipment for maintenance is a planned activity controlled in 

accordance with procedures and, as such, does not constitute a substantial degradation in the level 

of safety of the plant. A FIRE / EXPLOSION is an UNPLANNED activity and, as such, does 

constitute a substantial degradation in the level of safety of the plant. In this situation, an Alert 

classification is warranted.
Revision 01/2003 5-H-12

The inclusion of a "report of VISIBLE DAMAGE" should not be interpreted as mandating a lengthy 

damage assessment prior to classification. No attempt is made in this EAL to assess the actual 

magnitude of the damage. The occurrence of the EXPLOSION with reports of evidence of damage 

is sufficient for declaration. The declaration of an Alert and the activation of the Technical Support 

Center will provide the Emergency Director with the resources needed to perform these damage 

assessments. The Emergency Director also needs to consider any security aspects of the

EXPLOSIONs, if applicable. 
Revision 01/2003 5-H-13

HAZARDS AND OTHER CONDITIONS

AFFECTING PLANT SAFETY

HA3

Initiating Condition -- ALERT

Release of Toxic or Flammable Gases Within or Contiguous to a VITAL AREA Which 

Jeopardizes Operation of Systems Required to Maintain Safe Operations or Establish 

or Maintain Safe Shutdown.

Operating Mode Applicability: All

Example Emergency Action Levels: (1 or 2)

1. Report or detection of toxic gases within or contiguous to a VITAL AREA in concentrations that 

may result in an atmosphere IMMEDIATELY DANGEROUS TO LIFE AND HEALTH (IDLH).

2. Report or detection of gases in concentration greater than the LOWER FLAMMABILITY LIMIT

within or contiguous to a VITAL AREA.

Basis:

This IC is based on gases that affect the safe operation of the plant. This IC applies to buildings and 

areas contiguous to plant VITAL AREAs or other significant buildings or areas (i.e., service water 

pump house). The intent of this IC is not to include buildings (e.g., warehouses) or other areas that 

are not contiguous or immediately adjacent to plant VITAL AREAs. It is appropriate that increased 

monitoring be done to ascertain whether consequential damage has occurred. Escalation to a 

higher emergency class, if appropriate, will be based on System Malfunction, Fission Product 

Barrier Degradation, Abnormal Rad Levels / Radioactive Effluent, or Emergency Director Judgment 

ICs. 

EAL #1 is met if measurement of toxic gas concentration results in an atmosphere that is IDLH 

within a VITAL AREA or any area or building contiguous to VITAL AREA. Exposure to an IDLH 

atmosphere will result in immediate harm to unprotected personnel, and would preclude access to 

any such affected areas.

EAL #2 is met when the flammable gas concentration in a VITAL AREA or any building or area 

contiguous to a VITAL AREA exceed the LOWER FLAMMABILITY LIMIT. Flammable gasses, such 

as hydrogen and acetylene, are routinely used to maintain plant systems (hydrogen) or to repair 

equipment/components (acetylene - used in welding). This EAL addresses concentrations at which 

gases can ignite/support combustion. An uncontrolled release of flammable gasses within a facility 

structure has the potential to affect safe operation of the plant by limiting either operator or

equipment operations due to the potential for ignition and resulting equipment damage/personnel 

injury. Once it has been determined that an uncontrolled release is occurring, then sampling must 

be done to determine if the concentration of the released gas is within this range.
Revision 01/2003 5-H-14

HAZARDS AND OTHER CONDITIONS

AFFECTING PLANT SAFETY

HA4

Initiating Condition -- ALERT

Confirmed Security Event in a Plant PROTECTED AREA.

Operating Mode Applicability: All

Example Emergency Action Levels: (1 or 2)

1. INTRUSION into the plant PROTECTED AREA by a HOSTILE FORCE.

2. Other security events as determined from (site-specific) Safeguards Contingency Plan and 

reported by the (site-specific) security shift supervision

Basis:

This class of security events represents an escalated threat to plant safety above that contained in 

the NOUE. A confirmed INTRUSION report is satisfied if physical evidence indicates the presence 

of a HOSTILE FORCE within the PROTECTED AREA. 

Consideration should be given to the following types of events when evaluating an event against the 

criteria of the site specific Security Contingency Plan: SABOTAGE, HOSTAGE / EXTORTION, and 

STRIKE ACTION. The Safeguards Contingency Plan identifies numerous events/conditions that 

constitute a threat/compromise to a Station’s security. Only those events that involve Actual or 

Potential Substantial degradation to the level of safety of the plant need to be considered. The 

following events would not normally meet this requirement; (e.g., Failure by a Member of the 

Security Force to carry out an assigned/required duty, internal disturbances, loss/compromise of 

safeguards materials or strike actions).

INTRUSION into a VITAL AREA by a HOSTILE FORCE will escalate this event to a Site Area 

Emergency.

Reference is made to (site-specific) security shift supervision because these individuals are the 

designated personnel on-site qualified and trained to confirm that a security event is occurring or 

has occurred. Training on security event classification confirmation is closely controlled due to the 

strict secrecy controls placed on the plant Security Plan.
Revision 01/2003 5-H-15

 HAZARDS AND OTHER CONDITIONS

AFFECTING PLANT SAFETY

HA5

Initiating Condition -- ALERT

Control Room Evacuation Has Been Initiated.

Operating Mode Applicability: All

Example Emergency Action Level:

1. Entry into (site-specific) procedure for control room evacuation.

Basis:

With the control room evacuated, additional support, monitoring and direction through the Technical 

Support Center and/or other emergency response facility is necessary. Inability to establish plant 

control from outside the control room will escalate this event to a Site Area Emergency.
Revision 01/2003 5-H-16

HAZARDS AND OTHER CONDITIONS

AFFECTING PLANT SAFETY

HA6

Initiating Condition -- ALERT

Other Conditions Existing Which in the Judgment of the Emergency Director Warrant 

Declaration of an Alert.

Operating Mode Applicability: All

Example Emergency Action Level: 

1. Other conditions exist which in the judgment of the Emergency Director indicate that events 

are in process or have occurred which involve actual or likely potential substantial degradation 

of the level of safety of the plant. Any releases are expected to be limited to small fractions of 

the EPA Protective Action Guideline exposure levels .

Basis:

This EAL is intended to address 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 Alert emergency class.
Revision 01/2003 5-H-17

HAZARDS AND OTHER CONDITIONS

AFFECTING PLANT SAFETY

HS1

Initiating Condition – SITE AREA EMERGENCY

Confirmed Security Event in a Plant VITAL AREA.

Operating Mode Applicability: All

Example Emergency Action Levels: (1 or 2)

1. INTRUSION into the plant VITAL AREA by a HOSTILE FORCE.

2. Other security events as determined from (site-specific) Safeguards Contingency Plan and 

reported by the (site-specific) security shift supervision

Basis:

This class of security events represents an escalated threat to plant safety above that contained in 

the Alert IC in that a HOSTILE FORCE has progressed from the PROTECTED AREA to a VITAL 

AREA. 

Consideration should be given to the following types of events when evaluating an event against the 

criteria of the site specific Security Contingency Plan: SABOTAGE and HOSTAGE / EXTORTION. 

The Safeguards Contingency Plan identifies numerous events/conditions that constitute a

threat/compromise to a Station’s security. Only those events that involve Actual or Likely Major 

failures of plant functions needed for protection of the public need to be considered. The following 

events would not normally meet this requirement; (e.g., Failure by a Member of the Security Force 

to carry out an assigned/required duty, internal disturbances, loss/compromise of safeguards 

materials or strike actions).

Loss of Plant Control would escalate this event to a GENERAL EMERGENCY.

Reference is made to (site-specific) security shift supervision because these individuals are the 

designated personnel on-site qualified and trained to confirm that a security event is occurring or 

has occurred. Training on security event classification confirmation is closely controlled due to the 

strict secrecy controls placed on the plant Security Plan.
Revision 01/2003 5-H-18

HAZARDS AND OTHER CONDITIONS

AFFECTING PLANT SAFETY

HS2

Initiating Condition – SITE AREA EMERGENCY

Control Room Evacuation Has Been Initiated and Plant Control Cannot Be Established.

Operating Mode Applicability: All

Example Emergency Action Level:

1. Control room evacuation has been initiated.

AND

Control of the plant cannot be established per (site-specific) procedure within (site-specific) 

minutes.

Basis:

Expeditious transfer of safety systems has not occurred but fission product barrier damage may 

not yet be indicated. The intent of this IC is to capture those events where control of the plant 

cannot be reestablished in a timely manner. Site-specific time for transfer based on analysis or 

assessments as to how quickly control must be reestablished without core uncovering and/or core 

damage. This time should not exceed 15 minutes without additional justification. The determination 

of whether or not control is established at the remote shutdown panel is based on Emergency 

Director (ED) judgment. The ED is expected to make a reasonable, informed judgment within the 

site-specific time for transfer that the licensee has control of the plant from the remote shutdown 

panel. 

The intent of the EAL is to establish control of important plant equipment and knowledge of

important plant parameters in a timely manner. Primary emphasis should be placed on those 

components and instruments that supply protection for and information about safety functions. 

Typically, these safety functions are reactivity control (ability to shutdown the reactor and maintain it 

shutdown), reactor water level (ability to cool the core), and decay heat removal (ability to maintain 

a heat sink) for a BWR. The equivalent functions for a PWR are reactivity control, RCS inventory, 

and secondary heat removal.

Escalation of this event, if appropriate, would be by Fission Product Barrier Degradation, Abnormal 

Rad Levels/Radiological Effluent, or Emergency Director Judgment ICs. 
Revision 01/2003 5-H-19

HAZARDS AND OTHER CONDITIONS

AFFECTING PLANT SAFETY

HS3

Initiating Condition – SITE AREA EMERGENCY

Other Conditions Existing Which in the Judgment of the Emergency Director Warrant 

Declaration of Site Area Emergency.

Operating Mode Applicability: All

Example Emergency Action Level:

1. Other conditions exist which in the judgment of the Emergency Director indicate that events 

are in process or have occurred which involve actual or likely major failures of plant functions 

needed for 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 EAL is intended to address 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 class description for Site Area Emergency.
Revision 01/2003 5-H-20

 HAZARDS AND OTHER CONDITIONS

AFFECTING PLANT SAFETY

HG1

Initiating Condition – GENERAL EMERGENCY

Security Event Resulting in Loss Of Physical Control of the Facility.

Operating Mode Applicability: All

Example Emergency Action Level:

1. A HOSTILE FORCE has taken control of plant equipment such that plant personnel are 

unable to operate equipment required to maintain safety functions.

Basis:

This IC encompasses conditions under which a HOSTILE FORCE has taken physical control of 

VITAL AREAs (containing vital equipment or controls of vital equipment) required to maintain safety 

functions and control of that equipment cannot be transferred to and operated from another

location. Typically, these safety functions are reactivity control (ability to shut down the reactor and 

keep it shutdown) reactor water level (ability to cool the core), and decay heat removal (ability to 

maintain a heat sink) for a BWR. The equivalent functions for a PWR are reactivity control, RCS 

inventory, and secondary heat removal. If control of the plant equipment necessary to maintain 

safety functions can be transferred to another location, then the above initiating condition is not met. 

This EAL should also address loss of physical control of spent fuel pool cooling systems if

imminent fuel damage is likely (e.g., freshly off-loaded reactor core in pool). 

Loss of physical control of the control room or remote shutdown capability alone may not prevent 

the ability to maintain safety functions per se. Design of the remote shutdown capability and the 

location of the transfer switches should be taken into account. 
Revision 01/2003 5-H-21

HAZARDS AND OTHER CONDITIONS

AFFECTING PLANT SAFETY

HG2

Initiating Condition – GENERAL EMERGENCY

Other Conditions Existing Which in the Judgment of the Emergency Director Warrant 

Declaration of General Emergency.

Operating Mode Applicability: All

Example Emergency Action Level:

1. Other conditions exist which in the judgment of the Emergency Director indicate that events 

are in process or have occurred which involve actual or imminent substantial core

degradation or melting with potential for loss of containment integrity. Releases can be

reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for 

more than the immediate site area.

Basis:

This EAL is intended to address 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 General Emergency class.
Revision 01/2003 5-S-1

Recognition Category S

System Malfunction

INITIATING CONDITION MATRIX

NOUE ALERT SITE AREA EMERGENCY GENERAL EMERGENCY

SU1 Loss of All Offsite Power to 

Essential Busses for Greater 

Than 15 Minutes.

Op. Modes: Power Operation, 

Startup, Hot Standby, Hot 

Shutdown

SA5 AC power capability to essential 

busses reduced to a single 

power source for greater than 15 

minutes such that any additional 

single failure would result in 

station blackout.

Op. Modes: Power Operation, 

Startup, Hot Standby, Hot 

Shutdown

SS1 Loss of All Offsite Power and 

Loss of All Onsite AC Power to

Essential Busses.

Op. Modes: Power Operation, 

Startup, Hot Standby, Hot 

Shutdown

SG1 Prolonged Loss of All Offsite 

Power and Prolonged Loss of 

All Onsite AC Power to Essential 

Busses.

Op. Modes: Power Operation, 

Startup, Hot Standby, Hot 

Shutdown

SA2 Failure of Reactor Protection 

System Instrumentation to Complete or Initiate an Automatic 

Reactor Scram Once a Reactor 

Protection System Setpoint Has 

Been Exceeded and Manual 

Scram Was Successful.

Op. Modes: Power Operation, 

Startup, Hot Standby 

SS2 Failure of Reactor Protection 

System Instrumentation to Complete or Initiate an Automatic 

Reactor Scram Once a Reactor 

Protection System Setpoint Has 

Been Exceeded and Manual 

Scram Was NOT Successful.

Op. Modes: Power Operation, 

Startup 

SG2 Failure of the Reactor Protection 

System to Complete an Automatic Scram and Manual Scram 

was NOT Successful and There 

is Indication of an Extreme 

Challenge to the Ability to Cool 

the Core.

Op. Modes: Power Operation, 

Startup

SU2 Inability to Reach Required 

Shutdown Within Technical 

Specification Limits.

Op. Modes: Power Operation, 

Startup, Hot Standby, Hot 

Shutdown

SA3 Deleted SS4 Complete Loss of Heat 

Removal Capability.

Op. Modes: Power Operation, 

Startup, Hot Standby, Hot 

Shutdown

SU3 UNPLANNED Loss of Most or All 

Safety System Annunciation or 

Indication in The Control Room 

for Greater Than 15 Minutes

Op. Modes: Power Operation, 

Startup, Hot Standby, Hot 

Shutdown

SA4 UNPLANNED Loss of Most or All 

Safety System Annunciation or 

Indication in Control Room With 

Either (1) a SIGNIFICANT 

TRANSIENT in Progress, or (2) 

Compensatory Non-Alarming 

Indicators are Unavailable.

Op. Modes: Power Operation, 

Startup, Hot Standby, Hot 

Shutdown

SS6 Inability to Monitor a 

SIGNIFICANT TRANSIENT in 

Progress.

Op. Modes: Power Operation, 

Startup, Hot Standby, Hot 

Shutdown
Revision 01/2003 5-S-2

Recognition Category S

System Malfunction

INITIATING CONDITION MATRIX

SU7 Deleted SA1 Deleted SS3 Loss of All Vital DC Power.

Op. Modes: Power Operation, 

Startup, Hot Standby, Hot 

Shutdown

SU4 Fuel Clad Degradation.

Op. Modes: Power Operation, 

Startup, Hot Standby, Hot 

Shutdown

SU5 RCS Leakage.

Op. Modes: Power Operation, 

Startup, Hot Standby, Hot 

Shutdown

SS5 Deleted

SU6 UNPLANNED Loss of All Onsite 

or Offsite Communications 

Capabilities. 

Op. Modes: Power Operation, 

Startup, Hot Standby, Hot 

Shutdown

SU8 Inadvertent Criticality.

Op Modes: Hot Standby, Hot 

Shutdown
Revision 01/2003 5-S-3

SYSTEM MALFUNCTION

SU1

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes.

Operating Mode Applicability: Power Operation

Startup

Hot Standby

Hot Shutdown

Example Emergency Action Level:

1. Loss of power to (site-specific) transformers for greater than 15 minutes.

AND

At least (site-specific) emergency generators are supplying power to emergency busses.

Basis:

Prolonged loss of AC power reduces required redundancy and potentially degrades the level of 

safety of the plant by rendering the plant more vulnerable to a complete Loss of AC Power (e.g., 

Station Blackout). Fifteen minutes was selected as a threshold to exclude transient or momentary 

power losses.

Plants that have the capability to cross-tie AC power from a companion unit may take credit for the 

redundant power source in the associated EAL for this IC. Inability to effect the cross-tie within 15 

minutes warrants declaring a NOUE.
Revision 01/2003 5-S-4

 SYSTEM MALFUNCTION

SU2

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

Inability to Reach Required Shutdown Within Technical Specification Limits.

Operating Mode Applicability: Power Operation

Startup

Hot Standby

Hot Shutdown

Example Emergency Action Level:

1. Plant is not brought to required operating mode within (site-specific) Technical Specifications 

LCO Action Statement Time.

Basis:

Limiting Conditions of Operation (LCOs) require the plant to be brought to a required shutdown 

mode when the Technical Specification required configuration cannot be restored. Depending on 

the circumstances, this may or may not be an emergency or precursor to a more severe condition. 

In any case, the initiation of plant shutdown required by the site Technical Specifications requires a 

one hour report under 10 CFR 50.72 (b) Non-emergency events. The plant is within its safety 

envelope when being shut down within the allowable action statement time in the Technical

Specifications. An immediate NOUE is required when the plant is not brought to the required 

operating mode within the allowable action statement time in the Technical Specifications.

Declaration of a NOUE is based on the time at which the LCO-specified action statement time 

period elapses under the site Technical Specifications and is not related to how long a condition 

may have existed. Other required Technical Specification shutdowns that involve precursors to 

more serious events are addressed by other System Malfunction, Hazards, or Fission Product 

Barrier Degradation ICs.
Revision 01/2003 5-S-5

SYSTEM MALFUNCTION

SU3

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

UNPLANNED Loss of Most or All Safety System Annunciation or Indication in The 

Control Room for Greater Than 15 Minutes

Operating Mode Applicability: Power Operation

Startup

Hot Standby

Hot Shutdown

Example Emergency Action Level:

1. UNPLANNED loss of most or all (site-specific) annunciators or indicators associated with 

safety systems for greater than 15 minutes.

Basis:

This IC and its associated EAL are intended to recognize the difficulty associated with monitoring 

changing plant conditions without the use of a major portion of the annunciation or indication 

equipment. 

Recognition of the availability of computer based indication equipment is considered (e.g., SPDS, 

plant computer, etc.).

Quantification of "Most" is arbitrary, however, it is estimated that if approximately 75% of the safety 

system annunciators or indicators are lost, there is an increased risk that a degraded plant

condition could go undetected. It is not intended that plant personnel perform a detailed count of the 

instrumentation lost but use the value as a judgment threshold for determining the severity of the 

plant conditions. 

It is further recognized that most plant designs provide redundant safety system indication powered 

from separate uninterruptable power supplies. While failure of a large portion of annunciators is 

more likely than a failure of a large portion of indications, the concern is included in this EAL due to 

difficulty associated with assessment of plant conditions. The loss of specific, or several, safety 

system indicators should remain a function of that specific system or component operability status. 

This will be addressed by the specific Technical Specification. The initiation of a Technical

Specification imposed plant shutdown related to the instrument loss will be reported via 10 CFR 

50.72. If the shutdown is not in compliance with the Technical Specification action, the NOUE is 

based on SU2 "Inability to Reach Required Shutdown Within Technical Specification Limits."

(Site-specific) annunciators or indicators for this EAL must include those identified in the Abnormal 

Operating Procedures, in the Emergency Operating Procedures, and in other EALs (e.g., area, 

process, and/or effluent rad monitors, etc.).

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

Due to the limited number of safety systems in operation during cold shutdown, refueling, and 

defueled modes, no IC is indicated during these modes of operation.
Revision 01/2003 5-S-6

This NOUE will be escalated to an Alert if a transient is in progress during the loss of annunciation 

or indication.
Revision 01/2003 5-S-7

SYSTEM MALFUNCTION

SU4

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

Fuel Clad Degradation.

Operating Mode Applicability: Power Operation

Startup

Hot Standby

Hot Shutdown

Example Emergency Action Levels: (1 or 2)

1. (Site-specific) radiation monitor readings indicating fuel clad degradation greater than

Technical Specification allowable limits.

2. (Site-specific) coolant sample activity value indicating fuel clad degradation greater than

Technical Specification allowable limits.

Basis:

This IC is included as a NOUE because it is considered to be a potential degradation in the level of 

safety of the plant and a potential precursor of more serious problems. EAL #1 addresses sitespecific radiation monitor readings such as BWR air ejector monitors, PWR failed fuel monitors, 

etc., that provide indication of fuel clad integrity. EAL #2 addresses coolant samples exceeding 

coolant technical specifications for iodine spike. Escalation of this IC to the Alert level is via the 

Fission Product Barrier Degradation Monitoring ICs. Though the referenced Technical Specification 

limits are mode dependent, it is appropriate that the EAL’s be applicable in all modes, as they 

indicate a potential degradation in the level of safety of the plant. The companion IC to SU4 for the 

Cold Shutdown/Refueling modes is CU5. 
Revision 01/2003 5-S-8

 SYSTEM MALFUNCTION

SU5

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

RCS Leakage.

Operating Mode Applicability: Power Operation

Startup

Hot Standby

Hot Shutdown

Example Emergency Action Levels: (1 or 2)

1. Unidentified or pressure boundary leakage greater than 10 gpm.

2. Identified leakage greater than 25 gpm.

Basis:

This IC is included as a NOUE because it may be a precursor of more serious conditions and, as 

result, is considered to be a potential degradation of the level of safety of the plant. The 10 gpm 

value for the unidentified and pressure boundary leakage was selected as it is observable with 

normal control room indications. Lesser values must generally be determined through timeconsuming surveillance tests (e.g., mass balances). The EAL for identified leakage is set at a 

higher value due to the lesser significance of identified leakage in comparison to unidentified or 

pressure boundary leakage. In either case, escalation of this IC to the Alert level is via Fission 

Product Barrier Degradation ICs.
Revision 01/2003 5-S-9

SYSTEM MALFUNCTION

SU6

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

UNPLANNED 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)

1. Loss of all (site-specific list) onsite communications capability affecting the ability to perform 

routine operations.

2. Loss of all (site-specific list) offsite communications capability. 

Basis:

The purpose of this IC and its associated EALs is to recognize a loss of communications capability 

that either defeats the plant operations staff ability to perform routine tasks necessary for plant 

operations or the ability to communicate problems with offsite authorities. The loss of offsite

communications ability is expected to be significantly more comprehensive than the condition 

addressed by 10 CFR 50.72.

The availability of one method of ordinary offsite communications is sufficient to inform state and 

local authorities of plant problems. This EAL is intended to be used only when extraordinary means 

(e.g., relaying of information from radio transmissions, individuals being sent to offsite locations, 

etc.) are being utilized to make communications possible. 

Site-specific list for onsite communications loss must encompass the loss of all means of routine 

communications (e.g., commercial telephones, sound powered phone systems, page party system 

(Gaitronics) and radios / walkie talkies).

Site-specific list for offsite communications loss must encompass the loss of all means of

communications with offsite authorities. This should include the ENS, commercial telephone lines, 

telecopy transmissions, and dedicated phone systems. 
Revision 01/2003 5-S-10

 SYSTEM MALFUNCTION

SU8

Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT

Inadvertent Criticality.

OPERATING MODE APPLICABILITY Hot Standby

Hot Shutdown

Example Emergency Action Level: (1 or 2)

1. An UNPLANNED extended positive period observed on nuclear instrumentation.

2. An UNPLANNED sustained positive startup rate observed on nuclear instrumentation.

Basis:

This IC addresses inadvertent criticality events. While the primary concern of this IC is criticality 

events that occur in Cold Shutdown or Refueling modes (NUREG 1449, Shutdown and Low-Power 

Operation at Commercial Nuclear Power Plants in the United States), the IC is applicable in other 

modes in which inadvertent criticalities are possible. This IC indicates a potential degradation of the 

level of safety of the plant, warranting a NOUE classification. This IC excludes inadvertent

criticalities that occur during planned reactivity changes associated with reactor startups (e.g., 

criticality earlier than estimated). The Cold Shutdown/Refueling IC is CU8.

This condition can be identified using period monitors/startup rate monitor. The term “extended“ is 

used in order to allow exclusion of expected short term positive periods/startup rates from planned 

control rod movements for PWRs and BWRs (such as shutdown bank withdrawal for PWRs). 

These short term positive periods/startup rates are the result of the increase in neutron population 

due to subcritical multiplication.

Escalation would be by the Fission Product Barrier Matrix, as appropriate to the operating mode at 

the time of the event, or by Emergency Director Judgment.
Revision 01/2003 5-S-11

 SYSTEM MALFUNCTION

SA2

Initiating Condition -- ALERT

Failure of Reactor Protection System Instrumentation to Complete or Initiate an 

Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been 

Exceeded and Manual Scram Was Successful.

Operating Mode Applicability: Power Operation

Startup

Hot Standby

Example Emergency Action Level:

1. Indication(s) exist that indicate that reactor protection system setpoint was exceeded and 

automatic scram did not occur, and a successful manual scram occurred.

Basis:

This condition indicates failure of the automatic protection system to scram the reactor. This 

condition is more than a potential degradation of a safety system in that a front line automatic 

protection system did not function in response to a plant transient and thus the plant safety has 

been compromised, and design limits of the fuel may have been exceeded. An Alert is indicated 

because conditions exist that lead to potential loss of fuel clad or RCS. Reactor protection system 

setpoint being exceeded, rather than limiting safety system setpoint being exceeded, is specified 

here because failure of the automatic protection system is the issue. A manual scram is any set of 

actions by the reactor operator(s) at the reactor control console which causes control rods to be 

rapidly inserted into the core and brings the reactor subcritical (e.g., reactor trip button, Alternate 

Rod Insertion). Failure of manual scram would escalate the event to a Site Area Emergency.
Revision 01/2003 5-S-12

SYSTEM MALFUNCTION

SA4

Initiating Condition -- ALERT

UNPLANNED Loss of Most or All Safety System Annunciation or Indication in Control 

Room With Either (1) a SIGNIFICANT TRANSIENT in Progress, or (2) Compensatory 

Non-Alarming Indicators are Unavailable.

Operating Mode Applicability: Power Operation

Startup

Hot Standby

Hot Shutdown

Example Emergency Action Level:

1. UNPLANNED loss of most or all (site-specific) annunciators or indicators associated with 

safety systems for greater than 15 minutes.

AND

Either of the following: (a or b)

a. A SIGNIFICANT TRANSIENT is in progress.

OR

b. Compensatory non-alarming indications are unavailable.

Basis:

This IC and its associated EAL are intended to recognize the difficulty associated with monitoring 

changing plant conditions without the use of a major portion of the annunciation or indication 

equipment during a transient. Recognition of the availability of computer based indication equipment 

is considered (e.g., SPDS, plant computer, etc.).

"Planned" loss of annunciators or indicators includes scheduled maintenance and testing activities.

Quantification of "Most" is arbitrary, however, it is estimated that if approximately 75% of the safety 

system annunciators or indicators are lost, there is an increased risk that a degraded plant

condition could go undetected. It is not intended that plant personnel perform a detailed count of the 

instrumentation lost but use the value as a judgment threshold for determining the severity of the 

plant conditions. It is also not intended that the Shift Supervisor be tasked with making a judgment 

decision as to whether additional personnel are required to provide increased monitoring of system 

operation.

It is further recognized that most plant designs provide redundant safety system indication powered 

from separate uninterruptable power supplies. While failure of a large portion of annunciators is 

more likely than a failure of a large portion of indications, the concern is included in this EAL due to 

difficulty associated with assessment of plant conditions. The loss of specific, or several, safety 
Revision 01/2003 5-S-13

system indicators should remain a function of that specific system or component operability status. 

This will be addressed by the specific Technical Specification. The initiation of a Technical

Specification imposed plant shutdown related to the instrument loss will be reported via 10 CFR 

50.72. If the shutdown is not in compliance with the Technical Specification action, the NOUE is 

based on SU2 "Inability to Reach Required Shutdown Within Technical Specification Limits."

Site-specific annunciators or indicators for this EAL must include those identified in the Abnormal 

Operating Procedures, in the Emergency Operating Procedures, and in other EALs (e.g., area, 

process, and/or effluent rad monitors, etc.).

"Compensatory non-alarming indications" in this context includes computer based information such 

as SPDS. This should include all computer systems available for this use depending on specific 

plant design and subsequent retrofits. If both a major portion of the annunciation system and all 

computer monitoring are unavailable, the Alert is required.

Due to the limited number of safety systems in operation during cold shutdown, refueling and 

defueled modes, no IC is indicated during these modes of operation.

This Alert will be escalated to a Site Area Emergency if the operating crew cannot monitor the 

transient in progress.
Revision 01/2003 5-S-14

SYSTEM MALFUNCTION

SA5

Initiating Condition -- ALERT

AC power capability to essential busses reduced to a single power source for greater 

than 15 minutes such that any additional single failure would result in station blackout.

Operating Mode Applicability: Power Operation

Startup

Hot Standby

Hot Shutdown

Example Emergency Action Level:

1. AC power capability to site-specific essential busses reduced to a single power source for 

greater than 15 minutes

AND

Any additional single failure will result in station blackout.

Basis:

This IC and the associated EALs are intended to provide an escalation from IC SU1, "Loss of All 

Offsite Power To Essential Busses for Greater Than 15 Minutes." The condition indicated by this IC 

is the degradation of the offsite and onsite power systems such that any additional single failure 

would result in a station blackout. This condition could occur due to a loss of offsite power with a 

concurrent failure of one emergency generator to supply power to its emergency busses. Another 

related condition could be the loss of all offsite power and loss of onsite emergency diesels with 

only one train of emergency busses being backfed from the unit main generator, or the loss of 

onsite emergency diesels with only one train of emergency busses being backfed from offsite 

power. The subsequent loss of this single power source would escalate the event to a Site Area 

Emergency in accordance with IC SS1, "Loss of All Offsite and Loss of All Onsite AC Power to 

Essential Busses."

At multi-unit stations, the EALs should allow credit for operation of installed design features, such 

as cross-ties or swing diesels, provided that abnormal or emergency operating procedures address 

their use. However, these stations must also consider the impact of this condition on other shared 

safety functions in developing the site specific EAL.
Revision 01/2003 5-S-15

 SYSTEM MALFUNCTION

SS1

Initiating Condition -- SITE AREA EMERGENCY

Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses.

Operating Mode Applicability: Power Operation

Startup

Hot Standby

Hot Shutdown

Example Emergency Action Level:

1. Loss of power to (site-specific) transformers.

AND

Failure of (site-specific) emergency generators to supply power to emergency busses.

AND

Failure to restore power to at least one emergency bus within (site-specific) minutes from the 

time of loss of both offsite and onsite AC power.

Basis:

Loss of all AC power compromises all plant safety systems requiring electric power including RHR, 

ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolonged loss of all AC power will 

cause core uncovering and loss of containment integrity, thus this event can escalate to a General 

Emergency. The (site-specific) time duration should be selected to exclude transient or momentary 

power losses, but should not exceed 15 minutes.

Escalation to General Emergency is via Fission Product Barrier Degradation or IC SG1, "Prolonged 

Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power."

Consideration should be given to operable loads necessary to remove decay heat or provide 

Reactor Vessel makeup capability when evaluating loss of AC power to essential busses. Even 

though an essential bus may be energized, if necessary loads (i.e., loads that if lost would inhibit 

decay heat removal capability or Reactor Vessel makeup capability) are not operable on the

energized bus then the bus should not be considered operable. If this bus was the only energized 

bus then a Site Area Emergency per SS1 should be declared.
Revision 01/2003 5-S-16

SYSTEM MALFUNCTION

SS2

Initiating Condition -- SITE AREA EMERGENCY

Failure of Reactor Protection System Instrumentation to Complete or Initiate an 

Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been 

Exceeded and Manual Scram Was NOT Successful.

Operating Mode Applicability: Power Operation

Startup

Example Emergency Action Level:

1. Indication(s) exist that automatic and manual scram were not successful.

Basis:

Automatic and manual scram are not considered successful if action away from the reactor control 

console was required to scram the reactor.

Under these conditions, the reactor is producing more heat than the maximum decay heat load for 

which the safety systems are designed. A Site Area Emergency is indicated because conditions 

exist that lead to imminent loss or potential loss of both fuel clad and RCS. Although this IC may be 

viewed as redundant to the Fission Product Barrier Degradation IC, its inclusion is necessary to 

better assure timely recognition and emergency response. Escalation of this event to a General 

Emergency would be via Fission Product Barrier Degradation or Emergency Director Judgment 

ICs.
Revision 01/2003 5-S-17

SYSTEM MALFUNCTION

SS3

Initiating Condition -- SITE AREA EMERGENCY

Loss of All Vital DC Power.

Operating Mode Applicability: Power Operation

Startup

Hot Standby

Hot Shutdown

Example Emergency Action Level:

1. Loss of All Vital DC Power based on (site-specific) bus voltage indications for greater than 15 

minutes.

Basis:

Loss of all DC power compromises ability to monitor and control plant safety functions. Prolonged 

loss of all DC power will cause core uncovering and loss of containment integrity when there is 

significant decay heat and sensible heat in the reactor system. Escalation to a General Emergency 

would occur by Abnormal Rad Levels/Radiological Effluent, Fission Product Barrier Degradation, or 

Emergency Director Judgment ICs. Fifteen minutes was selected as a threshold to exclude

transient or momentary power losses.
Revision 01/2003 5-S-18

SYSTEM MALFUNCTION

SS4

Initiating Condition -- SITE AREA EMERGENCY

Complete Loss of Heat Removal Capability.

Operating Mode Applicability: Power Operation

Startup

Hot Standby

Hot Shutdown

Example Emergency Action Level:

1. Loss of core cooling and heat sink (PWR).

1. Heat Capacity Temperature Limit Curve exceeded (BWR).

Basis:

This EAL addresses complete loss of functions, including ultimate heat sink, required for hot 

shutdown with the reactor at pressure and temperature. Reactivity control is addressed in other 

EALs. For BWRs the loss of heat removal function is indicated by the Heat Removal Capability 

Temperature Limit Curve being exceeded.

Under these conditions, there is an actual major failure of a system intended for protection of the 

public. Thus, declaration of a Site Area Emergency is warranted. Escalation to General Emergency 

would be via Abnormal Rad Levels / Radiological Effluent, Emergency Director Judgment, or

Fission Product Barrier Degradation ICs.
Revision 01/2003 5-S-19

 SYSTEM MALFUNCTION

SS6

Initiating Condition -- SITE AREA EMERGENCY

Inability to Monitor a SIGNIFICANT TRANSIENT in Progress.

Operating Mode Applicability: Power Operation

Startup

Hot Standby

Hot Shutdown

Example Emergency Action Level:

1. a. Loss of most or all (site-specific) annunciators associated with safety systems.

AND

b. Compensatory non-alarming indications are unavailable.

AND

c. Indications needed to monitor (site-specific) safety functions are unavailable.

AND

d. SIGNIFICANT TRANSIENT in progress.

Basis:

This IC and its associated EAL are intended to recognize the inability of the control room staff to 

monitor the plant response to a transient. A Site Area Emergency is considered to exist if the 

control room staff cannot monitor safety functions needed for protection of the public.

(Site-specific) annunciators for this EAL should be limited to include those identified in the Abnormal 

Operating Procedures, in the Emergency Operating Procedures, and in other EALs (e.g., rad 

monitors, etc.)

"Compensatory non-alarming indications" in this context includes computer based information such 

as SPDS. This should include all computer systems available for this use depending on specific 

plant design and subsequent retrofits.

 (Site-specific) indications needed to monitor safety functions necessary for protection of the public 

must include control room indications, computer generated indications and dedicated annunciation 

capability. The specific indications should be those used to determine such functions as the ability 

to shut down the reactor, maintain the core cooled, to maintain the reactor coolant system intact, 

and to maintain containment intact.

"Planned" and “UNPLANNED” actions are not differentiated since the loss of instrumentation of this 

magnitude is of such significance during a transient that the cause of the loss is not an ameliorating 

factor.
Revision 01/2003 5-S-20

Quantification of "Most" is arbitrary, however, it is estimated that if approximately 75% of the safety 

system annunciators or indicators are lost, there is an increased risk that a degraded plant

condition could go undetected. It is not intended that plant personnel perform a detailed count of the 

instrumentation lost but use the value as a judgment threshold for determining the severity of the 

plant conditions. It is also not intended that the Shift Supervisor be tasked with making a judgment 

decision as to whether additional personnel are required to provide increased monitoring of system 

operation.
Revision 01/2003 5-S-21

SYSTEM MALFUNCTION

SG1

Initiating Condition -- GENERAL EMERGENCY

Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power to 

Essential Busses.

Operating Mode Applicability: Power Operation

Startup

Hot Standby

Hot Shutdown

Example Emergency Action Level:

1. Loss of power to (site-specific) transformers.

AND

Failure of (site-specific) emergency diesel generators to supply power to emergency busses.

AND

Either of the following: (a or b)

a. Restoration of at least one emergency bus within (site-specific) hours is not likely

OR

b. (Site-Specific) Indication of continuing degradation of core cooling based on Fission 

Product Barrier monitoring.

Basis:

Loss of all AC power compromises all plant safety systems requiring electric power including RHR, 

ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolonged loss of all AC power will 

lead to loss of fuel clad, RCS, and containment. The (site-specific) hours to restore AC power can 

be based on a site blackout coping analysis performed in conformance with 10 CFR 50.63 and 

Regulatory Guide 1.155, "Station Blackout," as available. Appropriate allowance for offsite

emergency response including evacuation of surrounding areas should be considered. Although 

this IC may be viewed as redundant to the Fission Product Barrier Degradation IC, its inclusion is 

necessary to better assure timely recognition and emergency response.

This IC is specified to assure that in the unlikely event of a prolonged station blackout, timely 

recognition of the seriousness of the event occurs and that declaration of a General Emergency 

occurs as early as is appropriate, based on a reasonable assessment of the event trajectory.

The likelihood of restoring at least one emergency bus should be based on a realistic appraisal of 

the situation since a delay in an upgrade decision based on only a chance of mitigating the event 

could result in a loss of valuable time in preparing and implementing public protective actions.
Revision 01/2003 5-S-22

In addition, under these conditions, fission product barrier monitoring capability may be degraded. 

Although it may be difficult to predict when power can be restored, it is necessary to give the 

Emergency Director a reasonable idea of how quickly (s)he may need to declare a General

Emergency based on two major considerations:

1. Are there any present indications that core cooling is already degraded to the point that Loss 

or Potential Loss of Fission Product Barriers is imminent? (Refer to Tables 3 and 4 for more 

information.)

2. If there are no present indications of such core cooling degradation, how likely is it that power 

can be restored in time to assure that a loss of two barriers with a potential loss of the third 

barrier can be prevented?

Thus, indication of continuing core cooling degradation must be based on Fission Product Barrier 

monitoring with particular emphasis on Emergency Director judgment as it relates to imminent 

Loss or Potential Loss of fission product barriers and degraded ability to monitor fission product 

barriers.
Revision 01/2003 5-S-23

SYSTEM MALFUNCTION

SG2

Initiating Condition -- GENERAL EMERGENCY

Failure of the Reactor Protection System to Complete an Automatic Scram and Manual 

Scram was NOT Successful and There is Indication of an Extreme Challenge to the 

Ability to Cool the Core.

Operating Mode Applicability: Power Operation

Startup

Example Emergency Action Level:

1. Indications exist that automatic and manual scram were not successful.

AND

Either of the following: (a or b)

a. Indication(s) exists that the core cooling is extremely challenged.

OR

b. Indication(s) exists that heat removal is extremely challenged.

Basis:

Automatic and manual scram are not considered successful if action away from the reactor control 

console is required to scram the reactor.

Under the conditions of this IC and its associated EALs, the efforts to bring the reactor subcritical 

have been unsuccessful and, as a result, the reactor is producing more heat than the maximum 

decay heat load for which the safety systems were designed. Although there are capabilities away 

from the reactor control console, such as emergency boration in PWRs, or standby liquid control in 

BWRs, the continuing temperature rise indicates that these capabilities are not effective. This 

situation could be a precursor for a core melt sequence.

For PWRs, the extreme challenge to the ability to cool the core is intended to mean that the core 

exit temperatures are at or approaching 1200 degrees F or that the reactor vessel water level is 

below the top of active fuel. For plants using CSFSTs, this EAL equates to a Core Cooling RED 

condition and an entry into function restoration procedure FR-S.1. For BWRs, the extreme

challenge to the ability to cool the core is intended to mean that the reactor vessel water level 

cannot be restored and maintained above Minimum Steam Cooling RPV Water Level as described 

in the EOP bases.

Another consideration is the inability to initially remove heat during the early stages of this

sequence. For PWRs, if emergency feedwater flow is insufficient to remove the amount of heat 

required by design from at least one steam generator, an extreme challenge should be considered 

to exist. For plants using CSFSTs, this EAL equates to a Heat Sink RED condition. For BWRs, 
Revision 01/2003 5-S-24

considerations include inability to remove heat via the main condenser, or via the suppression pool 

or torus (e.g., due to high pool water temperature).

In the event either of these challenges exist at a time that the reactor has not been brought below 

the power associated with the safety system design (typically 3 to 5% power) a core melt sequence 

exists. In this situation, core degradation can occur rapidly. For this reason, the General Emergency 

declaration is intended to be anticipatory of the fission product barrier matrix declaration to permit 

maximum offsite intervention time.
Revision 01/2003 A.1

Appendix A

Basis for Radiological Effluent Initiating Conditions

Introduction

This appendix supplements the basis information provided in Section 5 for initiating conditions AU1, 

AA1, AS1, and AG1. Since the publication of revision 2 of this methodology, there have been 

numerous questions raised as utilities worked to implement the IC and EALs. Additional feedback 

was provided by the staff of the Nuclear Regulatory Commission. It became apparent that the brief 

basis provided for each IC was not sufficient. When revision 3 of this document was in preparation, 

it was decided to incorporate this appendix to provide the needed additional guidance and

clarification. The NUMARC/NESP-007 effluent IC/EALs represent a departure from previous EAL 

practice and understanding these differences and their technical bases will facilitate site specific 

implementation of the NUMARC/NESP-007 classification methodology.

This appendix will be structured into seven major sections. They are:

1. Purpose of the effluent ICs/EALs and their relationship to other ICs/EALs

2. Explanation of the ICs

3. Explanation of the example EALs and their relationship to the ICs 

4. Interface between the ICs/EALs and the Offsite Dose Calculation Manual (ODCM)

5. Monitor setpoints versus EAL thresholds.

6. The impact of meteorology

7. The impact of source term

A.1 Purpose of the Effluent ICs/EALs

ICs AU1, AA1, AS1, and AG1 provide classification thresholds for UNPLANNED and/or uncontrolled 

releases of radioactivity to the environment. In as much as the purpose of emergency planning at 

nuclear power plants is to minimize the consequences of radioactivity releases to the environment, 

these ICs would appear to be controlling. However, classification of emergencies on the basis of 

radioactivity releases is not optimum, particularly those classifications based on radiation monitor 

indications. Such classifications can be deficient for several reasons, including:

• In significant emergency events, a radioactivity release is seldom the initiating event, but 

rather, is the consequence of some other condition. Relying on an indication of a release 

may not be sufficiently anticipatory.

• The relationship between an effluent monitor indication caused by a release and the 

offsite conditions that result is a function of several parameters (e.g., meteorology, 

source term) which can change in value by orders of magnitude between normal and 

emergency conditions and from event to event. The appropriateness of these

classifications is dependent on how well the parameter values assumed in preestablishing the classification thresholds match those that are present at the time of the 

incident.

Section 3.3 of NUMARC/NESP-007 emphasizes the need for accurate assessment and

classification of events, recognizing that over-classification, as well as under-classification, is to be 

avoided. Primary emphasis is intended to be placed on plant conditions in classifying emergency 

events. Effluent ICs were included, however, to provide a basis for classifying events that cannot be 

readily classified on the basis of plant condition alone. Plant condition ICs are included to address 

the precursors to radioactivity release in order to ensure anticipatory action. The effluent ICs do not 
Revision 01/2003 A.2

stand alone, nor do the plant condition ICs. The inclusion of both categories more fully addresses 

the potential event spectrum and compensates for potential deficiencies in either. This is a case in 

which the whole is greater than the sum of the parts.

From the discussion that follows, it should become clear how the various aspects of the

NUMARC/NESP-007 effluent ICs/EALs work together to provide for reasonably accurate and timely 

emergency classifications. While some aspects of the radiological effluent EALs may appear to be 

potentially unconservative, one also needs to consider IC/EALs in other recognition categories that 

compensate for this condition. During site specific implementation of these ICs/EALs, changes to 

some of these aspects might appear advantageous. While site specific changes are anticipated, 

caution must be used to ensure that these changes do not impact the overall effectiveness of the 

ICs / EALs.

A.2. Initiating Conditions

There are four radiological effluent ICs provided in NUMARC/NESP-007. The IC and the

fundamental basis for the ultimate classification for the four classifications are:

General (AG1) Offsite Dose Resulting from an Actual or Imminent Release of Gaseous 

Radioactivity Exceeds 1000 mR TEDE or 5000 mR Thyroid CDE for the 

Actual or Projected Duration of the Release Using Actual Meteorology.

Site Area (AS1) Offsite Dose Resulting from an Actual or Imminent Release of Gaseous 

Radioactivity Exceeds 100 mR TEDE or 500 mR Thyroid CDE for the 

Actual or Projected Duration of the Release.

Alert (AA1) Any UNPLANNED Release of Gaseous or Liquid Radioactivity to the

Environment that Exceeds 200 Times Radiological Technical

Specifications for 15 Minutes or Longer.

NOUE (AU1) Any UNPLANNED Release of Gaseous or Liquid Radioactivity to the

Environment that Exceeds Two Times Radiological Technical

Specifications for 60 Minutes or Longer.

The fundamental basis of AU1 and AA1 ICs differs from that for AS1 and AG1 ICs. It is important to 

understand the differences. 

• The Radiological Effluent Technical Specifications (RETS) (similar controls are included 

in the ODCMs of those facilities that implemented Generic Letter 89-01) are associated 

with particular offsite doses and dose rate limits. For showing compliance with these 

limits, facility Offsite Dose Calculation Manuals (ODCM) establish methodologies for 

establishing effluent monitor alarm setpoints, based on defined source term and

meteorology assumptions.

• AU1 and AA1 are NOT based on these particular values of offsite dose or dose rate but, 

rather, on the loss of plant control implied by a radiological release that exceeds a 

specified multiple of the RETS release limits for a specified period of time. 

• The RETS multiples are specified only to distinguish AU1 and AA1 from non-emergency 

conditions and from each other. While these multiples obviously correspond to an offsite 

dose, the classification emphasis is on a release that does not comply with a license 

commitment for an extended period of time. 

• While some of the example EALs for AU1 and AA1 use indications of offsite dose rates 

as symptoms that the RETS may be exceeded, the IC, and the classification, are NOT

concerned with the particular value of offsite dose. While there may be quantitative 

inconsistencies involved with this protocol, the qualitative basis of the EAL, i.e., loss of 

plant control, is not affected. 
Revision 01/2003 A.3

• The basis of the AS1 and AG1 ICs IS a particular value of offsite dose for the event 

duration. AG1 is set to the value of the EPA PAG. AS1 is a fraction (10%) of the EPA 

PAG. As such, these ICs are consistent with the fundamental definitions of a Site Area 

and General Emergency.

A.3 Example Emergency Action Levels

For each of the classifications, NUMARC/NESP-007 provides some example emergency action 

levels and bases. Ideally, the example EALs would correspond numerically with the thresholds 

expressed in the respective IC. Two cases are applicable to the effluent EALs:

1. The EAL corresponds numerically to the threshold in the respective IC. For example, a 

field survey result of 1000 mrem/hr for a projected release duration of one hour 

corresponds directly to AG1.

2. The EAL corresponds numerically to the threshold in the respective IC under certain 

assumed conditions. For example, an effluent monitor reading that equates to 100 

mrem for the projected duration of the release corresponds numerically to AS1 if the 

actual meteorology, source term, and release duration matches that used in

establishing the monitor thresholds. 

There are four typical example EALs:

• Effluent Monitor Readings: These EALs are pre-calculated values that correspond to the 

condition identified in the IC for a given set of assumptions. 

• Field Survey Results: These example EALs are included to provide a means to address 

classifications based on results from field surveys. 

• Perimeter Monitor Indications: For sites having them, perimeter monitors can provide a 

direct indication of the offsite consequences of a release. 

• Dose Assessment Results: These example EALs are included to provide a means to 

address classifications based on dose assessments. 

A.3.1 Effluent Monitor Readings

As noted above, these EALs are pre-calculated values that correspond to the condition identified in 

the IC for a given set of assumptions. The degree of correlation is dependent on how well the 

assumed parameters (e.g., meteorology, source term, etc.) represent the actual parameters at the 

time of the emergency. 

AS1 and AG1

Classifications should be made under these EALs if VALID (e.g., channel check, comparison to 

redundant/diverse indication, etc.) effluent radiation monitor readings exceed the pre-calculated 

thresholds. In a change from previous versions of this methodology, confirmation by dose

assessments is no longer required as a prerequisite to the classification. Nonetheless, dose

assessments are important components of the overall accident assessment activities when

significant radioactivity releases have occurred or are projected. Dose assessment results, when 

they become available, may serve to confirm the validity of the effluent radiation monitor EAL, may 

indicate that an escalation to a higher classification is necessary, or may indicate that the

classification wasn’t warranted. AS1 and AG1 both provide that, if dose assessment results are 

available, the classification should be based on the basis of the dose assessment result rather than 

the effluent radiation monitor EAL.

AU1 and AA1

ODCMs provide a methodology for determining default and batch-specific effluent monitor alarm 

setpoints pursuant to Standard Technical Specification (STS) 3.3.3.9. These setpoints are intended 
Revision 01/2003 A.4

to show that releases are within STS 3.11.2.1. The applicable limits are 500 mrem/year whole body 

or 3000 mrem/year skin from noble gases. (Inhalation dose rate limits are not addressed here since 

the specified surveillance involves collection and analysis of composite samples. This after-the-fact 

assessment could not be an made in a timely manner conducive to accident classification.) These 

setpoints are calculated using default source terms or batch-specific sample isotopic results and 

annual average χ/Q. Since the meteorology data is pre-defined, there is a direct correlation 

between the monitor setpoints and the RETS limits. Although the actual χ/Q may be different, 

NUREG-1022, Event Reporting Guidelines 10 CFR 50.72 and 50.73, provided "...Annual average 

meteorological data should be used for determining offsite airborne concentrations of radioactivity 

to maintain consistency with the technical specifications (TS) for reportability thresholds." The 

ODCM methodology is based on long term continuous releases. However, its use here in a short 

term release situation is appropriate. Remember that the AU1 and AA1 ICs are based on a loss of 

plant control indicated by the failure to comply with a multiple of the RETS release limits for an 

extended period and that the ODCM provides the methodology for showing compliance with the 

RETS.

To obtain the EAL thresholds, multiply the ODCM setpoint for each monitor by 2 (AU1) or 200 (AA1). 

It would be preferable to reference "2 x ODCM Setpoint" or "200 x ODCM Setpoint" as the EAL 

threshold. In this manner, the EAL would always change in step with changes in the ODCM setpoint 

(e.g., for a batch or special release. In actual practice, there may be an "warning" and a "high" alarm 

setpoint. The setpoint that is closest in value to the RETS limit should be used. Facility ODCMs 

may lower the actual setpoint to provide an administrative "safety margin". Also, if there is more 

than one unit or release stack on the site, the RETS limits may be apportioned. Two possible 

approaches to obtain the EAL thresholds are:

• The "2x" and "200x" multiples could be increased to address the reduced setpoints. For 

example, if the stack monitor were set to 50% of the RETS limit, the EAL threshold could 

be set to "4x" and "400x" the setpoint on that monitor. 

• The reduced setpoints could be ignored and the "2x" and "200x" multiples used as 

specified. While numerically conservative, using a single set of multipliers would

probably be desirable from a human engineering standpoint.

In a change from previous versions of this methodology, confirmation by dose assessments is no 

longer required as a prerequisite to the classification. While assessments with real meteorology 

may have provided a basis for escalating to AS1 (or AG1), the assessments could not confirm the 

AU1 or AA1 classifications since compliance with the RETS is demonstrated using annual average

meteorology – not – actual meteorology. 

Nonetheless, dose assessments are important components of the overall accident assessment 

activities when significant radioactivity releases have occurred or are projected. Dose assessment 

results, when they become available, may indicate that an escalation to a higher classification is 

necessary. AS1 and AG1 both provide that, if dose assessment results are available, the

classification should be based on the basis of the dose assessment result rather than the effluent 

radiation monitor EAL.

In typical practice, the radiological effluent monitor alarms would have been set, on the basis of 

ODCM requirements, to indicate a release that could exceed the RETS limits. Alarm response 

procedures call for an assessment of the alarm to determine whether or not RETS have been 

exceeded. Utilities typically have methods for rapidly assessing an abnormal release in order to 

determine whether or not the situation is reportable under 10 CFR 50.72. Since a radioactivity 

release of a magnitude comparable to the RETS limits will not create a need for offsite protective 

measures, it would be reasonable to use these abnormal release assessment methods to initiate 

dose assessment techniques using actual meteorology and projected source term and release 

duration.
Revision 01/2003 A.5

A.3.2 Perimeter Monitor, Field Survey Results, Dose Projection Results

AS1 and AG1

The perimeter monitor and field survey results are included to provide a means for classification 

based on actual measurements. There is a 1:1 correlation (with consideration of release duration) 

between these EALs and the IC since all are dependent on actual meteorology.

Dose projection result EALs are included to provide a basis for classification based on results from 

assessments triggered at lower emergency classifications. If the dose assessment results are 

available at the time that the classification is made, the results should be used in conjunction with 

this EAL for classifying the event rather than the effluent radiation monitor EAL.

Although the IC references TEDE and thyroid CDE as criteria, field survey results and perimeter 

monitor indications will generally not be reported in these dose quantities, but rather in terms of a 

dose rate. For this reason, the field survey EALs are based on a β-γ dose rate and a thyroid CDE 

value, both assuming one hour of exposure (or inhalation). If individual site analyses indicate a 

longer or shorter duration for the period in which the substantial portion of the activity is released, 

the longer duration should be used for the field survey and/or perimeter monitor EALs.

AU1 and AA1

As discussed previously, the threshold in these ICs is based on exceeding a multiple of the RETS 

for an extended period. The applicable RETS limit is the instantaneous dose rate provided in 

Standard Technical Specification (STS) 3.11.2.1. While these three EALs are also expressed in 

dose rate, they are dependent on actual meteorology. However, compliance with the RETS is 

demonstrated using annual average meteorology. Due to this, the only time that there would be a 

1:1 correlation between the IC and these EALs is when the value of the actual meteorology

matched the annual average -- an unlikely situation. For this reason, these EALs can only be 

indirect indicators that the RETS may be exceeded. The three example EALs are consistent with 

the fundamental basis of AU1 and AA1, that of a uncontrolled radioactivity release that indicates a 

loss of plant control. A dose rate, at or beyond the site boundary, greater than 0.1 mR/hr for 60 

minutes or 10.0 mR/hr for 15 minutes is consistent with this fundamental basis, regardless of the 

lack of numerical correlation to the RETS. The time periods chosen for the NOUE AU1 (60 minutes) 

and Alert AA1 (15 minutes) are indicative of the relative risks based on the loss of ability to terminate 

a release.

The numeric values shown in AU1 and AA1 are based on a release rate not exceeding 500 mrem 

per year, converted to a rate of: 500 ÷ 8766 = 0.057 mR/hr. If we take a multiple of 2, as specified in 

the NOUE threshold, this equates to a dose rate of about 0.11 mR/hr, which rounds to the 0.1 

mR/hr specified in AU1. Similarly for the AA1 EALs, we obtain 10 mR/hr. 

In AU1 and AA1, reference is made to automatic real-time dose assessment capability. In AS1 and 

AG1, the reference is to dose assessment. This distinction was made since it is unlikely that a dose 

assessment using manual methods would be initiated without some prior indication, e.g., a effluent 

monitor EAL. 

A.4 Interface Between ODCM and ICs/EALs

For AU1 and AA1, a strong link was established with the facility's ODCM. It was the intent of the 

NUMARC/NESP EAL Task Force to have the AU1 and AA1 EALs indexed to the ODCM alarm 

setpoints. This was done for several reasons:

• To allow the EALs to use the monitor setpoints already in place in the facility ODCM, 

thus eliminating the need for a second set of values as the EALs. The EAL could 

reference "2x ODCM Setpoint" or "200x ODCM Setpoint" for the monitors addressed in 
Revision 01/2003 A.6

the ODCM. Extensive calculations would only be necessary for monitors not addressed 

in the ODCM.

• To take advantage of the alarm setpoint calculational methodology already documented 

in the facility ODCM.

• To ensure that the operators had an alarm to indicate the abnormal condition. If the 

monitor EAL threshold was less than the default ODCM setpoint, the operators could be 

in the position of having exceeded an EAL and not knowing it.

• To simplify the IC/EAL by eliminating the need to address planned and UNPLANNED 

releases, continuous or batch releases, monitored or unmonitored releases. Any release 

that complies with the radiological effluent technical specifications (RETS) (or ODCM 

controls for utilities that have implemented GL 89-01) would not exceed a monitor EAL 

threshold.

• To eliminate the possibility of a planned release (e.g., containment / drywell purge) 

resulting in effluent radiation monitor readings that exceed an classification threshold 

that was based on a different calculation method. ODCMs typically require specific 

alarm setpoints for such releases. If the release can be authorized under the provisions 

of the ODCM/RETS, an emergency classification is not warranted. If the monitor EAL 

threshold is indexed to the ODCM setpoint (e.g., "...2 x ODCM setpoint...") the monitor 

EAL will always change in step with the ODCM setpoint.

• Although the ODCM is intended to address long term routine releases, its use here for 

short term releases is appropriate. The IC is specified in terms of a release that exceeds 

RETS for an extended period of time. Compliance to the RETS is shown using the 

ODCM methodology.

A.5 Setpoints versus Monitor EALs

Effluent monitors typically have provision for two separate alarm setpoints associated with the level 

of measured radioactivity. (There may be other alarms for parameters such as low sample flow.) 

These setpoints are typically established by the facility ODCM. As such, at most sites the values of 

the monitor EAL thresholds will not be implemented as actual alarm setpoints, but would be 

tabulated in the classification procedure. If the monitor EAL thresholds are calculated as suggested 

herein they will be higher than the ODCM alarm setpoints by at least a factor of two (i.e., AU1). This 

alarm alerts the operator to compare the monitor indication to the EAL thresholds. The

NUMARC/NESP-007 effluent EALs do NOT require alarm setpoints based on the monitor EALs. 

However, if spare alarm channels are available (e.g., high range channels), the monitor EAL

threshold could be used as the alarm setpoint.

A.6 The Impact of Meteorology

The existence of uncertainty between actual event meteorology and the meteorology assumed in 

establishing the EALs was identified above. It is important to note that uncertainty is present 

regardless of the meteorology data set assumed. The magnitude of the potential difference and, 

hence, the degree of conservatism will depend on the data set selected. Data sets that are intended 

to ensure low probability of under-conservative assessments have a high probability of being overconservative. For nuclear power plants, there are different sets of meteorological data used for 

different purposes. The two primary sets are:

• For accident analyses purposes, sector χ/Q values are set at that value that is 

exceeded only 0.5% of the hours wind blows into the sector. The highest of the 16 

sector values is the maximum sector χ/Q value. The site χ/Q value is set at that value 
Revision 01/2003 A.7

that is exceeded only 5% of the hours for all sectors. The higher of the sector or site χ/Q 

values is used in accident analyses.

• For routine release situations, annual average χ/Q values are calculated for specified 

receptor locations and at standard distances in each of the 16 radial sectors. In setting 

ODCM alarm set points, the annual average χ/Q value for the most restrictive receptor 

at or beyond the site boundary is used. The sector annual average χ/Q value is 

normalized for the percentage of time that the wind blows into that sector. In an actual 

event, the wind direction may be into the affected sector for the entire release duration. 

Many sites experience typical sector χ/Qs that are 10-20 times higher than the

calculated annual average for the sector.

In developing the effluent EALs, the NUMARC EAL Task Force elected to use annual average 

meteorology for establishing effluent monitor EAL thresholds. This decision was based on the 

following considerations.

• Use of the accident χ/Qs, may be too conservative. For some sites, the difference 

between the accident χ/Q and the annual average χ/Q can be a factor of 100-1000. With 

this difference in magnitude, the calculated monitor EALs for AS1 or AG1 might actually 

be less than the ODCM alarm setpoints, resulting in unwarranted classifications for 

releases that might be in compliance with ODCM limits.

• The ODCM and the RETS are based in part on annual average χ/Q (non-normalized). 

ODCMs already provide alarm setpoints based on annual average χ/Q that could be 

used for AU1 and AA1.

• Use of a χ/Q more restrictive than the χ/Q used to establish ODCM alarm setpoints 

could create a situation in which the EAL value would be less than the ODCM setpoint. 

In this case, the operators would have no alarm indication to alert them of the

emergency condition.

• Use of one χ/Q value for AU1 and AA1 and another for AS1 and AG1 might result in 

monitor EALs that would not progress from low to high classifications. Instead, the AS1 

and AA1 EALs might overlap.

Plant specific consideration must be made to determine if annual average meteorology is

adequately conservative for site specific use. If not one of the two more conservative techniques 

described above should be selected. It is incumbent upon the licensee to ensure that the selection 

is properly implemented to provide consistent classification escalation.

The impact of the differences between the assumed annual average meteorology and the actual 

meteorology depends on the particular EAL.

• For the AU1 and AA1 effluent monitor EALs, there is no impact since the IC and the 

EALs are based on annual average meteorology by definition.

• For the field survey, perimeter monitor, and dose assessment results EALs in AS1 and 

AG1, there is no impact since the IC and these EALs are based on actual meteorology.

• For the AS1 and AG1 effluent monitor EALs, there may be differences since the IC is 

based on actual meteorology and the monitor EALs are calculated on the basis of 

annual average meteorology or, on a site specific basis, one of the more conservative 

derivatives of annual average meteorology. This is considered as acceptable in that 

dose assessments using actual meteorology will be initiated for significant radioactivity 

releases. Needed escalations can be based on the results of these assessments. As 
Revision 01/2003 A.8

discussed previously, this delay was deemed to be acceptable since in significant

release situations, the plant condition EALs should provide the anticipatory

classifications necessary for the implementation of offsite protective measures.

• For the field survey, perimeter monitor, and dose assessment results EALs in AU1 and 

AA1, there is an impact. These three EALs are dependent on actual meteorology.

However, the threshold values for all of the AU1 and AA1 EALs are based on the 

assumption of annual average meteorology. If the actual and annual average

meteorology were equal, the IC and all of the EALs would correlate. Since it is likely that 

the actual meteorology will exceed the annual average meteorology, there will be

numerical inconsistencies between these EALs and the IC. The three example EALs are 

consistent with the fundamental basis of AU1 and AA1, that of a uncontrolled

radioactivity release that indicates a loss of plant control. A dose rate, at or beyond the 

site boundary, greater than 0.1 mR/hr for 60 minutes or 10.0 mR/hr for 15 minutes is 

consistent with this fundamental basis, regardless of the lack of numerical correlation to 

the RETS.

A.7 The Impact of Source Term

The ODCM methodology should be used for establishing the monitor EAL thresholds for these ICs. 

The ODCM provides a default source term based on expected releases. In many cases, the ODCM 

source term is derived from expected and/or design releases tabulated in the FSAR. 

For AS1 and AG1, the bases suggests the use of the same source terms used for establishing 

monitor EAL thresholds for AU1 and AA1. This guidance is provided to avoid potential overlaps 

between effluent monitor EALs for AA1 and AS1. Other source terms may be appropriate. In any 

case, efforts should be made to obtain and use best estimate (For Example: NUREG 1465), as 

opposed to conservative, source terms for all four ICs.

Even if the same source term is used for all four ICs, the analyst must consider the impact of overly 

conservative iodine to noble gas ratios. The AU1 and AA1 IC thresholds are based on external noble 

gas exposure. The AS1 and AG1 ICs are based on either TEDE or thyroid CDE. TEDE includes a 

contribution from inhalation exposure (i.e., CEDE) while the thyroid CDE is due solely to inhalation 

exposure. The inhalation exposure is sensitive to the iodine concentration in the source term. Since 

AU1 and AA1 are based on noble gases, and AS1 and AG1 are dependent on noble gases and

iodine, an over conservative iodine to noble gas ratio could result in AS1 and AG1 monitor EAL 

thresholds that either overlap or are too close to the AA1 monitor EAL thresholds.

As with meteorology, assessment of source terms has uncertainty. This uncertainty is

compensated for by the anticipatory classifications provided by ICs in other recognition categories.
Revision 01/2003 B.1

Appendix B

Basis For Implementation of Category C, D, AND E

Initiating Conditions By NUREG-0654/FEMA-REP-1 Users

NUMARC/NESP-007, Revision 2 (January 1992), was originally developed and approved as an 

alternative emergency action level (EAL) methodology that could be used in lieu of NUREG-0654 

Appendix 1 example initiating conditions (ICs). Regulatory Guide 1.101, “Emergency Planning and 

Preparedness for Nuclear Power Plants,” Revision 3, allowed licensees to continue using

previously approved ICs/EALs based on NUREG-0654 or submit for approval ICs/EALs based on 

NUMARC/NESP-007 to satisfy provisions of 10 CFR 50.47 (b) (4) and Section IV.B of Appendix E, 

“Emergency Planning and Preparedness for Production and Utilization Facilities,” to 10 CFR Part 

50.

In 1994, the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of 

Radiation Safety and Safeguards, Emergency Preparedness Branch developed a Branch Technical 

Position paper to support the NRC regional staffs’ technical review of EAL changes proposed by 

licensees who had not fully implemented the guidance in NUMARC/NESP-007. This paper provided 

examples of acceptable changes licensees could make to site-specific EALs developed from 

NUREG-0654 utilizing the technical bases under the example EALs in NUMARC/NESP-007. The 

NUREG-0654 example initiating conditions addressed by this position paper were Site Area

Emergency #9, Alerts #9 and #11, and Notification of Unusual Events #1, #4, #8, #9, #11 (partially), 

and #15. It was noted that licensees could provide the NRC other changes that utilized

NUMARC/NESP-007 guidance for evaluation on their individual merits and their complement to the 

licensee’s classification scheme as a whole. Note that the Branch Technical position was

subsequently incorporated into EPPOS 1.

During development of NEI 97-03 Revision 3 it was determined enhancements to address 

classification of events that occur during periods of plant shutdowns and refueling were needed. 

Additionally the need for EALs for permanently defueled sites and ISFSIs were identified. The 

decision was made to defer these enhancements for inclusion in NEI 99-01 Revision 4. These 

ICs/EALs were fashioned so they could be implemented throughout the industry, i.e., by licensees 

with EALs based on either NUMARC/NESP-007 or NUREG-0654 example ICs. NUREG-0654-

based EAL users may develop site-specific thresholds by using the bases contained in Appendix C, 

Basis for Cold Shutdown/Refueling ICs, Appendix D, Basis for Permanently Defueled Stations ICs, 

and Appendix E, Basis for ISFSI ICs. 

Adjustments to existing NUREG-0654-based EAL sets may be warranted coincident with the 

implementation of these site-specific Recognition Category C, D and/or E ICs/EALs. For example, 

it may be appropriate to define (if not previously defined) or redefine the mode applicability of an EAL 

based on NUREG-0654 Example IC General Emergency #2 (Loss of 2 of 3 Fission Product 

Barriers with potential loss of 3RD barrier) when adding the CG1 IC/EAL. Application of this approach 

may eliminate potential mode applicability conflicts or overlaps in existing and new IC/EALs. It is not 

the intent of this document to require NUREG-0654 users to adopt mode applicability or implement 

Recognition Category C in its entirety. 

The guidance which addresses cold shutdown/refueling IC/EALs in NEI 99-01 is intended to 

address both NUMARC/NESP-007 and NUREG-0654 users. For NUREG-0654 users, the scope of 

the cold shutdown/refueling initiative is limited to the “new” IC/EALs (CU2, CU4, CA1, CA2, and 

CG1), CA4 (compare with NUREG-054 Example Alert 10), and CS1 and CS2 (partially related to 

NUREG-0654 Example Site Area 10).
Revision 01/2003 C.1

Appendix C

Basis for Cold Shutdown/Refueling Initiating Conditions

Introduction

Recognition Category C is a new category of IC/EALs. Recognition Category S IC/EALs that were 

only applicable in cold shutdown or refueling have been removed from Recognition Category S and 

incorporated into the Recognition Category C. Those Category S IC/EALs that had applicability in 

modes other than just cold shutdown and/or refueling had their applicability changed and now exist 

in Recognition Categories C and S. In order to adequately address shutdown loss of inventory and 

loss of decay heat removal capability events new IC/EALs were added. The following matrix shows 

the relationship of Category C to Category S.

Category C 

IC/EAL

Category S

IC/EAL

New Significantly 

Revised 

Described in 

this Appendix

CU1 SU5 X

CU2 X X

CU3 SU1

CU4 X X

CU5 SU4

CU6 SU6

CU7 SU7

CU8 SU8 X

CA1 X X

CA2 X X

CA3 SA1

CA4 SA3 X X

CS1 SS5 X X

CS2 SS5 X X

CG1 X X

Recognition Category C completely replaces Recognition Category S when in Cold Shutdown and 

Refueling modes. It should be noted that the applicable Recognition Category A and H IC/EALs still 

apply when in Cold Shutdown and Refueling modes. Recognition Category F is not applicable to 

either the Cold Shutdown or Refueling modes.

Planning assumptions addressed in the development of the Category C IC/EALs are:

1. Variability of Initial Conditions - There will be a wide variability of initial conditions for the events 

addressed herein due to different plant configurations that could occur during shutdown periods. 

During power operations, the Fission Product Barrier Matrix classifies events on the loss or 

challenge to the fission product barriers. During shutdown conditions, these barriers may have 

intentionally been defeated. For this reason, these EALs are function and performance-based 

to the extent possible.

2. Redundancy and Diversity of Instruments - The redundancy and diversity of instruments 

typically used during power operations may be unavailable during shutdown periods. For 

example, in BWRs, loss of forced flow through the shutdown cooling, reactor recirculation, or 

reactor cleanup systems may result in the loss of accurate reactor coolant temperature

measurement. In some PWRs core exit thermocouples are disconnected prior to removing the 

reactor vessel head. Loss of forced decay heat removal flow may then render RCS loop or 
Revision 01/2003 C.2

RHR inlet temperature instruments readings invalid. For this reason, these EALs provide for 

alternative site specific time-based EALs in addition to the instrumentation EALs.

3. Available Decay Heat - The potential for core damage is directly related to the amount of decay 

heat available. Events that occur earlier in shutdown will have the potential for greater

consequence than will events that occur later in shutdown. This threshold would be reached 

sooner for events that occur early in a shutdown than those that occur late in a shutdown. For 

this reason, these EALs provide thresholds based on temperature increase.

The core damage potential is a function of the latent heat available and the capability of systems 

to remove the heat. During shutdown evolutions redundancy of many system components may 

have been intentionally decreased to facilitate maintenance therefore potentially increasing the 

probability that an event could lead to core damage. 

Available decay heat decreases from time of core shutdown. Approximately 6% of full power 

core thermal heat is available immediately after shutdown. At 30 days from shutdown, available 

decay heat is approximately 0.1% reactor power. Therefore the threat of core damage due to 

decay heat generation decreases over time. Typically, refueling mode is not entered until 100 

hours after shutdown effectively limiting the availability of decay heat.

4. Release Potential - The radionuclide inventory in the core is approximately 0.6 Ci/watt following 

extended operation at power. Thus, at shutdown, the core inventory for a typical 3000 Mwt 

reactor may be as much as 1.8E9 Ci, of which more than 1.0E7 Ci is iodine. With the 8.3 day 

half-life of I-131, there is a potential for significant radioactivity release well into a shutdown 

period.

5. Compatibility - The format and specific wording of the example EALs is expected to be modified 

to be compatible with individual utility nomenclature and procedure writing guidelines, provided 

that the intent of the example EAL is maintained.

6. Operating Experience - For BWRs, the shutdown EALS are intended to address concerns 

raised by NRC Office for Analysis and Evaluation of Operational Data (AEOD) Report

AEOD/EG09, "BWR Operating Experience Involving Inadvertent Draining of the Reactor

Vessel," dated August 8, 1986. This report states: “In broadest terms, the dominant causes of 

inadvertent reactor vessel draining are related to the operational and design problems

associated with the residual heat removal system when it is entering into or exiting from the 

shutdown cooling mode. During this transitional period water is drawn from the reactor vessel, 

cooled by the residual heat removal system heat exchangers (from the cooling provided by the 

service water system), and returned to the reactor vessel. First, there are piping and valves in 

the residual heat removal system which are common to both the shutdown cooling mode and 

other modes of operation such as low pressure coolant injection and suppression pool cooling. 

These valves, when improperly positioned, provide a drain path for reactor coolant to flow from 

the reactor vessel to the suppression pool or the radwaste system. Second, establishing or 

exiting the shutdown cooling mode of operation is entirely manual, making such evolutions 

vulnerable to personnel and procedural errors. Third, there is no comprehensive valve interlock 

arrangement for all the residual heat removal system valves that could be activated during 

shutdown cooling. Collectively, these factors have contributed to the repetitive occurrences of 

the operational events involving the inadvertent draining of the reactor vessel.”
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INITIATING CONDITIONS

The four initiating conditions classify the shutdown event on the basis of the Potential Loss or Loss 

of one or more of the cold shutdown barrier functions.

General Emergency (GE)

The GE is declared on the occurrence of the loss of function of all three barriers. If all three barriers 

are lost, the ability to maintain fission product inventory within the containment no longer exists. This 

represents a direct path for radioactive inventory to be released to the environment. This is 

consistent with the definition of a GE.

Site Area Emergency (SAE)

The two IC/EALs identified as SAE events are considered to involve the actual or likely losses of 

plant functions needed for the protection of the public. These IC/EALs represent a loss of one 

fission product barrier with the potential or actual loss of a second barrier. Additionally the IC/EALs 

address the status of the containment boundary in the classification scheme. The fact that these 

IC/EALs call for a SAE reflects the lower latent energy available to cause core melt. These IC/EALs 

also reflect the decreased availability of motive force for release of core activity in the unlikely event 

that fuel damage should occur. This is consistent with the fundamental definition of an SAE.

Alert

The four IC/EALs identified as Alert events are considered to represent substantial degradation in 

the level of safety of the plant. This is consistent with the fundamental definition of an Alert.

Notification of Unusual Event (NOUE)

The eight IC/EALs identified as NOUE events are considered to represent potential degradation in 

the level of safety of the plant. This is consistent with the fundamental definition of an Unusual 

Event.

EXAMPLE EALs

The Recognition Category C example IC/EALs are based on 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. A number of variables, such as initial vessel level, or shutdown heat removal system 

design, can have a significant impact on heat removal capability challenging the fuel clad barrier. 

The Loss example EAL represents the inability to restore and maintain RPV level to above the top of 

active fuel. Fuel damage is probable if RPV level cannot be restored, as available decay heat will 

cause boiling, further reducing the RPV level.

CU1, CU2, CU4, CA1, CA2, and CA4 IC/EALs are provided to serve as precursors to a loss of heat 

removal due to loss of inventory or function. CS1 and CS2 example EALs represent a significant 

loss of RCS inventory. The magnitude of this loss of water indicates that makeup systems have 

not been effective and may not be capable of preventing further RPV level decrease and potential 

core uncovery. EAL modifiers such as “for X minutes” are used when indications of level are 

unavailable to indicate the lack of a success path to restore inventory and the potential for core 
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uncovery. In the context of CS1 and CS2 EALs, “containment closure” is the action taken to secure 

primary or secondary containment and its associated structures, systems, and components as a 

functional barrier to fission product release under existing plant conditions. CG1 example EALs 

indicate that core uncovery due to loss of inventory has occurred for a period that could result in 

core damage. Additionally EALs that describe challenge to the Containment Barrier are included. 
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Appendix D

Basis for Permanently Defueled Station Initiating Conditions

Introduction

Recognition Category D is a new category that provides IC/EALs for Permanently Defueled 

stations. Category D was written to provide a stand alone set of IC/EALs for Permanently Defueled 

Stations. IC/EALs from Recognition Category A, C, F, S, and H were reviewed for applicability and 

where applicable have been included to address all Permanently Defueled station events.

A Permanently Defueled station is basically a spent fuel storage facility. This appendix is based on 

the assumption that the spent fuel was generated by an operating nuclear power station under a 

10CFR50 license that has ceased operations and intends to store the spent fuel for some period of 

time. The spent fuel is stored in a pool of water that serves as both the cooling medium for decay 

heat and shielding from direct radiation. The primary functions of this pool configuration become 

the emphasis of emergency classification methodology.

When in the permanently defueled condition, the licensee receives approval for exemption from 

specific emergency planning requirements. These exemptions must be approved by the NRC. 

The source term and relative risks associated with pool storage are the basis for maintaining only 

an onsite emergency plan. Calculations are provided in the licensing process that quantify

radioactive releases associated with plausible accidents as documented in the stations Safety 

Analysis Report (SAR).

D.1 Purpose of the Permanently Defueled ICs/EALs

The emergency classifications used are those provided by NUREG 0654/FEMA Rep.1. The NOUE 

classifications provide an increased awareness for abnormal conditions. The Alert classifications 

are specific to the actual or potential effects on the spent fuel in storage. The source term and 

motive force available in the permanently defueled condition is insufficient to warrant classifications 

of Site Area Emergency or General Emergency levels. Analyses for the credible design basis 

accidents are provided in the SAR.

Section 3.3 of NUMARC/NESP-007 emphasizes the need for accurate assessment and

classification of events, recognizing that over-classification, as well as under-classification, is to be 

avoided. Primary emphasis is intended to be placed on observable conditions in classifying

emergency events. In the permanently defueled condition, these conditions are primarily

associated with the spent fuel, the spent fuel pool systems used to provide cooling, and shielding. 

Effluent IC/EALs were included, however, to provide a basis for classifying events that cannot be 

readily classified based on observable condition alone. 

D.2. Initiating Conditions

There are two radiological effluent IC/EALs provided. The IC/EALs and the fundamental basis for 

classifications are:

Alert (D-AA1) Any UNPLANNED Release of Gaseous or Liquid Radioactivity to the

Environment that Meets or Exceeds 200 times the Technical Specification 

Release Limit for 15 Minutes or Longer.
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NOUE (D-AU1) Any UNPLANNED Release of Gaseous or Liquid Radioactivity to the

Environment that Meets or Exceeds 2 times the Technical Specification 

Release Limit for 60 Minutes or Longer.

D-AU1 and D-AA1 are NOT based on these particular values of offsite dose or dose rate but, 

rather, on the loss of plant control implied by a radiological release that exceeds a specified multiple 

of the RETS release limits for a specified period of time. 

IC/EALs D-AU1 and D-AA1 provide classification thresholds for UNPLANNED and/or uncontrolled 

releases of radioactivity to the environment. Calculations supporting the release rates specified in 

the EAL threshold values should be provided which quantify expected doses at the Restricted Area 

Boundary. The major isotope of concern in the permanently defueled condition is Kr-85.

Alert (D-AA2) UNCONTROLLED increase in plant radiation levels that impede

operations.

NOUE (D-AU2) UNCONTROLLED increase in plant radiation levels.

IC/EALs D-AU2 and D-AA2 provide classification thresholds for UNPLANNED and/or uncontrolled 

increases of radiation levels. These IC/EALs are concerned with unexpected increases in radiation 

levels within the facility that may affect operations. The Alert IC/EAL is specific to areas that will 

result in exposure to plant personnel. An increase of 100 mR/hr must also be accompanied by 

some impeded operations. The 100 mR/hr is arbitrary and may be set at a reasonable value for a 

specific application with justification for that value provided. The value of 15 mR/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 in this threshold without 

averaging, as a 30 day duration implies an event potentially more significant than an Alert. The 

NOUE uses a moderate increase in monitored radiation level that is not the result of a planned 

evolution and the source of the increase is not immediately recognized and controlled. The value 

selected (25 mR/hr) is arbitrary and may be set at a reasonable value for a specific application with 

justification for that value provided. This IC/EAL is included to raise awareness of an abnormal 

condition. 

One system malfunction is provided that is directly related to the permanently defueled condition 

methodology. The Spent Fuel pool inventory and temperature are the primary parameters that 

indicate the potential for fuel damage. 

NOUE (D-SU1) Decrease in Spent Fuel Pool level OR temperature increase that is not the 

result of a planned evolution.

The Site Specific value for decreasing level should be based on either the Technical Specification 

value for Spent Fuel Pool level or a calculated level that will result in prohibitive radiation levels in the 

Fuel Building. Justification for the level used in the EAL threshold value should allow for time to 

correct the level decrease prior to classification. 

The site-specific temperature should be chosen based on the starting point for fuel damage

calculations in the SAR. Typically, this temperature is 1250 to 1500F. Spent Fuel Pool temperature

is normally maintained well below this point thus allowing time to correct the cooling system 

malfunction prior to classification.
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It is assumed that the level and temperature thresholds described above result from an unplanned 

evolution. The NOUE is thus used to heighten awareness of control problems associated with 

spent fuel pool inventory or temperature control. Both of these conditions would have a long leadtime before fuel damage could occur due to decay heat.

Alert (D-HA1) Confirmed security event in the Fuel Building or Control Room.

NOUE (D-HU1) Confirmed security event with potential loss of level of safety of the plant.

A confirmed INTRUSION report is satisfied if physical evidence indicates the presence of a

HOSTILE FORCE within the Fuel Handling Building or Control Room. An Alert classification is 

warranted to account for the potential fuel damage that may be inflicted by a HOSTILE FORCE.

The NOUE is based on (site-specific) Site Security Plans. Security events that do not represent a 

potential degradation in the level of safety of the plant are reported under 10 CFR 73.71 or in some 

cases under 10 CFR 50.72. 

Reference is made to (site-specific) security shift supervision because these individuals are the 

designated personnel on-site qualified and trained to confirm that a security event is occurring or 

has occurred. Training on security event classification confirmation is closely controlled due to the 

strict secrecy controls placed on the plant Security Plan.

Alert (D-HA2) Other conditions judged warranting declaration of ALERT.

NOUE (D-HU2) Other conditions judged warranting declaration of an UNUSUAL EVENT

The Emergency Director has the discretion to classify events based on the classification level 

definitions. This discretion should be used when conditions or events are observed and no specific 

IC/EAL is apparent. A discretionary Alert will provide the onshift crew with additional personnel to 

address the abnormal condition. The NOUE will heighten awareness of the abnormal condition. 

NOUE (D-HU3) Natural or destructive phenomena inside the Protected Area affecting the 

ability to maintain spent fuel integrity.

Natural and destructive phenomena are classified at the NOUE level because of the unknown 

factors of the effects when they occur. Escalation to an Alert is through the observable effects of 

the natural and destructive phenomena via D- AA2.
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Appendix E

Basis for ISFSI Initiating Conditions

Introduction

An Independent Spent Fuel Storage Installation (ISFSI) is a complex that is designed and

constructed for the interim storage of spent nuclear fuel and other radioactive materials associated 

with spent fuel storage. An ISFSI which is located on the site of another facility may share common 

utilities/services and be physically connected with the other facility yet still be considered

independent provided, that such sharing of utilities and services or physical connections does not: 

(1) Increase the probability or consequences of an accident or malfunction of components,

structures, or systems that are important to safety; or (2) reduce the margin of safety as defined in 

the basis for any technical specification of either facility.

A Dry Cask Storage System (DCSS) may be used to store spent nuclear fuel under either a sitespecific or general license to operate an ISFSI. At present, any holder of an active reactor operating 

license under 10 CFR Part 50, has the authority to construct and operate an ISFSI under the 

provisions of the general license. Requirements for construction and pre-operational activities of 

such an ISFSI are discussed in Subparts K and L of 10 CFR Part 72. The requirements for 

pursuing a site-specific ISFSI license are discussed in Subparts B and C of 10 CFR Part 72.

E.1 Purpose of the ISFSI IC/EALs

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 the 

public health and safety. This evaluation shows that the maximum offsite dose to a member of the 

public offsite due to an accidental release of radioactive materials would not exceed 1 rem effective 

dose equivalent or an intake of 2 milligrams of soluble uranium (due to chemical toxicity).

The Final Rule governing Emergency Planning Licensing Requirements for Independent Spent Fuel 

Storage Facilities was posted in the Federal Register on June 22, 1995 (Federal Register Volume 

60, Number 120 June 22, 1995, Pages 32430-32442). The rule indicated that a significant amount 

of the radioactive material contained within a cask must escape its packaging and enter the 

atmosphere for there to be a significant environmental impact resulting from an accident involving 

the dry storage of spent nuclear fuel. There are two primary factors that protect the public health 

and safety from this unlikely dry storage radioactive material release event.

The first deals with regulatory requirements imposed on the design for the cask. Regulatory

requirements have sufficient safety margins so that (during normal storage cask handling

operations, off-normal events, adverse environmental conditions, and severe natural phenomena) 

the casks can not release a significant part of its inventory to the atmosphere.

The second factor deals with the cask general design criteria. The cask criteria requires that 1) 

design provides confinement safety functions during the unlikely but credible design basis events, 

2) the fuel clad must be protected against degradation that leads to gross rupture, and 3) the fuel 

must be retrievable. These general design criteria place an upper bound on the energy a cask can 

absorb before the fuel is damaged. No credible dynamic events were identified that could impart 

such significant amounts of energy to a storage cask after that cask is placed at the ISFSI. The 

second factor also considers the lack of dispersal mechanisms and the age of the spent fuel. There 
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is no significant dispersal mechanism for the radioactive material contained within a storage cask. 

Spent fuel required to be stored in an ISFSI must be cooled for at least 1 year. Based on the design 

limitations of most cask systems, the majority of spent fuel is cooled greater than 5 years. At this 

age, spent fuel has a heat generation rate that is too low to cause significant particulate dispersal in 

the unlikely event of a cask confinement boundary failure. Consequently, formal offsite planning is 

not required because the postulated worst-case accident involving an ISFSI has insignificant 

consequences to the public health and safety. 

10 CFR 72.32 provides two means for satisfying its requirements. 10 CFR 72.32 (a) requires that 

the application for an ISFSI be accompanied by an Emergency Plan. 10 CFR 72.32 (c) allows that 

the emergency plan required by 10 CFR 50.47 for a nuclear power reactor licensed for operation by 

the Commission shall be deemed to satisfy the requirements for an ISFSI located on the site or 

located within the exclusion area as defined in 10 CFR 100. 10 CFR 72.32 (a) requires that an 

ISFSI Emergency Plan include a classification system for classifying accidents as ”alerts”. In 

contrast to the 10 CFR 72.32 requirements, regulations governing 10 CFR 50.47 emergency plans 

specify four emergency classes: (1) notification of unusual events, (2) alert, (3) site area

emergency, and (4) general emergency, and require a determination of the adequacy of onsite and 

offsite emergency plans. 10 CFR 72.212(b)(6) requires that a general licensee review its reactor 

emergency plan to determine if its effectiveness is decreased and make necessary changes.

The expectations for 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, i.e., to provide assistance if requested. Even with regard to activation of a licensee’s 

emergency response organization (ERO), the ERO for a 10 CFR 72.32 emergency plan is not that 

prescribed under a 10 CFR 50.47 emergency plan, e.g., no Emergency Technical Support.

Consequently, the “alerts” contemplated by 10 CFR 72.32, have been classified as NOUEs herein. 

To do otherwise could lead to an inappropriate response posture on the part of offsite response 

organizations.

NUREG-1567, Standard Review Plan for Spent Fuel Dry Storage Facilities, descriptions of initiating 

events appear below:

• Fire onsite that might affect radioactive material of systems important to safety

• Severe natural phenomenon projected to occur that might affect radioactive material or 

systems important to safety (e.g., flood, tsunami, hurricane, tidal surge, hurricane force 

winds)

• Severe natural phenomenon or other incidents have occurred that may have affected 

radioactive material or systems important to safety, but initial assessment is not complete 

(e.g., beyond design basis earthquake, flood, tsunami, hurricane, tidal surge, hurricane force 

winds, tornado missiles, explosion, release of flammable gas)

• Elevated radiation levels or airborne contamination levels within the facility indicate severe 

loss of control (factor of 100 over normal levels)

• Ongoing security compromise (greater than 15 minutes)

• Accidental release of radioactivity within building confinement barrier (pool or waste 

management facility)

• Discovery of condition that creates a criticality hazard

• Other conditions that warrant precautionary activation of the licensee’s emergency 

response organization

Note that 10 CFR 72.32 also discusses emergency planning license application requirements for 

Monitored Retrievable Storage Facilities (MRS) and for ISFSIs that may process and/or repackage 

spent fuel. 10 CFR 72.32 (b) requires that an Emergency Plan for an MRS or one of these more 

complex ISFSIs include a classification system for classifying accidents as "alerts" or "site area 
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emergencies." NUREG-1567 provides a list of events that may initiate a site area emergency at 

one of these facilities. However, these facilities are beyond the scope of this discussion.

NUREG-1536, Standard Review Plan for Dry Cask Storage Systems, provides guidance for

performing safety reviews of applications for approval of spent fuel DCSS. The principal purposes 

of the DCSS Standard Review Plan (SRP) are to ensure the quality and consistency of staff 

reviews and to establish a well-defined basis from which to evaluate proposed changes in the 

scope of reviews. 

Accidents and events associated with natural phenomena may share common regulatory and 

design limits. By contrast, anticipated occurrences (off-normal conditions) are distinguished, in part, 

from accidents or natural phenomena by the appropriate regulatory guidance and design criteria. 

For example, the radiation dose from an off-normal event must not exceed the limits specified in 10 

CFR Part 20 and 10 CFR 72.104(a), whereas the radiation dose from an accident or natural 

phenomenon must not exceed the specifications of 10 CFR 72.106(b). Accident conditions may 

also have different allowable structural criteria.

According to NUREG 1536, the following accidents should be evaluated in the SAR. Because of the 

NRC's defense-in-depth approach, each should be evaluated regardless of whether it is highly 

unlikely or highly improbable. These do not constitute the only accidents that should be addressed if 

the SAR is to serve as a reference for accidents for the site-specific application. Others that may 

be derived from a hazard analysis could include accidents resulting from operational error,

instrument failure, lightning, and other occurrences. Accident situations that are not credible

because of design features or other reasons should be identified and justified in the SAR. 

• Section 2.0-V.2.b(3) - Accident Conditions 

(a) Cask Drop 

(b) Cask Tipover 

(c) Fire 

(d) Fuel Rod Rupture 

(e) Leakage of the Confinement Boundary 

(f) Explosive Overpressure 

(g) Air Flow Blockage 

• Section 2.0-V.2.b(4) - Natural Phenomena Events 

(a) Flood 

(b) Tornado 

(c) Earthquake 

(d) Burial under Debris 

(e) Lightning 

(f) Other natural phenomena events (including seiche, tsunami, and hurricane) 

The emergency classifications used are those provided by NUREG 0654/FEMA Rep.1. NOUE 

classifications provide an increased awareness for abnormal conditions. The source term and 

motive force available at a simple ISFSI is insufficient to warrant classifications above the NOUE 

level using the 10 CFR 50 emergency classification scheme.

Section 3.3 of NUMARC/NESP-007 emphasizes the need for accurate assessment and

classification of events. It is intended that primary emphasis be placed on observable conditions in 

classifying emergency events. For an ISFSI, these conditions are primarily associated with the 

CONFINEMENT BOUNDARY of a loaded fuel storage cask.
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E.2. Initiating Conditions

NOUE (E-HU1) Damage to a loaded cask CONFINEMENT BOUNDARY.

The Emergency Director has the discretion to classify events based on the classification level 

definitions. This discretion should be used when conditions or events are observed and no specific 

IC/EAL is apparent. The NOUE will heighten awareness of the abnormal condition. Natural

phenomena events and accident conditions are classified at the NOUE level in the event that a 

loaded cask CONFINEMENT BOUNDARY is damaged or violated.

NOUE (E-HU2) Security event with potential loss of level of safety of the ISFSI.

The NOUE is based on (site-specific) ISFSI Security Plans. Security events that do not represent a 

potential degradation in the level of safety are reported under 10 CFR 73.71 or in some cases under 

10 CFR 50.72. 

Reference is made to (site-specific) security shift supervision because these individuals are the 

designated personnel on-site qualified and trained to confirm that a security event is occurring or 

has occurred. Training on security event classification confirmation is closely controlled due to the 

strict secrecy controls placed on the plant Security Plan.

The term “Protected Area” is defined by 10 CFR 73.2(a) as “an area encompassed by physical 

barriers and to which access is controlled.” Response to, and classification of, an un-neutralized 

intrusion into an operating plant’s Protected Area and a separate ISFSI Protected Area would differ 

significantly. The Final Rule governing Physical Protection for Spent Nuclear Fuel and High-Level 

Radioactive Waste (Federal Register Volume 63, Number 94, dated May 15, 1998, Pages 26955-

26963), stated that “[t]he Commission believes that the appropriate level of physical protection for 

spent fuel and high-level radioactive waste lies somewhere between industrial-grade security and 

the level that is required at operating power reactors. The Commission also notes that the nature of 

spent fuel and of its storage mechanisms offers unique advantages in protecting the material.” 

Further, “[t]he Commission never intended that onsite physical protection personnel at an ISFSI 

would provide a response to a safeguards event other than calling for assistance from local law 

enforcement or other designated response force unless their timely response could not be

ensured.” 10 CFR 73.51 calls for unarmed watchmen, not armed guards. Therefore, it is 

reasonable to treat the “Protected Area” around the operating units and around the ISFSI differently 

with regard to the EALs governing classification unless the ISFSI Protected Area is bounded by the 

plant Protected Area. This same rationale may be applicable to natural or other hazards in addition 

to Security events.