NUREG-1520

Standard Review Plan

for the Review of a License Application

for a Fuel Cycle Facility

Final Report

March 2002

Office of Nuclear Material Safety and Safeguards

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001
1 Standard format and content guides for Fundamental Nuclear Material Control Plans 

Abstract iii NUREG-1520

ABSTRACT

This “Standard Review Plan (SRP) for the Review of a License Application for a Fuel Cycle

Facility” (NUREG-1520) provides guidance to the staff reviewers in the U.S. Nuclear Regulatory

Commission (NRC), Office of Nuclear Material Safety and Safeguards (NMSS) who perform

safety and environmental impact reviews of applications to construct or modify and operate

nuclear fuel cycle facilities. As such, this SRP ensures the quality, uniformity, and predictability

of the staff reviews. This SRP also makes information about licensing acceptance criteria

widely available to interested members of the public and the regulated industry. Each SRP

section addresses the responsibilities of the staff reviewers, the matters that they review, the

Commission's regulations pertinent to specific technical matters, the acceptance criteria used

by the staff, the process and procedures used to accomplish the review, and the conclusions

that are appropriate to summarize the review.

This SRP also addresses the long-standing health, safety, and environmental protection

requirements of Title 10, Parts 20 and 70, of the Code of Federal Regulations (10 CFR Parts 20

and 70) as well as the amended accident safety requirements reflected in the new Subpart H of

10 CFR Part 70. For example, the chapters concerning radiation safety, environmental

protection, emergency management, and decommissioning contain acceptance criteria that are

primarily set by regulations that remained unaffected by the recent revision to 10 CFR Part 70.

The new Subpart H of 10 CFR Part 70 identifies risk-informed performance requirements and

requires applicants and existing licensees to conduct an integrated safety analysis (ISA) and

submit an ISA Summary, as well as other information. Chapters 3 (ISA) and 11 (Management

Measures) of this SRP are the primary chapters that address the staff’s review in relation to the

performance and other related requirements of Subpart H. 

This SRP is not a substitute for NRC regulations and compliance is not required. The

approaches and methods in this report are provided for information only. Methods and solutions

different from those described in this report will be acceptable if they provide a basis for the staff

to make the determination needed to issue or continue a license.

This SRP focuses on safety and environmental impact reviews. Review criteria applicable to

the safeguards sections of license applications were developed earlier and are published in

NUREGs 1280 and 1065.1
Contents v NUREG-1520

CONTENTS

Page

ABSTRACT ................................................................. iii

EXECUTIVE SUMMARY........................................................ xi

ACRONYMS AND ABBREVIATIONS .............................................. xiii

GLOSSARY ................................................................ xvi

INTRODUCTION ............................................................ xix

1 GENERAL INFORMATION ................................................. 1-1

1.1 FACILITY AND PROCESS DESCRIPTION ................................ 1-1

1.1.1 Purpose of Review ............................................ 1-1

1.1.2 Responsibility for Review ....................................... 1-1

1.1.3 Areas of Review .............................................. 1-1

1.1.4 Acceptance Criteria ........................................... 1-1

1.1.5 Review Procedures ........................................... 1-2

1.1.6 Evaluation Findings ........................................... 1-2

1.1.7 References .................................................. 1-3

1.2 INSTITUTIONAL INFORMATION ......................................... 1-3

1.2.1 Purpose of Review ............................................ 1-3

1.2.2 Responsibility for Review ....................................... 1-3

1.2.3 Areas of Review .............................................. 1-3

1.2.4 Acceptance Criteria ........................................... 1-3

1.2.5 Review Procedures ........................................... 1-5

1.2.6 Evaluation Findings ........................................... 1-5

1.2.7 References .................................................. 1-5

1.3 SITE DESCRIPTION .................................................. 1-6

1.3.1 Purpose of Review ............................................ 1-6

1.3.2 Responsibility for Review ....................................... 1-6

1.3.3 Areas of Review .............................................. 1-6

1.3.4 Acceptance Criteria ........................................... 1-7

1.3.5 Review Procedures ........................................... 1-8

1.3.6 Evaluation Findings ........................................... 1-8

1.3.7 References .................................................. 1-9

2 ORGANIZATION AND ADMINISTRATION ..................................... 2-1

2.1 PURPOSE OF REVIEW ............................................... 2-1

2.2 RESPONSIBILITY FOR REVIEW ........................................ 2-1

2.3 AREAS OF REVIEW .................................................. 2-1

2.4 ACCEPTANCE CRITERIA ............................................. 2-2

2.4.1 Regulatory Requirements ...................................... 2-2

2.4.2 Regulatory Guidance .......................................... 2-2

2.4.3 Regulatory Acceptance Criteria .................................. 2-2

2.5 REVIEW PROCEDURES .............................................. 2-3

2.5.1 Acceptance Review ........................................... 2-3

2.5.2 Safety Evaluation ............................................. 2-3

2.6 EVALUATION FINDINGS ............................................... 2-4

2.7 REFERENCES ...................................................... 2-4
Contents vi NUREG-1520

3 INTEGRATED SAFETY ANALYSIS (ISA) AND ISA SUMMARY ..................... 3-1

3.1 PURPOSE OF REVIEW ............................................... 3-1

3.1.1 Safety Program and ISA Commitments ........................... 3-1

3.1.2 ISA Summary ................................................ 3-2

3.2 RESPONSIBILITY FOR REVIEW ........................................ 3-2

3.3 AREAS OF REVIEW .................................................. 3-2

3.3.1 Safety Program and ISA Commitments ........................... 3-3

3.3.2 ISA Summary and ISA Documentation ............................ 3-4

3.4 ACCEPTANCE CRITERIA ............................................. 3-7

3.4.1 Regulatory Requirements ...................................... 3-7

3.4.2 Regulatory Guidance .......................................... 3-8

3.4.3 Regulatory Acceptance Criteria .................................. 3-8

3.5 REVIEW PROCEDURES ............................................. 3-28

3.5.1 Acceptance Review .......................................... 3-28

3.5.2 Safety Evaluation ............................................ 3-29

3.6 EVALUATION FINDINGS .............................................. 3-33

3.7 REFERENCES ..................................................... 3-34

APPENDIX A EXAMPLE PROCEDURE FOR ACCIDENT SEQUENCE EVALUATION ... 3-A-1

A.1 RISK MATRIX DEVELOPMENT ....................................... 3-A-2

A.2 CONSEQUENCE CATEGORY ASSIGNMENT ........................... 3-A-5

A.3 LIKELIHOOD CATEGORY ASSIGNMENT ............................... 3-A-6

A.4 ASSESSING EFFECTIVENESS OF IROFS .............................. 3-A-7

A.5 RISK INDEX EVALUATION SUMMARY.................................. 3-A-8

A.6 ACCIDENT SUMMARY AND RISK INDEX ASSIGNMENT FOR TABLE A-7 .... 3-A-14

A.7 DETERMINATION OF LIKELIHOOD CATEGORY IN TABLE A-8 ............ 3-A-16

A.8 DETERMINATION OF FAILURE FREQUENCY INDEX NUMBERS IN 

TABLE A-9 ....................................................... 3-A-16

A.9 DETERMINATION OF FAILURE PROBABILITY INDEX NUMBERS IN 

TABLE A-10 ...................................................... 3-A-17

A.10 DETERMINING MANAGEMENT MEASURES FOR IROFS ................ 3-A-17

A.11 RISK-INFORMED REVIEW OF IROFS ............................... 3-A-19

4 RADIATION PROTECTION ................................................ 4-1

4.1 PURPOSE OF REVIEW ............................................... 4-1

4.2 RESPONSIBILITY FOR REVIEW ........................................ 4-1

4.3 AREAS OF REVIEW .................................................. 4-1

4.4 ACCEPTANCE CRITERIA ............................................. 4-2

4.4.1 Commitment to Radiation Protection Program Implementation ......... 4-2

4.4.2 Commitment to an ALARA Program .............................. 4-3

4.4.3 Organization and Personnel Qualifications ......................... 4-4

4.4.4 Commitment to Written Procedures .............................. 4-5

4.4.5 Training ..................................................... 4-5

4.4.6 Ventilation and Respiratory Protection Programs .................... 4-7

4.4.7 Radiation Surveys and Monitoring Programs ....................... 4-8

4.4.8 Additional Program Commitments ............................... 4-12

4.5 REVIEW PROCEDURES ............................................. 4-12

4.5.1 Acceptance Review .......................................... 4-12

4.5.2 Safety Evaluation ............................................ 4-13
Contents vii NUREG-1520

4.6 EVALUATION FINDINGS .............................................. 4-13

4.7 REFERENCES ..................................................... 4-13

5 NUCLEAR CRITICALITY SAFETY ........................................... 5-1

5.1 PURPOSE OF REVIEW ............................................... 5-1

5.2 RESPONSIBILITY FOR REVIEW ........................................ 5-1

5.3 AREAS OF REVIEW .................................................. 5-1

5.3.1 Management of the NCS Program ................................ 5-1

5.3.2 Organization and Administration ................................. 5-2

5.3.3 Management Measures ........................................ 5-2

5.3.4 Methodologies and Technical Practices ........................... 5-3

5.4 ACCEPTANCE CRITERIA ............................................. 5-4

5.4.1 Regulatory Requirements ...................................... 5-4

5.4.2 Regulatory Guidance .......................................... 5-4

5.4.3 Regulatory Acceptance Criteria .................................. 5-5

5.5 REVIEW PROCEDURES ............................................. 5-18

5.5.1 Acceptance Review .......................................... 5-19

5.5.2 Safety Evaluation ............................................ 5-19

5.6 EVALUATION FINDINGS .............................................. 5-19

5.7 REFERENCES ..................................................... 5-20

6 CHEMICAL PROCESS SAFETY ............................................ 6-1

6.1 PURPOSE OF REVIEW ............................................... 6-1

6.2 RESPONSIBILITY FOR REVIEW ........................................ 6-1

6.3 AREAS OF REVIEW .................................................. 6-1

6.4 ACCEPTANCE CRITERIA ............................................. 6-3

6.4.1 Regulatory Requirements ...................................... 6-3

6.4.2 Regulatory Guidance .......................................... 6-3

6.4.3 Regulatory Acceptance Criteria .................................. 6-3

6.5 PROCEDURES FOR REVIEW ......................................... 6-6

6.5.1 Acceptance Review ........................................... 6-6

6.5.2 Safety Evaluation ............................................. 6-6

6.6 EVALUATION FINDINGS ............................................... 6-7

6.7 REFERENCES ...................................................... 6-8

7 FIRE SAFETY ........................................................... 7-1

7.1 PURPOSE OF REVIEW ............................................... 7-1

7.2 RESPONSIBILITY FOR REVIEW ........................................ 7-1

7.3 AREAS OF REVIEW .................................................. 7-1

7.4 ACCEPTANCE CRITERIA ............................................. 7-2

7.4.1 Regulatory Requirements ...................................... 7-2

7.4.2 Regulatory Guidance .......................................... 7-2

7.4.3 Regulatory Acceptance Criteria .................................. 7-2

7.5 REVIEW PROCEDURES .............................................. 7-5

7.5.1 Acceptance Review ........................................... 7-5

7.5.2 Safety Evaluation ............................................. 7-6

7.6 EVALUATION FINDINGS ............................................... 7-6

7.7 REFERENCES ...................................................... 7-7
Contents viii NUREG-1520

8 EMERGENCY MANAGEMENT .............................................. 8-1

8.1 PURPOSE OF REVIEW ............................................... 8-1

8.2 RESPONSIBILITY FOR REVIEW ........................................ 8-1

8.3 AREAS OF REVIEW .................................................. 8-1

8.3.1 Emergency Plan .............................................. 8-2

8.3.2 Evaluation That No Emergency Plan is Required .................... 8-2

8.4 ACCEPTANCE CRITERIA ............................................. 8-2

8.4.1 Regulatory Requirements ...................................... 8-2

8.4.2 Regulatory Guidance .......................................... 8-3

8.4.3 Regulatory Acceptance Criteria .................................. 8-3

8.4.3.1 Emergency Plan .............................................. 8-3

8.4.3.2 Evaluation That No Emergency Plan Is Required ................... 8-11

8.5 AMENDMENTS OR CHANGES TO THE EMERGENCY PLAN ................ 8-13

8.6 REVIEW PROCEDURES ............................................. 8-13

8.6.1 Acceptance Review .......................................... 8-13

8.6.2 Safety Evaluation ............................................ 8-13

8.7 EVALUATION FINDINGS .............................................. 8-14

8.8 REFERENCES ..................................................... 8-15

9 ENVIRONMENTAL PROTECTION .......................................... 9-1

9.1 PURPOSE OF REVIEW ............................................... 9-1

9.2 RESPONSIBILITY FOR REVIEW ........................................ 9-1

9.3 AREAS OF REVIEW .................................................. 9-1

9.3.1 Environmental Report ......................................... 9-1

9.3.2 Environmental Protection Measures .............................. 9-2

9.4 ACCEPTANCE CRITERIA ............................................. 9-3

9.4.1 Regulatory Requirements ...................................... 9-3

9.4.2 Regulatory Guidance .......................................... 9-4

9.4.3 Regulatory Acceptance Criteria .................................. 9-5

9.5 REVIEW PROCEDURES ............................................. 9-16

9.5.1 Environmental Report ........................................ 9-16

9.5.2 Environmental Protection ...................................... 9-16

9.6 EVALUATION FINDINGS .......................................... 9-17

9.6.1 Introduction ................................................. 9-17

9.6.2 NEPA Documentation and Coordination .......................... 9-18

9.6.3 Safety Evaluation ............................................ 9-20

9.7 REFERENCES ..................................................... 9-21

10 DECOMMISSIONING .................................................... 10-1

10.1 PURPOSE OF REVIEW ............................................. 10-1

10.2 RESPONSIBILITY FOR REVIEW ...................................... 10-1

10.3 AREAS OF REVIEW ................................................ 10-1

10.4 ACCEPTANCE CRITERIA ........................................... 10-2

10.4.1 Regulatory Requirements ...................................... 10-2

10.4.2 Regulatory Guidance .......................................... 10-2

10.5 REVIEW PROCEDURES ............................................ 10-2

10.5.2 Safety Review ............................................... 10-3

10.6 EVALUATION FINDINGS ............................................. 10-3

10.7 REFERENCES .................................................... 10-3
Contents ix NUREG-1520

11 MANAGEMENT MEASURES .............................................. 11-1

11.1 PURPOSE OF REVIEW ............................................. 11-1

11.2 RESPONSIBILITY FOR REVIEW ...................................... 11-1

11.3 AREAS OF REVIEW ................................................ 11-2

11.3.1 Configuration Management (CM) ................................. 11-2

11.3.2 Maintenance ................................................. 11-3

11.3.3 Training and Qualifications ..................................... 11-4

11.3.4 Procedures ................................................. 11-4

11.3.5 Audits and Assessments ....................................... 11-6

11.3.6 Incident Investigations ......................................... 11-6

11.3.7 Records Management ......................................... 11-6

11.3.8 Other QA Elements ........................................... 11-7

11.4 ACCEPTANCE CRITERIA ........................................... 11-8

11.4.1 Regulatory Requirements ...................................... 11-8

11.4.2 Regulatory Guidance .......................................... 11-8

11.4.3 Regulatory Acceptance Criteria .................................. 11-9

11.5 REVIEW PROCEDURES ........................................... 11-24

11.5.1 Acceptance Review .......................................... 11-24

11.5.2 Safety Evaluation ............................................ 11-24

11.6 EVALUATION FINDINGS ............................................ 11-30

11.6.1 CM ....................................................... 11-30

11.6.2 Maintenance ................................................ 11-31

11.6.3 Training and Qualification ..................................... 11-32

11.6.4 Procedures ................................................ 11-32

11.6.5 Audits and Assessments ...................................... 11-32

11.6.6 Incident Investigations ........................................ 11-32

11.6.7 Records Management ........................................ 11-33

11.6.8 Other QA Elements .......................................... 11-33

11.7 REFERENCES ................................................... 11-34

APPENDIX A CHECKLIST FOR PROCEDURES ............................... 11-A-1

APPENDIX B RECORDS .................................................. 11-B-1

Examples of Records ..................................................... 11-B-1

APPENDIX A FILING STANDARDS FOR SUBMITTALS ............................. A-1

Proprietary Information ........................................................ A-1

Classified Information ......................................................... A-1

Style and Composition ........................................................ A-1

Physical Specifications ....................................................... A-2
Contents x NUREG-1520

Tables

Table A-1: Consequence Severity Categories Based on 10 CFR 70.61 ............... 3-A-2

Table A-2: Likelihood Categories Based on 10 CFR 70.61 ......................... 3-A-3

Table A-3: Risk Matrix With Risk Index Values ................................... 3-A-3

Table A-4: Qualitative Categorization of IROFS .................................. 3-A-4

Table A-5: Consequence Severity Categories Based on 10 CFR 70.61 ............... 3-A-6

Table A-6: Event Likelihood .................................................. 3-A-7

Table A-7: Example Accident Sequence Summary and Risk Index Assignment ........ 3-A-9

Table A-8: Determination of Likelihood Category ................................. 3-A-9

Table A-9: Failure Frequency Index Numbers .................................. 3-A-10

Table A-10: Failure Probability Index Numbers .................................. 3-A-11

Table A-11: Failure Duration Index Numbers ................................... 3-A-11

Table A-12: Accident Sequence Descriptions .................................. 3-A-13

Table A-13: Descriptive List of Items Relied on for Safety ......................... 3-A-13
Executive Summary xi NUREG-1520

EXECUTIVE SUMMARY

This “Standard Review Plan (SRP) for the Review of a License Application for a Fuel Cycle

Facility” (NUREG-1520) provides U.S. Nuclear Regulatory Commission (NRC) guidance for

reviewing and evaluating the health, safety, and environmental protection aspects of applications

for licenses to possess and use special nuclear material (SNM) to produce nuclear reactor fuel. 

This guidance also applies to the review and evaluation of proposed amendments and license

renewal applications for nuclear fuel cycle facilities. 

The principal purpose of this SRP is to ensure the quality and uniformity of reviews conducted by

the staff of the NRC’s Office of Nuclear Material Safety and Safeguards (NMSS). This SRP also

provides a well-defined foundation from which to evaluate proposed changes in the scope, level

of detail, and acceptance criteria of reviews. Another important purpose of this SRP is to make

information about regulatory reviews widely available and to improve communication and

understanding of the staff review process. In addition, because this SRP describes the scope,

level of detail, and acceptance criteria for reviews, it serves as regulatory guidance for applicants

who need to determine what information to present in a license application and related

documents.

This SRP addresses the long-standing health, safety, and environmental protection

requirements of Title 10, Parts 20 and 70, of the Code of Federal Regulations (10 CFR Part 70

and 10 CFR Part 20), as well as the newer accident safety requirements reflected in the new

Subpart H of 10 CFR Part 70. For example, the chapters concerning radiation safety,

environmental protection, emergency management, and decommissioning contain acceptance

criteria that are primarily set by regulations that remain unaffected by the recent revision to 

10 CFR Part 70. Review criteria applicable to the safeguards sections of license applications

were developed earlier and are published in NUREGs 1280 and 1065 which are standard format

and content guides for Fundamental Nuclear Material Control Plans for high enriched uranium

facilities and low enriched uranium facilities, respectively.

The new Subpart H of 10 CFR Part 70 identifies risk-informed performance requirements and

requires applicants and existing licensees to conduct an integrated safety analysis (ISA) and

submit an ISA Summary, as well as other information. Chapters 3 (ISA and ISA Summary) and

11 (Management Measures) of this SRP are the primary chapters that address the staff’s review

in relation to the performance and other related requirements of Subpart H.

 Each nuclear fuel cycle facility license application should contain a safety program description

that addresses all of the topics listed in the table of contents of this SRP, in the same order in

which they are presented in this document. In general terms, the requirements in 

10 CFR Part 70 specify the information that an applicant must supply in its safety program

description. This SRP compliments 10 CFR Part 70 by identifying the specific information to be

submitted by an applicant and evaluated by the staff. 

The major topics addressed within the safety program description of a facility license application

are discussed in separate chapters of this SRP, including general information, organization and

administration, integrated safety analysis, radiation safety, nuclear criticality safety, chemical

process safety, fire safety, emergency management, environmental protection,

decommissioning, and management measures. Each of these chapters contains seven

sections including (1) purpose of review, (2) responsibility for review, (3) areas of review, 
Executive Summary xii NUREG-1520

(4) acceptance criteria, (5) review procedures, (6) evaluation findings, and (7) references.

Prospective applicants should study the topic areas treated in the chapters of this SRP, paying

particular attention to the sections entitled “Areas of Review” and “Acceptance Criteria.” In

addition, in accordance with 10 CFR 70.62 and 70.65, applicants are required to submit an ISA

Summary in conjunction with the application.

This SRP provides information and guidance to assist the licensing staff and the applicant in

understanding the underlying objectives of the regulatory requirements, the relationships among

NRC requirements, the licensing process, the major guidance documents that the NRC staff

has prepared for licensing fuel cycle facilities, and information about aspects of the staff review

process set out in individual SRP sections. Staff analyses are intended to provide regulatory

confirmation of reasonable assurance of safe design and operation. A staff determination of

reasonable assurance leads to a decision to issue or renew a license or to approve an

amendment to the license. If the staff determines that an application contains inadequate

information or commitments, the staff will inform the applicant of what is needed and the basis

on which the determination was made.

The “Acceptance Criteria” delineated in this SRP are intended to communicate the underlying

objectives, but do not represent the only means of satisfying those objectives. Rather an

applicant should tailor its safety program to the particular features of its facility. If an applicant

chooses approaches other than those presented in this SRP, the applicant should identify the

portions of its license application that differ from the design approaches and acceptance criteria

of the SRP, and should demonstrate how the proposed alternatives provide an acceptable

method of complying with the Commission's regulations. The staff retains the responsibility to

make an independent determination concerning the adequacy of the applicants’s proposed

approaches.
Acronyms xiii NUREG-1520

ACRONYMS AND ABBREVIATIONS

AEC active engineered control

AEGL acute exposure guideline level

ALARA as low as is reasonably achievable

ANS American Nuclear Society

ANSI American National Standards Institute

ASTM American Society for Testing and Materials

BDC baseline design criteria

CAAS criticality accident alarm system

CFR Code of Federal Regulations

CM configuration management

DFP decommissioning funding plan

DP decommissioning plan

EA environmental assessment

EAL emergency action level

EIS environmental impact statement

ERPG Emergency Response Planning Guidelines

FHA fire hazards analysis

FM facility maintenance

FONSI finding of no significant impact

HEPA high efficiency particulate

HS&E health, safety, and environment

ICRP International Commission on Radiological Protection

IROFS item(s) relied on for safety

ISA integrated safety analysis
Acronyms xiv NUREG-1520

ISO International Organization for Standardization

LIB Licensing and International Safeguards Branch

MDC minimum detectable concentration

MOU memorandum of understanding

NCS nuclear criticality safety

NCRP National Council on Radiation Protection and Measurements

NEPA National Environmental Policy Act

NFPA National Fire Protection Association

NIST National Institute of Standards and Technology

NMSS Nuclear Material Safety and Safeguards, Office of (NRC)

NRC U.S. Nuclear Regulatory Commission

OSHA Occupational Safety and Health Administration

PEC passive engineered control

PEL permissible exposure limit

PHA process hazard analysis

P&ID piping and instrumentation diagram

PM preventive maintenance

QA quality assurance

QC quality control

RAI request for additional information

RWP radiation work permits

SECY Office of the Secretary of the Commission (NRC)

SER safety evaluation report

SNM special nuclear material

SRP standard review plan
Acronyms xv NUREG-1520

TEDE total effective dose equivalent

TWA time-weighted average

UA Underwriters Laboratories
Glossary xvi NUREG-1520

GLOSSARY

This glossary defines technical/industry terms that are used consistently throughout this SRP, or

references the related definitions in either 10 CFR 20.1003 or 10 CFR 70.4. This glossary does

not define terms that may have different connotations in different contexts; such terms are

defined in the various chapters of this SRP.

Active engineered control

(AEC)

A physical device that uses active sensors, electrical

components, or moving parts to maintain safe process

conditions without any required human action.

Accident sequence An unintended sequence of events that, given the failure of

certain IROFS identified in the sequence, would result in

environmental contamination, radiation exposure, release of

radioactive material, inadvertent nuclear criticality, or exposure

to hazardous chemicals (provided that the chemicals are

produced from licensed radioactive material). The term

“accident” may be used interchangeably with “accident

sequence.”

Acute This term is defined in 10 CFR 70.4.

Administrative control Either an augmented administrative control or a simple

administrative control, as defined herein.

Augmented administrative

control

A procedurally required or prohibited human action, combined

with a physical device that alerts the operator that the action is

needed to maintain safe process conditions, or otherwise adds

substantial assurance of the required human performance.

Available and reliable to

perform their function

when needed

This term is defined in 10 CFR 70.4.

Baseline design criteria A set of criteria specifying design features and management

measures that are required and acceptable under certain

conditions for new processes or facilities specified in 

10 CFR 70.64. In general, these criteria are the acceptance

criteria that apply to safety design for new facilities and new

processes, as described in the chapters of this SRP.

Configuration

management (CM)

This term is defined in 10 CFR 70.4. 

Controlled area This term is defined in 10 CFR 20.1003. 
Glossary xvii NUREG-1520

Controlled parameter A measurable parameter that is maintained within a specified

range by one or more specific controls to ensure the safety of

an operation.

Consequence Any result of interest caused by an event or sequence of

events. In this context, “adverse consequence” refers to

adverse health or safety effects on either workers, the public,

or the environment.

Critical mass of special

nuclear material (SNM) 

This term is defined in 10 CFR 70.4.

Double contingency

protection

A characteristic or attribute of a process that has incorporated

sufficient safety factors so that at least two unlikely,

independent, and concurrent changes in process conditions

are required before a nuclear criticality accident is possible.

Engineered control See active engineered control or a passive engineered control.

External event An event for which the likelihood cannot be altered by changes

to the regulated facility or its operation. This would include all

natural phenomena events, plus airplane crashes, explosions,

toxic releases, fires, etc., occurring near or on the plant site. 

Hazardous chemicals

produced from licensed

materials

This term is defined in 10 CFR 70.4. 

Integrated safety analysis

(ISA)

This term is defined in 10 CFR 70.4. 

Integrated safety analysis

summary

This term is defined in 10 CFR 70.4.

Items relied on for safety

(IROFS)

This item is defined in 10 CFR 70.4. All safety controls, as

defined in this SRP, are IROFS.

Management measures This term is defined in 10 CFR 70.4.

Mitigative control A control intended to reduce the consequences of an accident

sequence, not to prevent it. When a mitigative control works

as intended, the results of the sequence are called the

mitigated consequences.

Natural phenomena event Earthquakes, floods, tornadoes, tsunamis, hurricanes, and

other events that occur in the natural environment and could
Glossary xviii NUREG-1520

adversely affect safety. 

Natural phenomena

events may be credible or

incredible, depending on

their likelihood of

occurrence. 

New processes at existing

facilities

Systems-level or facility-level design changes to process

equipment, process technology, facility layout, or types of

licensed material possessed or used. Generally, this

definition does not include component-level design changes

or equipment replacement.

Passive engineered

control

A device that uses only fixed physical design features to

maintain safe process conditions without any required human

action. 

Preventive control A control intended to prevent an accident (i.e., any of the

radiological or chemical consequences described in 

10 CFR 70.61).

Safety control A system, device, or procedure that is intended to regulate a

device, process, or human activity to maintain a safe state. 

Controls may be engineered controls or administrative

(procedural) controls, and may be either preventive or

mitigative, as defined herein. 

Safe process conditions The defined ranges or sets of acceptable values of one or

more controlled parameters.

Simple administrative

control

A procedural human action that is prohibited or required to

maintain safe process conditions.

Unacceptable

performance deficiencies 

This term is defined in 10 CFR 70.4. 

Worker This term is defined in 10 CFR 70.4.
Introduction xix NUREG-1520

INTRODUCTION

This “Standard Review Plan (SRP) for the Review of a License Application for a Fuel Cycle

Facility” (NUREG-1520) provides U.S. Nuclear Regulatory Commission (NRC) guidance for

reviewing and evaluating the health, safety, and environmental protection aspects of applications

for licenses to possess and use special nuclear material (SNM) to produce nuclear reactor fuel. 

This guidance also applies to the review and evaluation of proposed amendments and license

renewal applications for nuclear fuel cycle facilities. 

The principal purpose of this SRP is to ensure the quality and uniformity of reviews conducted by

the staff of the NRC’s Office of Nuclear Material Safety and Safeguards (NMSS). This SRP also

provides a well-defined foundation from which to evaluate proposed changes in the scope, level

of detail, and acceptance criteria of reviews. Another important purpose of this SRP is to make

information about regulatory reviews widely available and to improve communication and

understanding of the staff review process. In addition, because this SRP describes the scope,

level of detail, and acceptance criteria for reviews, it serves as regulatory guidance for applicants

who need to determine what information to present in a license application and related

documents.

This SRP addresses the long-standing health, safety, and environmental protection

requirements of Title 10, Parts 20 and 70, of the Code of Federal Regulations (10 CFR Part 70

and 10 CFR Part 20), as well as the newer accident safety requirements reflected in the new

Subpart H of 10 CFR Part 70. For example, the chapters concerning radiation safety,

environmental protection, emergency management, and decommissioning contain acceptance

criteria that are primarily set by regulations that remain unaffected by the recent revision to 

10 CFR Part 70. Review criteria applicable to the safeguards sections of license applications

were developed earlier and are published in NUREGs 1280 and 1065 which are standard format

and content guides for Fundamental Nuclear Material Control Plans for high enriched uranium

facilities and low enriched uranium facilities, respectively.

The new Subpart H of 10 CFR Part 70 identifies risk-informed performance requirements and

requires applicants and existing licensees to conduct an integrated safety analysis (ISA) and

submit an ISA Summary, as well as other information. Chapters 3 (ISA) and 11 (Management

Measures) of this SRP are the primary chapters that address the staff’s review in relation to the

performance and other related requirements of Subpart H. For new facilities that have not

already been designed, built, licensed and operated, Subpart H also requires adherence to

baseline design criteria (BDC), as specified in 10 CFR 70.64.

This SRP is a guidance document that is intended for use during the review of license

applications, license renewal applications, and amendment applications. This SRP does not

preclude licensees or applicants from suggesting alternative approaches to those specified in

the SRP to demonstrate compliance with applicable regulations. 

In reviewing a license application, renewal application, or license amendment for a fuel cycle

facility, the staff must determine whether there is reasonable assurance that the facility can and

will be operated in a manner that will not be inimical to the common defense and security, and

will adequately protect the health and safety of workers, the public, and the environment. To

carry out this responsibility, the staff evaluates the information that the applicant provides and,

through independent assessments, determines whether the applicant has proposed an
Introduction xv NUREG-1520

adequate safety program that is compliant with regulatory requirements. To assist the staff in

carrying out this responsibility, this SRP clearly states and identifies those standards, criteria,

and bases that the staff will use in reaching licensing decisions. 

An applicant submits a complete description of the safety program for the possession and use

of SNM to show how it will ensure compliance with the applicable requirements. The safety

program must be described in sufficient detail to permit the staff to determine with reasonable

assurance that the facility is designed and will be operated without undue risk to the health and

safety of workers or the public. Before submitting a program description, an applicant should

have analyzed the facility in sufficient detail to conclude that it is designed and can be operated

safely. 

The requirements in 10 CFR 70.22, 10 CFR 70.23, and Subpart H to 10 CFR Part 70 specify, in

general terms, the information to be supplied in a safety program description. This SRP

supersedes and replaces draft Regulatory Guide 3.52, “Standard Format and Content for the

Health and Safety Sections of License Renewal Applications for Uranium Processing and Fuel

Fabrication.” As such, this SRP identifies the specific information to be submitted by an

applicant and evaluated by the staff. Prospective applicants should study the topic areas treated

in the chapters of this SRP and the sections within each chapter (specifically, the sections

headed “Areas of Review” and “Acceptance Criteria”). To facilitate the staff’s review, a license

application should contain a safety program description that addresses the contents of this SRP

in the same order as presented in this document. Material submitted in one location in a license

application may be referenced at another location to avoid unnecessary duplication. 

In accordance with 10 CFR 70.62 and 70.65, applicants are also required to submit an ISA

Summary in conjunction with the application. However, the ISA Summary will not be

incorporated in the license or license amendment issued by the NRC.

This SRP provides information and guidance to assist the licensing staff and the applicant in

understanding the underlying objectives of the regulatory requirements, the relationships among

NRC requirements, the licensing process, the major guidance documents that the NRC staff

has prepared for licensing fuel cycle facilities, and information about aspects of the staff review

process set out in individual SRP sections. Staff analyses are intended to provide regulatory

confirmation of reasonable assurance of safe design and operation. A staff determination of

reasonable assurance leads to a decision to issue or renew a license or to approve an

amendment. If the staff determines that an application contains inadequate descriptions or

commitments, the staff will inform the applicant of what is needed and the basis on which the

determination was made.

The “Acceptance Criteria” delineated in this SRP are intended to communicate the underlying

objectives, but do not represent the only means of satisfying those objectives. An applicant

should tailor its safety program to the particular features of its facility. If an applicant chooses

approaches other than those presented in this SRP, the applicant should identify the portions of

its license application that differ from the design approaches and acceptance criteria of the SRP,

and should document how the proposed alternatives provide an acceptable method of complying

with the Commission's regulations. The staff retains the responsibility to make an independent

determination concerning the adequacy of the applicants’s proposed approaches.

Each SRP chapter is structured to include the review (1) purpose, (2) responsibility, (3) areas,

(4) acceptance criteria, (5) procedure, (6) findings, and (7) references.
Introduction xxi NUREG-1520

Purpose of Review

This section presents a brief statement of the purpose and objectives of reviewing the subject

areas. It emphasizes the staff’s evaluation of the ways the applicant will achieve identified

performance objectives and ensures (through the review) that the applicant has used a multidisciplinary, systems-oriented approach to establish designs, controls, and procedures within

individual technical areas.

Responsibility for Review

This section identifies the NRC organization and individuals (by function) who are responsible for

evaluating the specific subject or functional area. If reviewers with expertise in other areas are to

participate in the evaluation, they also are identified by function. In general, the licensing project

manager has responsibility for the total review product, which is referred to as a safety

evaluation report (SER). However, an identified technical specialist will have primary

responsibility for a particular review topic (usually an SRP chapter), and one or more specialists

may have supporting responsibility. The overall application review is performed by this team of

specialist reviewers. Although they individually perform their review tasks, the reviews are

extensively coordinated and integrated to ensure consistency in approach and to promote riskinformed reviews. The licensing project manager oversees and directs the coordination of the

reviewers. The reviewers’ immediate line management has the responsibility to ensure that

qualified reviewers perform an adequate review.

Areas of Review

This section describes the topics, functions, systems, components, analyses, applicant

commitments, data, or other information that should be reviewed as part of the given subject

area of the license application. Because this section identifies information to be reviewed in

evaluating the adequacy of the application, it identifies the acceptable content of an applicant's

submittal in the areas discussed. The areas of review identified in this section obviate the need

for a separate standard format and content guide.

The topics identified in this section also set the content of the next two sections of the SRP, i.e.

“Acceptance Criteria” and “Review Procedures”, which should address, in the same order, the

topics set forth in this section as areas to be reviewed. This section also identifies the

information needed or the review expected from other NRC individuals to permit the individual

charged with primary review responsibility to complete the review. 

Acceptance Criteria

This section defines a set of applicable NRC acceptance criteria on the basis of regulatory

requirements, and these collectively establish the basis for assessing the acceptability of the

applicant's commitments relative to the design, programs, or functions within the scope of the

particular SRP section. Technical bases consist of specific criteria, such as NRC regulations,

regulatory guides, NUREG reports, and industry codes and standards. As such, the acceptance

criteria present positions and approaches that are acceptable to the staff. As noted above, they

are not considered the only acceptable positions or approaches, and others may be proposed by

an applicant.
Introduction xxii NUREG-1520

The NRC staff will interpret applicant commitments to follow an industry standard as a

commitment to adhere to all “shall” statements in the standard. Suggestions and

recommendations in the standards (so called “should” statements) will not be considered by the

staff as binding commitments by the applicant, unless the applicant specifically states an intent

to treat the “should” statements as binding commitments (i.e., treat as if they are “shall”

statements). The applicant may make such commitments as part of the description of the

safety program basis. If the staff finds that a definitive commitment to a “should” statement is

necessary to provide adequate protection, the reviewer will raise this as an issue in any request

for additional information (RAI) on specific licensing actions. However, applicants should note

that some industry or consensus standards specifically direct users to provide justifications for

not abiding by recommendations contained in the standards. For example, American Nuclear

Society (ANS) Standard 8.1, which relates to nuclear criticality safety, states that “when

recommendations are not implemented, justification shall be provided,” thus effectively mixing

“should” and “shall” statements. In such instances, applicants should be prepared to justify any

decisions to not abide by recommendations contained in the standards. 

This SRP presents acceptance criteria for each technical function area (e.g., nuclear criticality

safety, fire safety, radiation safety) and the management measures (e.g., configuration

management, maintenance, audits, and assessments) that an applicant uses to provide a level

of protection commensurate with the accident risk inherent in the proposed process activities. 

For example, at process stations (or for an entire process or sub-process) for which the

inherent risk to workers, the public, or the environment is demonstrably small, the applicant

needs to provide only those design and operating controls that ensure that small risk. The key

element in the staff’s evaluation is the applicant’s adequate demonstration of acceptable control

of risk, which then supports a competent and informed review by the NRC staff. 

Review Procedures

This section describes how the staff will perform the review. It generally describes procedures

that the reviewer should follow to achieve an acceptable scope and depth of review and to obtain

reasonable assurance that the applicant has provided appropriate commitments to ensure that it

will operate the facility safely. This could include identifying which licensee commitments the

reviewer needs to verify, and could include directing the reviewer to coordinate with others

having review responsibilities for other portions of the application than those assigned to the

reviewer. This section should provide whatever procedural guidance is necessary to evaluate

the applicant's level of achievement of the acceptance criteria.

Evaluation Findings

This section presents the type of positive conclusion that is sought, for the particular review

area, to support a decision to grant a license or amendment. The review must be adequate to

permit the reviewer to support this conclusion. For each section, a conclusion of this type will be

included in the staff's SER, in which the staff publishes the results of its review. The SER will

also contain a description of the review, including aspects that received special emphasis,

matters that the applicant modified during the review, matters that require additional information

or will be resolved in the future, aspects where the facility’s design or the applicant's proposals

deviate from the criteria in the SRP, and the bases for any deviations from the SRP or proposed

exemptions from the regulations. 
Introduction xxiii NUREG-1520

Staff may recommend in the SER, license conditions to address any issues that were not

previously resolved by an applicant's commitments. Such conditions are discussed with an

applicant before issuing the license (or license amendment) and become commitments to

performance in addition to those commitments that the applicant presented in the application.

References

This section lists references that the staff should consult during the review process. However,

depending on the action and approaches proposed by the applicant, they may not always be

relevant to the review.

Appendix A to this SRP provides additional guidance on filing standards for applications.
General Information 1-1 NUREG-1520

1 GENERAL INFORMATION

1.1 FACILITY AND PROCESS DESCRIPTION

1.1.1 Purpose of Review

The purpose of this review is to ascertain whether an application for a new, renewed, or

amended license includes an overview of the facility layout and a summary description of its

manufacturing processes. This overview will be used by all reviewers, NRC managers, and the

general public to understand the purpose of the facility and to obtain an overview of the design

of its processes. A more detailed description of the facility and its manufacturing processes is

contained in the ISA Summary.

1.1.2 Responsibility for Review

Primary: Licensing Project Manager

Secondary: None

Supporting: None

1.1.3 Areas of Review

The staff should review the general facility and process descriptions provided by the applicant,

which should include (1) scaled drawings showing the locations of facility buildings and other

major structures, hazardous materials storage areas, onsite roadways, railroad spurs or sidings,

and major ingress and egress routes for the site; (2) a text index with descriptive titles that

indicate the purpose of each feature; (3) the interrelationships among the features; (4) the

relationship of facility features to site features; (5) a narrative description of the flow of licensed

material through the facility’s manufacturing processes; and (6) the proximity of facility buildings

to the site boundary and nearby populations. This information should be consistent with that

presented in Chapter 9 “Environmental Protection” and Chapter 8 “Emergency Management” of

this SRP.

1.1.4 Acceptance Criteria

1.1.4.1 Regulatory Requirements

The regulations applicable to the areas of review in this SRP are 10 CFR 70.22, “Contents of

Applications,” and 10 CFR 70.65(b)(1), (2), and (3), “Additional Content of Applications.” 

1.1.4.2 Regulatory Guidance

There are no regulatory guides that apply to a general facility description for a fuel cycle facility. 

1.1.4.3 Acceptance Criteria

The reviewer will determine that the applicant’s presentations with respect to this section of the

SRP are acceptable if the following criteria are met:
General Information 1-2 NUREG-1520

(1) The application presents information at a level of detail that is appropriate for general

familiarization and understanding of the proposed facility and processes. This information

may be less detailed than that presented in the ISA Summary.

(2) The application summarizes the facility information contained in the ISA Summary. This

includes descriptions of the overall facility layout on scaled drawings, including the site’s

geographical features and facility structural features (such as buildings, towers, and tanks)

and transportation right-of-ways. In addition, the summary should describe the relationship

of specific facility features to the major processes that will be ongoing at the facility.

(3) The major chemical or mechanical processes involving licensable quantities of SNM are

described in summary form, based in part on information presented in the ISA Summary. 

This description should include reference to the building locations of major process

components; brief descriptions of the process steps; the chemical forms of SNM in process;

the maximum amounts of SNM in process in various building locations; and the types,

amounts, and discharge points of waste materials discharged to the environment from the

processes. 

(4) The application presents a summary identification of the raw materials, by-products, wastes,

and finished products of the facility. This information should include data regarding

expected levels of trace impurities or contaminants (particularly fission products or

transuranic elements) characterized by identity and concentration. In addition, this

summary should identify the proposed possession at the facility of any moderator or

reflector with special characteristics, such as beryllium or graphite.

1.1.5 Review Procedures

1.1.5.1 Acceptance Review

To begin the staff’s review the primary reviewer determines whether the application includes the

topics discussed in Section 1.1.3, “Areas of Review.” If the reviewer identifies significant

deficiencies in the application, the staff should request the applicant to submit additional

material before the start of the safety evaluation. With the complete submittal available, the

reviewer should examine the facility and process descriptions and assess their acceptability by

comparison with the acceptance criteria in Section 1.1.4.3 and consistency with information in

the ISA Summary. 

1.1.5.2 Safety Evaluation

The material to be reviewed is informational in nature, and no technical analysis is required. In

addition the information to be reviewed is used only as background for the more detailed

descriptions in later sections of the application. Therefore, the primary reviewer confirms

whether the descriptive information presented is consistent with the information presented in

the ISA Summary.

1.1.6 Evaluation Findings

If the license application provides sufficient information and the regulatory acceptance criteria in

Section 1.1.4.3 are appropriately satisfied, the staff concludes that this evaluation is complete. 

The reviewer writes material suitable for inclusion in the SER prepared for the entire

application. The report includes a summary statement of what was reviewed and why the

reviewer finds the submittal acceptable. The staff can document the review as follows:
General Information 1-3 NUREG-1520

The staff has reviewed the general facility description for [name of facility] according to

Section 1.1 of the Standard Review Plan. The applicant has adequately described 

(1) the facility and processes so that the staff has an overall understanding of the

relationships of the facility features and (2) the function of each feature. The applicant has

cross-referenced its general description with the more detailed descriptions elsewhere in the

application. The staff concludes that the applicant has complied with the general

requirements of 10 CFR 70.22, “Contents of Applications,” 10 CFR 70.60, “Applicability,”

and 10 CFR 70.65(b)(1), (2), and (3), “Additional Content of Applications,” as applicable to

this section. 

1.1.7 References

Code of Federal Regulations, Title 10, Part 70, “Domestic Licensing of Special Nuclear

Material,” U.S. Government Printing Office, Washington, DC.

1.2 INSTITUTIONAL INFORMATION

1.2.1 Purpose of Review

The purpose of this review is to establish whether the license application includes adequate

information identifying the applicant, the applicant’s characteristics, and the proposed activity.

1.2.2 Responsibility for Review

Primary: Licensing Project Manager

Secondary: None

Supporting: Office of the General Counsel; Office of Administration/Division of

Security

1.2.3 Areas of Review

Information provided for review should include the identity and address of the applicant’s facility

and corporate headquarters; corporate information sufficient to show the relationship of the

applicant’s organization relative to other corporate entities; the existence and extent of foreign

ownership or influence; financial information sufficient to indicate the resources available to the

applicant to pursue the activities for which the license is sought; the site location as legally

described in land records; a description of each proposed licensed activity in the form of

requested authorized uses; the type of license being applied for; and the type, quantity, and

form(s) of material(s) proposed to be used at the licensed facility.

1.2.4 Acceptance Criteria

1.2.4.1 Regulatory Requirements

The regulations applicable to the areas of review in this SRP are 10 CFR 70.22, “Contents of

Applications,” 10 CFR 70.65(b)(1), (2), and (3), “Additional Contents of Applications,” 

10 CFR 70.33, “Renewal of Licenses,” and 10 CFR Part 95, “Security Facility Approval and

Safeguarding of National Security Information and Restricted Data.”
General Information 1-4 NUREG-1520

1.2.4.2 Regulatory Guidance

There are no regulatory guides that apply to institutional information for a fuel cycle facility.

1.2.4.3 Regulatory Acceptance Criteria

The application is acceptable if the following criteria are met:

(1) Corporate Identity

The applicant has furnished its full name and address. The address of the fuel cycle facility

is provided if it is different from that of the applicant. If the application is for renewal, the

applicant identifies the number of the license to be renewed. A full description of the facility

site location (State, county, and municipality) is given. The State where the applicant is

incorporated or organized and the location of the principal office are indicated. If the

applicant is a corporation or other entity, the names and citizenship of its principal officers

are provided. The application should include any information known to the applicant

concerning the control or ownership, if any, exercised over the applicant by any alien,

foreign corporation, or foreign government. Primary ownership and relationships to other

components of the same ownership are explicitly described. The presence and operations

of any other company on the site to be licensed are fully described. 

(2) Financial Qualifications

A description of financial qualifications demonstrates the applicant’s current and continuing

access to the financial resources necessary to engage in the proposed activity in

accordance with 10 CFR 70.22(a)(8) and 10 CFR 70.23(a)(5). 

(3) Type, Quantity, and Form of Licensed Material

The application identifies the elemental name, maximum quantity, and specifications,

including the chemical and physical form(s), of the SNM that the applicant proposes to

acquire, deliver, receive, possess, produce, use, transfer, or store. For each SNM the

specifications include the isotopic content and amount of enrichment by weight percent. 

(4) Authorized Uses

A summary, non-technical narrative description is provided for each activity or process in

which the applicant proposes to acquire, deliver, receive, possess, produce, use, process,

transfer, or store SNM. The authorized uses of SNM proposed for the facility are described

and are consistent with the Atomic Energy Act of 1954, et seq. The description is

consistent with more detailed process descriptions submitted as part of the ISA Summary

reviewed under Chapter 3 of this SRP.

If the application is for a renewal, the applicant states the period of time for which license

renewal is requested, and why the renewal application should be considered timely in

accordance with 10 CFR Part 70.
General Information 1-5 NUREG-1520

(5) Special Exemptions or Special Authorizations

Specific requests for exemptions or authorizations of an unusual nature should be listed in

this section and justified in the appropriate technical section of the application.

(6) Security of Classified Information

If applicable, the applicant has requested and received a facility security clearance in

accordance with 10 CFR Part 95.

1.2.5 Review Procedures

1.2.5.1 Acceptance Review

To begin the staff’s review, the primary reviewer determines whether the application is complete

and addresses each issue in Section 1.2.3, “Areas of Review.” If the reviewer identifies

significant deficiencies in the application, the staff will request that the applicant submit

additional material before the start of the safety evaluation.

1.2.5.2 Safety Evaluation

The material to be reviewed is for the most part informational in nature, and detailed technical

analysis is generally not required beyond the acceptance criteria. The reviewer requests review

assistance, as needed, from the Division of Security and the Office of the General Counsel in

the review of corporate and financial information.

1.2.6 Evaluation Findings

If the information provided is consistent with the guidance of this SRP, the staff will conclude

that this evaluation is complete. The staff can document its review as follows:

The staff has reviewed the institutional information for [name of facility] according to

Section 1.2.of the Standard Review Plan. On the basis of the review, the NRC staff has

determined that the applicant has adequately described and documented the corporate

structure and financial information, and is in compliance with those parts of 

10 CFR 70.22 and 70.65 related to other institutional information. In addition, in accordance

with 10 CFR 70.22(a)(2) and (4), the applicant has adequately described the types, forms,

quantities, and proposed authorized uses of licensed materials to be permitted at this facility

as follows:

Material Form Quantity Authorized Use(s)

The applicant’s proposed activities are consistent with the Atomic Energy Act. The

applicant has provided all institutional information necessary to understand the

ownership, financial qualifications, location, planned activities, and nuclear materials to

be handled in connection with the requested license.

1.2.7 References

Code of Federal Regulations, Title 10, Part 70, “Domestic Licensing of Special Nuclear

Material,” U.S. Government Printing Office, Washington, DC.
General Information 1-6 NUREG-1520

1.3 SITE DESCRIPTION

1.3.1 Purpose of Review

The purpose of this review is to determine whether the information provided by an applicant

adequately describes the geographic, demographic, meteorologic, hydrologic, geologic, and

seismologic characteristics of the site and the surrounding area. The site description is a

summary of the information that the applicant used in preparing the environmental report,

emergency plan, and ISA Summary.

1.3.2 Responsibility for Review

Primary: Licensing Project Manager

Secondary: ISA Reviewer, Environmental Protection Reviewer, and Emergency Plan 

Reviewer

Supporting: Fuel Facility Inspection Staff 

1.3.3 Areas of Review

The information that the applicant presented in this section is summarized from the information

presented in more detail in the applicant’s environmental report, emergency plan, and ISA

Summary. The information that the NRC staff will review includes the following (as appropriate

for the facility being reviewed):

(1) Site Geography

a. site location: State, county, municipality, topographic quadrangle (eight 71/2 minute 

quadrants), site boundary, and controlled area boundary

b. major nearby highways

c. nearby bodies of water

d. any other significant geographic feature that may impact accident analysis within 1 

mile of the site (e.g., ridges, valleys, specific geologic structures)

(2) Demographics

a. latest census results for area of concern

b. description, distance, and direction to nearby population centers

c. description, distance, and direction to nearby public facilities (e.g., schools, hospitals,

parks)

d. description, distance, and direction to nearby industrial areas or facilities that may

present potential hazards (including other nearby nuclear facilities)

e. uses of land within 1 mile of the facility (i.e., residential, industrial, commercial,

agricultural)

f. uses of nearby bodies of water

(3) Meteorology

a. primary wind directions and average wind speeds
General Information 1-7 NUREG-1520

b. annual amount and forms of precipitation, as well as the design-basis values for

accident analysis of maximum snow or ice load and probable maximum precipitation

c. type, frequency, and magnitude of severe weather (e.g., lightning, tornado, hurricane)

and design-basis event summary descriptions for accident analysis

(4) Hydrology

a. characteristics of nearby rivers, streams, and bodies of water as appropriate

b. depth to the water table and potentiometric surface map

c. groundwater flow direction and velocity for the site

d. characteristics of the uppermost aquifer 

e. design-basis flood events used for accident analysis

(5) Geology

a. characteristics of soil types and bedrock

b. design-basis earthquake magnitudes and return periods used for accident analysis 

c. description of other geologic hazards (e.g., mass wasting)

1.3.4 Acceptance Criteria

1.3.4.1 Regulatory Requirements

Regulations applicable to the areas of review in this SRP are 10 CFR 70.22, “Contents of

Applications.”

1.3.4.2 Regulatory Guidance

There are no regulatory guides that apply to site descriptions for a fuel cycle facility.

1.3.4.3 Regulatory Acceptance Criteria

The site description will be considered acceptable if it includes the following:

(1) The summary briefly describes site geography, including its location relative to prominent

natural and man-made features (such as mountains, rivers, airports, population centers,

schools, commercial and manufacturing facilities). The summary also describes the site

boundary and the controlled area. 

(2) The summary provides population information on the basis of the most current available

census data to show population distribution as a function of distance from the facility.

(3) The application addresses appropriate meteorologic data, including a summary of designbasis values for accident analysis of maximum snow or ice load, and probable maximum

precipitation, as may be developed by the applicant and presented in the ISA Summary. 

The applicant presents appropriate design-basis values for lightning, high winds, tornado,

hurricane, and other severe weather conditions that are applicable to the site.

(4) The application includes a summary description of the hydrology and geology (including

seismicity) for the area, and cites the design-basis flood event for which the facility may be 
General Information 1-8 NUREG-1520

safely shut down. The application also provides earthquake accelerations for the site

associated with a 250-year and 500-year earthquake. 

(5) The applicant’s descriptions are consistent with the more detailed information presented

within the ISA Summary, the environmental report, and the emergency plan, if applicable. 

1.3.5 Review Procedures

1.3.5.1 Acceptance Review 

The staff will initially determine whether the application is complete and addresses all topics

discussed in Section 1.3.3, “Areas of Review.” The information in this section provides a

general summary of the bases reported in the ISA Summary and is consistent with the

applicant’s environmental report and emergency plan. The applicant may include references to

the more detailed data used to complete evaluations in the ISA Summary.

If the reviewer identifies significant deficiencies in the application, the staff will request that the

applicant submit additional material before the start of the safety evaluation.

For license renewals, the details necessary to support the information in the site description

summary may be referenced to a prior submittal or material included elsewhere in the renewal

application.

1.3.5.2 Safety Evaluation

The material described in this section of the SRP is informational, summarizing that contained

in the ISA Summary, environmental report, emergency plan, and other documents referenced

by the applicant. No technical analysis is required, as the primary reference for the information

is the ISA Summary. The applicant may also need to update this section to verify any

information changes made in response to the staff's environmental, emergency management,

and ISA Summary reviews.

1.3.6 Evaluation Findings

If the license application provides sufficient information and is consistent with the guidance in

this SRP, the staff will conclude that this evaluation is complete and the applicant’s site

description is acceptable. The staff can document its review as follows:

The staff has reviewed the site description for [name of facility] according to Section 1.3

of the Standard Review Plan. The applicant has adequately described and summarized

general information pertaining to (1) the site geography, including its location relative to

prominent natural and man-made features such as mountains, rivers, airports,

population centers, schools, and commercial and manufacturing facilities; (2) population

information on the basis of the most current available census data to show population

distribution as a function of distance from the facility; (3) meteorology, hydrology, and

geology for the site; and (4) applicable designbasis events. The reviewer verified that

the site description is consistent with the information used as a basis for the

environmental report, emergency management plan, and ISA Summary.
General Information 1-9 NUREG-1520

1.3.7 References

Code of Federal Regulations, Title 10, Part 70, “Domestic Licensing of Special Nuclear

Material,” U.S. Government Printing Office, Washington, DC.

Accession #: ML01370271
Organization and Administration 2-1 NUREG-1520

2 ORGANIZATION AND ADMINISTRATION

2.1 PURPOSE OF REVIEW

The purpose of the review of the applicant's organization and administration is to ensure that

the proposed management policies will provide reasonable assurance that the licensee plans,

implements, and controls site activities in a manner that ensures the safety of workers, the

public, and the environment. The review also ensures that the applicant has identified and

provided adequate qualification descriptions for key management positions.

2.2 RESPONSIBILITY FOR REVIEW

Primary: Licensing Project Manager

Secondary: None

Supporting: Primary Reviewers for other SRP Chapters (e.g., technical area chapters

and management measures chapters) and the Fuel Facility Inspection

Staff

2.3 AREAS OF REVIEW

The organizational structure and associated administrative program proposed by the applicant

should include administrative policies, procedures and management policies, qualifications of

key management positions, and describe how these will provide reasonable assurance that the

health, safety, and environmental (HS&E) protection functions will be effective.

For new facilities, or already licensed facilities undergoing major modifications, to ensure that all

HS&E functions and standards are met, the applicant should describe the comprehensive

management policies and procedures that will be used to manage and closely monitor the

facility design, engineering, and construction.

The application should address how the management policies ensure the establishment and

maintenance of design and operations. The administrative and management policies should

describe the relationships among major facility safety functions such as the ISA, management

measures for items relied on for safety (IROFS), radiation safety, nuclear criticality safety, fire

safety, chemical safety, environmental monitoring, and emergency planning. The applicant

should also describe its qualification criteria with regard to education, training, and experience

for key management positions. Management positions for which such criteria should be

described include the facility manager, operations manager, shift supervisor, and managers for

various safety and environmental disciplines. Alternative named management positions could

be proposed. Qualification criteria should be described generally, in terms of academic

credentials, formal continuing education, and work experience. For example, “...bachelor’s

degree in nuclear engineering or related scientific or engineering field, with 5 years experience

managing the operations of a nuclear fuel manufacturing facility.”
Organization and Administration 2-2 NUREG-1520

2.4 ACCEPTANCE CRITERIA

2.4.1 Regulatory Requirements

10 CFR Part 70.22, 70.23, and 70.62(d) require a management system and administrative

procedures for the effective implementation of HS&E functions concerning the applicant’s

corporate organization, qualifications of the staff, and adequacy of the proposed equipment,

facilities, and procedures to provide adequate safety for workers, the public, and the

environment.

2.4.2 Regulatory Guidance

There are no regulatory guides specific to the organization and administration description of

fuel cycle facilities. 

2.4.3 Regulatory Acceptance Criteria

The application is acceptable if the following criteria are met. Appropriate commitments

relevant to these criteria should be included in the applicant’s safety program description.

New Facilities or Facilities Undergoing Major Modifications (In addition to the criteria listed

below for existing facilities):

(1) The applicant has identified and functionally described the specific organizational groups

that are responsible for managing the design, construction, and operation of the facility. 

The application also includes organizational charts. 

(2) Clear, unambiguous management controls and communications exist among the

organizational units that are responsible for managing the design and construction of the

facility. 

(3) The personnel responsible for managing the design, construction, and operation of the

facility have substantive breadth and level of experience and are appropriately available. 

The qualifications, responsibilities, and authorities for key supervisory and management

positions with HS&E responsibilities are clearly defined in position descriptions that are

accessible to all affected personnel and to the NRC, upon request.

(4) The applicant has described specific plans to commission the facility’s startup and

operation, including the transition from the startup phase to operations under the direct

supervision of the applicant’s personnel responsible for safe operations.

Existing Facilities:

(1) The applicant has identified and functionally described the specific organizational groups

responsible for operating the facility and managing the development of design changes to

the facility. The application also includes organizational charts.

(2) The qualifications, responsibilities, and authorities of key supervisory and management

positions with HS&E responsibilities are clearly defined in position descriptions that are

accessible to affected persons and to the NRC, upon request. 
Organization and Administration 2-3 NUREG-1520

(3) In the organizational hierarchy, the HS&E organization(s) is independent of the operations

organization(s), allowing it to provide objective HS&E audit, review, or control activities. 

"Independent" means that neither organization reports to the other in an administrative

sense. (However, both may report to a common manager.) Lines of responsibility and

authority are clearly drawn.

(4) The individual delegated overall responsibility for the HS&E functions has the authority to

shut down operations if they appear to be unsafe and in that case, must approve restart of

shutdown operations. 

(5) The activities that are essential for effective implementation of the HS&E functions are

documented in formally approved, written procedures, prepared in compliance with a formal

document control program. 

(6) The applicant should commit to a simple mechanism, available for use by any person in the

plant, for reporting potentially unsafe conditions or activities to the HS&E organization. 

Reported concerns should be promptly investigated, assessed, and resolved.

(7) Effective lines of communication and authority among the organizational units involved in

the engineering, HS&E, and operations functions of the facility are clearly defined.

(8) The applicant has committed to establish formal management measures required to ensure

the availability and reliability of IROFS. Management measures are detailed in Chapter 11

of this SRP. 

(9) Written agreements exist with offsite emergency resources such as fire, police,

ambulance/rescue units, and medical services. This is addressed in more detail in Chapter

7, "Fire Safety," and Chapter 8, "Emergency Planning," of this SRP.

The applicant's safety program description includes commitments relevant to meeting the

acceptance criteria described above. 

2.5 REVIEW PROCEDURES

2.5.1 Acceptance Review

The primary reviewer should evaluate the application to determine whether it addresses the

“Areas of Review” discussed in Section 2.3, above. If the reviewer identifies significant

deficiencies, the staff will request that the applicant submit additional material before the start of

the safety evaluation.

2.5.2 Safety Evaluation

The primary reviewer should perform a safety evaluation with respect to the acceptance criteria

described in Section 2.4. The objective of the review is to ensure that the corporate-level

management and technical support structure, as demonstrated by organizational charts and

descriptions of functions and responsibilities, are clear with respect to assignments of primary

responsibility. The primary reviewer consults with the NRC inspection staff to verify that the

applicant’s management positions are adequately defined in terms of both numbers of persons

and their responsibilities, authorities, and required qualifications. The reviewer may visit the
Organization and Administration 2-4 NUREG-1520

site, if considered necessary to discuss and verify implementation of the acceptance criteria

with facility management.

On the basis of the foregoing, the supporting staff reviewers determine the overall acceptability

of the applicant's management system, management qualifications, organizational structure,

and administrative procedures. The reviewers should determine whether the acceptance

criteria of Section 2.4 are satisfied, and then prepare an SER in accordance with Section 2.6. 

2.6 EVALUATION FINDINGS

The staff's evaluation should verify that the license application provides sufficient information to

satisfy the regulatory requirements of Section 2.4.1 and that the regulatory acceptance criteria

in Section 2.4.3 have been appropriately considered in satisfying the requirements. On the

basis of this information, the staff should conclude that this evaluation is complete. The

reviewer should write material suitable for inclusion in the SER prepared for the entire

application. The SER should include a summary statement of what was evaluated and the

basis for the reviewers' conclusions. The staff can document the evaluation as follows:

The staff has reviewed the organization and administration for [name of facility]

according to Chapter 2 of the Standard Review Plan.

[For new facilities] The applicant has described (1) clear responsibilities and associated

resources for the design and construction of the facility and (2) its plans for managing

the project. [Insert a summary statement of what was evaluated and why the reviewer

finds the submittal acceptable.] The staff has reviewed these plans and commitments

and concludes that they provide reasonable assurance that an acceptable organization,

administrative policies, and sufficient competent resources have been established or are

committed, to satisfy the applicant's commitments for the design and construction of the

facility.

[For operating and new facilities] The applicant has described its organization and

management policies for providing adequate safety management and management

measures for the safe operation of the facility. [Insert a summary statement of what was

evaluated and why the reviewer finds the submittal acceptable.] The staff has reviewed

this information and concludes that the applicant has an acceptable organization,

administrative policies, and sufficient competent resources to provide for the safe

operation of the facility under both normal and abnormal conditions.

2.7 REFERENCES

U.S. Code of Federal Regulations, Title 10, Part 70, “Domestic Licensing of Special Nuclear

Material,” U.S. Government Printing Office, Washington, DC.

U.S. Nuclear Regulatory Commission, “Proposed Method for Regulating Major Materials

Licensees,” NUREG-1324, Sections 3.1, Organization Plan, and 3.2, Managerial Controls and

Oversight, 1992.

Accession #: ML013370277
Integrated Safety Analysis 3-1 NUREG-1520

3 INTEGRATED SAFETY ANALYSIS (ISA) AND ISA SUMMARY

3.1 PURPOSE OF REVIEW

An integrated safety analysis (ISA) identifies potential accident sequences in the facility=s

operations, designates items relied on for safety (IROFS) to either prevent such accidents or

mitigate their consequences to an acceptable level, and describes management measures to

provide reasonable assurance of the availability and reliability of IROFS. Applicants for new

licenses and persons holding 10 CFR Part 70 licenses on September 18, 2000, must perform

an ISA and submit a summary (referred to as an “ISA Summary”) to the NRC, for approval. 

The ISA Summary principally differs from the ISA by focusing on higher risk accident

sequences with consequences that could exceed the performance criteria of 10 CFR 70.61. 

The ISA Summary is a synopsis of the results of the ISA and contains information specified in 

10 CFR 70.65(b).

The ISA and supporting ISA documentation (such as piping and instrumentation diagrams,

criticality safety analyses, dose calculations, process safety information, and ISA worksheets)

are maintained at the facility. The NRC determines the acceptability of the applicant’s ISA by

reviewing a portion of the ISA documentation and any supporting documentation maintained

onsite and by reviewing and approving the applicant=s ISA Summary which, although not part of

the license application, is placed on the public docket. Neither the ISA nor the ISA Summary is

incorporated as part of the license.

Reviewers must confirm that an ISA Summary meets the regulatory requirements of 

10 CFR 70.65 and, specifically, that suitable IROFS and management measures have been

designated for higher-risk accident sequences and that programmatic commitments to maintain

the ISA and ISA Summary are acceptable. An applicant may submit, for NRC approval, one

ISA Summary for the entire facility, or multiple ISA summaries for individual processes (or

groups of processes) in the facility as they are completed. Reviews of ISA summaries may

necessitate examination of the ISA and its supporting documentation to confirm the

underpinnings of calculations, conclusions, and components of safety programs. 

This chapter provides guidance for NRC’s review of two types of information submitted by

applicants: 

(1) commitments regarding the applicant=s safety program including the ISA, pursuant to the

requirements of 10 CFR 70.62

(2) ISA summaries submitted in accordance with 10 CFR 70.62(c)(3)(ii) and 70.65

In the case of license applications (either initial or renewal), applicants would submit both types

of information. In the case of a license amendment, an applicant may submit either or both

types of information, as needed, to address the areas amended.

3.1.1 Safety Program and ISA Commitments

The purpose of the review of commitments related to the safety program, including the ISA, as

presented in the license application, renewal, or amendment, is to determine with reasonable

assurance that the applicant will accomplish the requirements of 10 CFR 70.61; 70.62(a)(1),

(2) and (3); 70.62(c)(1) and (2); 70.62(d); 70.64 for new facilities; and 70.72 for changes

requiring updates of the ISA.
Integrated Safety Analysis 3-2 NUREG-1520

3.1.2 ISA Summary

The purpose of the review of the ISA Summary is to establish reasonable assurance that the

applicant has performed the following tasks:

(1) Conducted an ISA of appropriate detail for each applicable process, using methods and

staff adequate to achieve the requirements of 10 CFR 70.62(c)(1) and (2).

(2) Identified and evaluated, in the ISA, all credible events (accident sequences) involving

process deviations or other events internal to the facility (e.g., explosions, spills, and fires);

and credible external events that could result in facility-induced consequences to workers,

the public, or the environment, that could exceed the performance requirements of 

10 CFR 70.61. As a minimum, external events normally include the following:

a. natural phenomena events such as floods, high winds, tornadoes, and earthquakes

b. fires external to the facility

c. transportation accidents and accidents at nearby industrial facilities

(3) Designated engineered and administrative IROFS, and correctly evaluated the set of

IROFS addressing each accident sequence, as providing reasonable assurance, through

preventive or mitigative measures, and through application of supporting management

measures (discussed in Chapter 11) that the performance requirements of 10 CFR 70.61

are met.

3.2 RESPONSIBILITY FOR REVIEW

Primary: Assigned Licensing Reviewer

Secondary: Technical Specialists in Specific Areas

Supporting: Fuel Facility Inspectors

3.3 AREAS OF REVIEW

This chapter addresses two types of submittals, including (1) those containing descriptive

commitments regarding the safety program, including the ISA, and (2) ISA summaries. The

descriptive commitments regarding the safety program should be found in license applications,

renewals, and amendments. ISA summaries may be submitted for an entire existing facility, a

new facility, a new process, or altered processes requiring revision of the ISA.

The safety program and ISA commitments and descriptions to be reviewed consist of 

(1) process safety information [10 CFR 70.62(b)], (2) methods used to perform the ISA, 

(3) qualifications of the team performing the ISA [10 CFR 70.62(c)(2)], (4) methods of

documenting and implementing the results of the ISA, (5) procedures to maintain the ISA

current when changes are made to the facility, and (6) management measures 

[10 CFR 70.62(d)]. These commitments and descriptions, as appropriate, will primarily be

documented in the license application within an ISA chapter. However, pursuant to Chapter 11

of this SRP, commitments and descriptions regarding management measures will be in a

separate chapter of an application.
Integrated Safety Analysis 3-3 NUREG-1520

The results of ISA analyses performed for compliance with Subpart H of 10 CFR Part 70 are

presented in an ISA Summary. This ISA Summary may be submitted with an application for a

new license, a license renewal, or a license amendment, but is not to be incorporated as part of

the license.

The ISA Summary submitted to the NRC, and portions of the ISA and ISA documentation

maintained onsite, will be reviewed to determine the adequacy of the applicant's ISA. The

contents of the ISA Summary are specified in 10 CFR 70.65 and include the following nine

topics:

(1) general description of the site

(2) general description of the facility

(3) description of facility processes, hazards, and types of accident sequences

(4) demonstration of compliance with 10 CFR 70.61 performance requirements

(5) description of the ISA team qualifications and ISA methods 

(6) descriptive list of IROFS

(7) description of acute chemical exposure standards used

(8) descriptive list of sole IROFS

(9) definition of the terms “credible,” “unlikely,” and “highly unlikely”

The ISA (referred to here as the standalone document that presents the results of the ISA) and

supporting ISA documentation (e.g., piping and instrumentation drawings, engineered IROFS

boundary descriptions, criticality safety analyses, dose calculations, process hazards analysis,

process safety information, ISA worksheets, etc.) will be maintained at the facility site. The

reviewer(s) may need to consult the ISA and its supporting documentation at the facility site to

establish the completeness and acceptability of the ISA or, in the case of an existing facility, to

visit the site to fully understand a process operation. For example, the reviewer(s) should

confirm that low-risk accident sequences that were not reported in the ISA Summary were

correctly identified and analyzed in the ISA.

3.3.1 Safety Program and ISA Commitments

The NRC reviews the application to determine whether the applicant=s commitments to

establish a safety program and to perform and maintain an ISA are adequate. In the following,

the phrase Aprocess node@ or Aprocess@ is used to refer to a single, reasonably compact piece

of equipment or workstation where a single unit process or processing step is conducted. A

typical fuel cycle facility is divided into several major process lines or areas, each consisting of

many process nodes. The areas of review for ISA commitments are as follows:

(1) The applicant's description of, and commitments to, a method for maintaining a current and

accurate set of process safety information, including information on the hazardous

materials, technology, and equipment used in each process. The applicant should explain

this activity in detail in the description of its configuration management program 

(Section 11.1, AConfiguration Management@).

(2) The applicant's description of, and commitments to, requirements for ISA team training and

qualifications (Section 11.4, ATraining and Qualification@) for those individuals who will

conduct and maintain the ISA and ISA Summary.

(3) The applicant's description of, and commitments to, ISA methods, method selection criteria,

or specific methods to be used for particular classes of process nodes (usually process
Integrated Safety Analysis 3-4 NUREG-1520

workstations). The review of the ISA method(s) includes evaluating the applicant=s methods

in the following specific areas:

a. hazard identification

b. process hazard analysis (accident identification)

c. accident sequence construction and evaluation

d. consequence determination and comparability to 10 CFR 70.61

e. likelihood categorization for determining compliance with 10 CFR 70.61

(4) The applicant's description of, and commitments to, management procedures for

conducting and maintaining the ISA. Specific review areas include the following applicant

procedures: 

a. performance of, and updates to, the ISA 

b. review responsibility 

c. ISA documentation

d. reporting of ISA Summary changes per 10 CFR 70.72(d)(1) and (3) 

e. maintenance of ISA records per 10 CFR 70.62(a)(2) 

3.3.2 ISA Summary and ISA Documentation

The NRC reviews the ISA Summary and, if necessary, the ISA and supporting ISA

documentation to determine whether there is reasonable assurance that the applicant has

performed a systematic evaluation of the hazards and has identified credible accident

sequences, IROFS and management measures that satisfy the performance requirements of 

10 CFR 70.61. The NRC confirms that credible accidents that result in a release of radioactive

material, a nuclear criticality event, or any other exposure to radiation resulting from use of

licensed material that exceeds the exposure limits stated in 10 CFR 70.61, are “highly unlikely”

or “unlikely,” as appropriate. In addition, the NRC reviews accidents involving hazardous

chemicals produced from licensed materials. That is, chemicals that are licensed materials or

have licensed materials as precursor compounds, or substances that physically or chemically

interact with licensed materials and that are toxic, explosive, flammable, corrosive, or reactive to

the extent that they endanger life or health. These include substances that are commingled

with licensed material or are produced by a reaction with licensed material. If a chemical

accident has the potential to cause, or reduce protection from, a radiation exposure accident,

then it also must be addressed. On the other hand, accident sequences having unmitigated

consequences that will not exceed the performance requirements of 10 CFR 70.61(c), once

identified as such, do not require reporting in the ISA Summary.

The areas of review for the ISA Summary are as follows:

(1) Site: The site description in the ISA Summary (see Section 1.3, ASite Description@) focuses

on those factors that could affect safety, such as geography, meteorology (e.g., high winds

and flood potential), seismology, demography, and nearby industrial facilities and

transportation routes.

(2) Facility: The facility description in the ISA Summary focuses on features that could affect

potential accidents and their consequences. Examples of these features are facility

location, facility design information, and the location and arrangement of buildings on the

facility site.
Integrated Safety Analysis 3-5 NUREG-1520

(3) Processes, Hazards and Accident Sequences: The process description in the ISA

Summary addresses each process that was analyzed as part of the ISA. Specific areas

reviewed include basic process function and theory, functions of major components and

their operation, process design and equipment, and process operating ranges and limits. 

This description must also include a list of the hazards (and interactions of hazards) for

each process and the accident sequences that could result from such hazards and for

which unmitigated consequences could exceed the performance requirements of 

10 CFR 70.61.

(4) Demonstration of Compliance with 10 CFR 70.61: For each applicable process, this section

presents the following information that should be developed in the ISA to demonstrate

compliance with the performance criteria of 10 CFR 70.61:

a. postulated consequences and comparison to the consequence levels identified in 

10 CFR 70.61, as well as information, such as inventory and release path factors

supporting the results of the consequence evaluation 

b. information showing how the applicant established the likelihoods of accident

sequences that could exceed the performance requirements of 10 CFR 70.61

c. information describing how designated IROFS protect against accident sequences that

could exceed the performance requirements of 10 CFR 70.61

d. information on management measures applied to the IROFS (addressed in greater

detail in Chapter 11)

e. information on how the criticality monitoring requirements of 10 CFR 70.24 are met

f. if applicable, how the baseline design criteria of 10 CFR 70.64 are addressed

(5) Team Qualifications and ISA Methods: This section should discuss the applicant=s ISA

team qualifications and ISA methods, as described in the ISA Summary. (If methods are

adequately described in the license application, there will be no need to duplicate this

information in the ISA Summary. Specific examples of the application of ISA methods

should be included in the ISA Summary to enable the reviewer(s) to understand their

selection and use.)

(6) List of IROFS: This list describes the IROFS for all intermediate- and high-consequence

accidents in sufficient detail to understand their safety function(s).

(7) Chemical Consequence Standards: This discussion identifies the applicant=s quantitative

standards for assessing the chemical consequence levels specified in 10 CFR 70.61, as

described in the ISA Summary.

(8) List of Sole IROFS: This list identifies those IROFS that are the sole item preventing or

mitigating an accident for which the consequences could exceed the performance

requirements of 10 CFR 70.61. 

(9) Definitions of “Unlikely”, “Highly Unlikely” and “Credible”: The applicant must define the

terms “unlikely,” “highly unlikely,” and “credible,” as used in the ISA Summary.
Integrated Safety Analysis 3-6 NUREG-1520

10 CFR 70.65(b) lists the types of information required to be submitted in an ISA Summary. 

This includes generic information, such as site description, ISA methods, and ISA team

qualifications. This also includes process-specific information, such as a list of IROFS, general

descriptions of types of accidents sequences, and “information demonstrating compliance with

10 CFR 70.61.” To meet the latter requirement, an applicant would have to provide, as a

minimum, likelihood and consequence information for each type of process accident sequence

identified in the ISA Summary. To permit the reviewer(s) to evaluate the effectiveness of the

applicant’s likelihood and consequence evaluation methods, the reviewer(s) should also

examine the analyses of some accident sequences that are not reported in the ISA Summary

for which the applicant established consequences not to exceed the performance requirements

of 10 CFR 70.61.

In some simple cases, the information normally contained in the ISA Summary process

descriptions and list of IROFS might be sufficient to enable the reviewer(s) to understand how

compliance is achieved when taken together with the description of ISA likelihood evaluation

methods and criteria. However, in general, a description of how the applicant’s ISA team

evaluated credible accident likelihood to be “highly unlikely” or “unlikely” should be supplied.

The reviewer(s) should evaluate the efficacy of the applicant’s ISA methods. To do this, in

addition to reviewing the description of the ISA methods, the reviewer(s) will need to understand

how these methods have been applied in practice to the wide diversity of process safety

designs in the facility. Examples included in the ISA Summary of how the methods are applied

to a representative sample of processes should allow the reviewer(s) to understand the

applicant’s ISA method(s). In addition, a thorough understanding of the applicant’s ISA

method(s) will enable the reviewer(s) to better select other processes for which additional

“vertical slice” reviews may need to be performed onsite. The method for selecting specific

processes or accidents for additional onsite reviews is described in Section 3.5 of this chapter,

“Review Procedures.” 

For an average-sized fuel fabrication facility, the ISA Summary should include a detailed

demonstration of the application of the ISA methods to three or four nuclear criticality accident

sequences, one fire accident sequence, and one environmental/radiological/chemical accident

sequence. The number and selection of accident sequences for which a demonstration of the

ISA method(s) should be included in the ISA Summary will depend on the (1) size and number

of processes at the facility, (2) number of accident sequences for which the consequences

could exceed the 10 CFR 70.61 performance requirements, (3) diversity of process designs,

and (4) types and numbers of designated IROFS.

The NRC review of the applicant’s example accident sequence evaluations included in the ISA

Summary is not a substitute for the “vertical slice” and “horizontal” reviews that should be

performed using detailed information at the site. This onsite evaluation of ISA documentation

and processes must be NRC-selected in order to confirm that the ISA was actually performed

as described in the ISA Summary.
Integrated Safety Analysis 3-7 NUREG-1520

3.4 ACCEPTANCE CRITERIA

3.4.1 Regulatory Requirements

10 CFR 70.62 specifies the requirement to establish and maintain a safety program, including

performance of an ISA. 10 CFR 70.62(c) specifies requirements for conducting an ISA

including a demonstration that credible high-consequence and intermediate-consequence

events meet the safety performance requirements of 10 CFR 70.61. The requirement to

prepare and submit an ISA Summary for NRC approval is stated in 10 CFR 70.65(b). 

10 CFR 70.65(b) also describes the contents of an ISA Summary. 10 CFR 70.72 sets forth

requirements for maintaining the ISA, ISA documentation and the ISA Summary current when

changes are made to the site, structures, processes, systems, equipment, components,

computer programs, and activities of personnel. 

The information to be included in the ISA Summary can be divided into four categories including

(1) site and facility characteristics, (2) ISA method(s), (3) hazards and accident analysis and 

(4) IROFS. The following table summarizes the information requirements of each category, the

corresponding regulatory citation, and the section of NUREG-1520, Chapter 3, in which the

expectations for such information are described.
Integrated Safety Analysis 3-8 NUREG-1520

Information Requirements for the ISA Summary

Information Category and Requirement 10 CFR Part 70 Regulatory

Citation

NUREG-1520, Chapter 3

Section Reference

Site and Facility Characteristics:

$Site description 70.65(b)(1) 3.4.3.2(1)

$Facility description 70.65(b)(2) 3.4.3.2(2)

$Criticality monitoring and alarms 70.65(b)(4) 3.4.3.2(4C)

$Compliance with baseline design criteria 70.64 (if applicable) 3.4.3.2(4D) 

ISA Method(s):

$ISA method(s) description 70.65(b)(5) 3.4.3.2(5)

$ISA team description 70.65(b)(5) 3.4.3.2(5)

$Quantitative standards for acute chemical

 exposures

70.65(b)(7) 3.4.3.2(7)

$Definition of “unlikely,” “highly unlikely,” and “credible” 70.65(b)(9) 3.4.3.2(9)

Hazards and Accident Analysis:

$Description of processes analyzed 70.65(b)(3) 3.4.3.2(3)

$Identification of hazards 70.65(b)(3) 3.4.3.2(3)

$Description of accident sequences 70.65(b)(3) 3.4.3.2(3)

$Characterization of high- and intermediate-

 consequence accident sequences

70.65(b)(3) 3.4.3.2(3)

Items Relied on For Safety:

$List and description of items relied on for safety (IROFS) 70.65(b)(6) 3.4.3.2(6)

$Description of IROFS’ link to accident sequences to show 

 10 CFR 70.61 compliance 

70.65(b)(6) 3.4.3.2(4) and (6)

$IROFS management measures 70.65(b)(4) 3.4.3.2(4B) and (6)

$List of sole IROFS 70.65(b)(8) 3.4.3.2(8)

3.4.2 Regulatory Guidance

Guidance applicable to performing an ISA and documenting the results is contained in 

NUREG-1513, “Integrated Safety Analysis Guidance Document,” May 2001. NUREG/CR-6410,

ANuclear Fuel Cycle Accident Analysis Handbook,@ March 1998, provides guidance on

acceptable methods for evaluating the chemical and radiological consequences of potential

accidents.

3.4.3 Regulatory Acceptance Criteria

The acceptance criteria for an ISA are contingent on meeting the relevant requirements of

10 CFR Part 70, “Domestic Licensing of Special Nuclear Material.” The ISA will form the basis

for the safety program by identifying potential accidents, designating IROFS and management

measures, and evaluating the likelihood and consequences of each accident sequence for

compliance with the performance requirements of 10 CFR 70.61. Some of the acceptance

criteria address the programmatic commitments made by the applicant to perform and maintain

an ISA. The remainder of the criteria address the ISA results, as documented in the ISA

Summary, and whether those documented results demonstrate that the applicant=s IROFS and
Integrated Safety Analysis 3-9 NUREG-1520

management measures can reasonably be expected to ensure that the relevant accident

sequences will meet the performance requirements of 10 CFR 70.61.

3.4.3.1 Safety Program and ISA Commitments

This section discusses the acceptance criteria for license commitments pertaining to the

facility’s safety program including the performance of an ISA. 10 CFR Part 70 contains a

number of specific safety program requirements related to the ISA. Acceptance criteria for the

content of the ISA Summary appear in Section 3.4.3.2. These include the primary

requirements that an ISA be conducted, and that it evaluate and show that the applicant=s

facility complies with the performance requirements of 10 CFR 70.61. For each component of

the safety program, there may be several necessary elements, including organization,

assignment of responsibilities, management policies, required activities, written procedures for

activities, use of industry consensus standards, and technical safety practices, among others.

The applicant=s commitments for each of the three elements of the safety program defined in 

10 CFR 70.62(a) should be acceptable if the applicant does the following:

(1) Process Safety Information

a. The applicant commits to compile and maintain an up-to-date database of processsafety information. Written process-safety information will be used in updating the ISA

and in identifying and understanding the hazards associated with the processes. The

compilation of written process-safety information should include information pertaining

to:

i. The hazards of all materials used or produced in the process, which should

include information on chemical and physical properties (such as toxicity, acute

exposure limits, reactivity, and chemical and thermal stability) such as are

included on Material Safety Data Sheets (meeting the requirements of 

29 CFR 1910.1200(g)).

ii. Technology of the process should include a block flow diagram or simplified

process flow diagram, a brief outline of the process chemistry, safe upper and

lower limits for controlled parameters (e.g., temperature, pressure, flow, and

concentration), and evaluation of the health and safety consequences of process

deviations. 

iii. Equipment used in the process should include general information on topics

such as the materials of construction, piping and instrumentation diagrams

(P&IDs), ventilation; design codes and standards employed, material and energy

balances, IROFS (e.g., interlocks, detection, or suppression systems), electrical

classification, and relief system design and design basis.

b. The applicant includes procedures and criteria for changing the ISA, along with a

commitment to design and implement a facility change mechanism that meets the

requirements of 10 CFR 70.72. The applicant should discuss the evaluation of the

change within the ISA framework, as well as procedures and responsibilities for

updating the facility’s ISA.
Integrated Safety Analysis 3-10 NUREG-1520

c. The applicant commits to engage personnel with appropriate experience and expertise

in engineering and process operations to maintain the ISA. The ISA team for a process

should consist of individuals who are knowledgeable in the facility=s ISA method(s) and

the operation, hazards, and safety design criteria of the particular process. 

(2) ISA

a. The applicant commits to conduct an ISA of appropriate complexity for each process,

such that it identifies (i) radiological hazards, (ii) chemical hazards that could increase

radiological risk, (iii) facility hazards that could increase radiological risk, (iv) potential

accident sequences, (v) consequences and likelihood of each accident sequence and

(vi) IROFS including the assumptions and conditions under which they support

compliance with the performance requirements of 10 CFR 70.61. The application is

acceptable if it describes sufficiently specific methods and criteria that would be effective

in accomplishing each of these tasks. Such effective methods and criteria are

described in NUREG-1513, NUREG-6410, item 5 of SRP Section 3.4.3.2, and 

Appendix A to this chapter.

b. The applicant commits to maintain the ISA and its supporting documentation so that it is

accurate and up-to-date by means of a suitable configuration management system and

to submit changes to the ISA Summary, to the NRC, in accordance with 

10 CFR 70.72(d)(1) and (3). The ISA must account for any changes made to the facility

or its processes (e.g., changes to the site, operating procedures, or control systems). 

Management policies, organizational responsibilities, revision time frame, and

procedures to perform and approve revisions to the ISA should be outlined succinctly. 

The applicant commits to evaluate any facility changes or changes in the process safety

information that may alter the parameters of an accident sequence by means of the

facility=s ISA method(s). For any revisions to the ISA, the applicant commits to use

personnel having qualifications similar to those of ISA team members who conducted

the original ISA.

c. The applicant commits to train personnel in the facility's ISA method(s) and/or to use

suitably qualified personnel to update and maintain the ISA and ISA Summary.

d. The applicant commits to evaluate proposed changes to the facility or its operations by

means of the ISA method(s) and to designate new or additional IROFS and appropriate

management measures as required. The applicant also agrees to promptly evaluate the

adequacy of existing IROFS and associated management measures and to make any

required changes that may be impacted by changes to the facility and/or its processes. 

If a proposed change results in a new type of accident sequence (e.g., different initiating

event, significant changes in the consequences) or increases the consequences and/or

likelihood of a previously analyzed accident sequence within the context of 

10 CFR 70.61, the applicant commits to promptly evaluate the adequacy of existing

IROFS and associated management measures and to make necessary changes, if

required.

e. The applicant commits to address any IROFS’ unacceptable performance deficiencies

that are identified through updates to the ISA. 

f. The applicant commits to maintain written procedures on site.
Integrated Safety Analysis 3-11 NUREG-1520

g. The applicant commits to establish all IROFS (if not already established) and to maintain

them so that they are available and reliable when needed. 

In citing industry consensus standards, the applicant should delineate specific commitments in

the standards that will be adopted. The applicant should provide justifications if a standard is

not adopted in its entirety.

(3) Management Measures

a. The applicant commits to establish management measures (which are evaluated using

SRP Chapter 11) that comprise the principal mechanism by which the reliability and

availability of each IROFS is ensured.

3.4.3.2 ISA Summary and ISA Documentation

Information in the ISA Summary should provide the basis for the reviewer(s) to conclude that

there is reasonable assurance that the identified IROFS will satisfy the performance

requirements of 10 CFR 70.61. To do this, the reviewer must conclude that the applicant=s ISA

program has the capability to identify appropriate IROFS, and that IROFS identified in the ISA

Summary are adequate to control the potential accidents of concern at the facility. The

accidents of concern are those for which the consequences would be at the high and

intermediate consequence levels, absent any preventive or mitigative controls. In this context,

adequacy means the capability of the IROFS to prevent the related accidents with sufficient

reliability, or to sufficiently mitigate their consequences so that the performance requirements of

10 CFR 70.61 can be met. To support such a review, sufficient information about an accident

sequence and the proposed IROFS must be included in the ISA Summary to allow the

reviewer(s) to assess their contributions to prevention or mitigation. The ISA Summary must

contain enough information concerning the ISA methods and the qualifications of the ISA team

who performed the ISA and any other resources employed to give the reviewer(s) confidence

that the potential accidents identified are reasonably complete.

In addition, the reviewer(s) need to determine that appropriate management measures will be in

place to ensure the availability and reliability of the identified IROFS, when needed. Review of

designated management measures is addressed in Chapter 11 of this SRP.

The following acceptance criteria address each of the content elements of the ISA Summary

required by 10 CFR 70.65(b). For new facilities, the reviewer(s) should also evaluate those

aspects of the design that address baseline design criteria of 10 CFR 70.64 applicable to

individual processes. Thus, the following content elements have defined acceptance criteria:

(1) general description of the site

(2) general description of the facility

(3) description of facility processes, hazards, and types of accident sequences

(4) demonstration of compliance with 10 CFR 70.61 performance requirements

(5) description of the ISA team qualifications and ISA methods

(6) descriptive list of IROFS

(7) description of acute chemical exposure standards used

(8) descriptive list of sole IROFS

(9) definitions of “credible,” “unlikely,” and “highly unlikely”
Integrated Safety Analysis 3-12 NUREG-1520

Detailed acceptance criteria for each element of the ISA Summary follow:

(1) Site

The description in the ISA Summary of the site for processing nuclear material is considered

acceptable if the applicant includes, or references, the following safety-related information, with

emphasis on those factors that could affect safety:

a. A description of the site geography, including its location, taking into account prominent

natural and man-made features such as mountains, rivers, airports, population centers,

possibly hazardous commercial and manufacturing facilities, transportation routes, etc.,

adequate to permit evaluation of: i) the likelihoods of accidents caused by external

factors; and ii) the consequences of potential accidents.

b. Population information, based on most recent census data, that shows population

distribution as a function of distance from the facility, adequate to permit evaluation of

regulatory requirements, including exposure of the public to consequences listed in 

10 CFR 70.61.

c. Characterization of natural phenomena (e.g., tornadoes, hurricanes, floods, and

earthquakes) and other external events sufficient to assess their impact on facility safety

and to assess their likelihood of occurrence. At a minimum, the 100-year flood should

be postulated, consistent with U.S. Army Corps of Engineers flood plain maps. The

applicant also provides earthquake accelerations for the site associated with a 250-year

and 500-year earthquake. The discussion identifies all design basis natural events for

the facility, indicates which events are considered incredible, and describes the basis for

that determination. The assessment also indicates which events could occur without

adversely impacting safety.

(2) Facility

The description of the facility is considered acceptable if the applicant identifies and describes

the general features that affect the reliability or availability of IROFS. If such information is

available elsewhere in the application, reference to the appropriate sections is considered

acceptable. The information provided should adequately support an overall understanding of

the facility structure and its general arrangement. As a minimum, the applicant adequately

identifies and describes:

a. The facility location and the distance from the site boundary in all directions, including

the distance to the nearest resident and distance to boundaries in the prevailing wind

directions. 

b. Restricted area and controlled area boundaries.

c. Design information regarding the resistance of the facility to failures caused by credible

external events, when those failures may produce consequences exceeding those

identified in 10 CFR 70.61.

d. The location and arrangement of buildings on the facility site.
Integrated Safety Analysis 3-13 NUREG-1520

(3) Processes, Hazards, and Accident Sequences

Processes

The description of the processes analyzed as part of the ISA [10 CFR 70.62(c)(1) (i-vi)] is

considered acceptable if it describes the following features in sufficient detail to permit an

understanding of the theory of operation, and to assess compliance with the performance

requirements of 10 CFR 70.61. A description at a systems level is acceptable, provided that

it permits the NRC reviewer to adequately evaluate (1) the completeness of the hazard and

accident identification tasks and (2) the likelihood and consequences of the accidents

identified. If the information is available elsewhere in the application and is adequate to

support the ISA, reference to the appropriate sections is considered acceptable. The

information provides an adequate explanation of how the IROFS reliably prevent the

process from exceeding safety limits for each high and intermediate consequence accident

sequence.

a. Basic process function and theory includes a general discussion of the basic theory of

the process. 

b. Major components includes the general arrangement, function, and operation of major 

components in the process. If appropriate, it also includes arrangement drawings and 

process schematics showing the major components and instrumentation, and chemical 

flow sheets showing compositions of the various process streams.

c. Process design and equipment include a discussion of process design, equipment, and 

instrumentation that is sufficiently detailed to permit an adequate understanding of the

results of the ISA. As appropriate, it includes schematics indicating safety

interrelationships of parts of the process. In particular, it is usually necessary for

criticality safety to diagram the location and geometry of the fissile and other materials in

the process, for both normal and bounding abnormal conditions. This can be done

using either schematic drawings or textual descriptions indicating the location and

geometry of fissile materials, moderators, etc., sufficient to permit an understanding of

how the IROFS limit the mass, geometry, moderation, reflection, etc. If such details are

not included in the ISA Summary, the information may be verified as part of an onsite

ISA review.

d. Process operating ranges and limits include the operating ranges and limits for

measured process variables (e.g., temperatures, pressures, flows, and compositions)

that are controlled by IROFS to ensure safe operations of the process. If such details

are not included in the ISA Summary, the information may be verified as part of an

onsite ISA review.

Hazards

The description of process hazards provided in the ISA Summary is acceptable if it

identifies, for each process, all types of hazards that are relevant to determine compliance

with the performance criteria of 10 CFR 70.61. That is, the acceptance criterion is

completeness. All hazards that could result in an accident sequence in which the

consequences could exceed the performance requirements of 10 CFR 70.61 should be

listed, even if later analysis of a particular hazard shows that resulting accident sequences

do not exceed these limits. Otherwise the reviewer(s) cannot determine completeness. 
Integrated Safety Analysis 3-14 NUREG-1520

General exclusion from consideration of certain hazards for an entire facility can be justified

by bounding case analyses showing that, for the conditions or credible inventories on site,

the performance requirements of 10 CFR 70.61 cannot be exceeded. In this case, the

bounding inventories or conditions, if under the control of the applicant, become IROFS. 

The list of process hazards is acceptable if the ISA Summary provides the following

information:

a. a list of materials (radioactive, fissile, flammable, and toxic) or conditions that could

result in hazardous situations (e.g., loss of containment of licensed nuclear material),

including the maximum intended inventory amounts and location(s) of the hazardous

materials at the facility 

b. potential interactions among materials or conditions that could result in hazardous

situations

Accident Sequences

The general description of types of accident sequences in the ISA Summary is acceptable if

the reviewer can determine the following considerations:

a. The applicant has identified all accidents for which the consequences could exceed the

performance requirements of 10 CFR 70.61.

b. The applicant has identified how the IROFS listed in the ISA Summary protect against

each such type of accident. 

General types of accident sequences differ if they consist of a different set of IROFS

failures. Thus, several processes, each using a set of IROFS that is functionally of the

same type (e.g., same mechanical, physical, and/or electrical principle of operation), can be

summarized as a single type of accident sequence and listed only once. However, the

individual processes covered by this system should be individually identified in a way that

the reviewer(s) can determine completeness in addressing all processes.

For this reason, it is not generally acceptable to merely list the type of hazard or the

controlled parameters without referencing the items relied on to control that parameter or

hazard. The description of general types of accident sequences is acceptable if it covers all

types of sequences, initiating events and IROFS failures. Initiating events may be either a

failure of an IROFS or an external event. Human errors can be initiating events or IROFS

failures. The description of a general type of accident sequence is acceptable if it permits

the reviewer to determine how each accident sequence for which the consequences could

exceed the performance requirements of 10 CFR 70.61 is protected against by IROFS or a

system of IROFS. 

One acceptable way to do this is to show a fault tree on which the basic events are IROFS

failures. Another acceptable way is to provide a table on which each row displays the

events in an accident sequence, such as in Appendix A, Table A-7, where, in general, each

event is a failure of an IROFS. Another acceptable way is to provide a narrative summary

for each process describing the sequence of events in each type of accident.

To demonstrate completeness, the description of general types of accident sequences must

be identified using systematic methods and consistent references. Therefore, each
Integrated Safety Analysis 3-15 NUREG-1520

description of a general type of accident sequence is acceptable if it meets the following

criteria: 

a. An acceptable method of hazard identification and process hazard analysis was used in

accordance with the criteria of NUREG-1513.

b. The selected method was correctly applied. 

c. The applicant did not overlook any accident sequence for which the consequences

could exceed the performance requirements of 10 CFR 70.61.

d. The applicant used a method of identifying facility processes that ensured identification

of all processes.

During the early phases of an ISA, accidents will be identified for which the consequences

may initially be unknown. These accidents will later be analyzed and may be shown to have

consequences that are less than the levels identified in 10 CFR 70.61.

The ISA Summary need not list as a separate type of accident sequence, every conceivable

permutation of an accident. Accidents having characteristics that all fall in the same

categories can be grouped as a single type of accident in the ISA Summary, provided that

the following conditions are fulfilled: 

a. The initiating events have the same effect on the system.

b. They all consist of failures of the same IROFS or system of IROFS. 

c. They all result in violation of the safety limit on the same parameter.

d. They all result in the same type and severity categories of consequences. 

(4) Information Demonstrating Compliance with the Performance Requirements of 

10 CFR 70.61, Including (a) Accident Sequence Evaluation and IROFS Designation, 

(b) Management Measures, (c) Requirements of Criticality Monitoring, and 

(d) Requirements for New Facilities or New Processes at Existing Facilities

a. Accident Sequence Evaluation and IROFS Designation

10 CFR 70.65(b)(4) requires that the ISA Summary contain Ainformation that demonstrates

compliance with the performance criteria of 10 CFR 70.61.@ Since the requirements of 

10 CFR 70.61 are expressed in terms of consequences and likelihoods of events, the ISA

Summary should provide sufficient information to demonstrate the following considerations:

i. Credible high-consequence events are highly unlikely.

ii. Credible intermediate-consequence events are unlikely.

The performance requirements of 10 CFR 70.61 have three elements, including 

(a) completeness, (b) consequences, and (c) likelihood. 
Integrated Safety Analysis 3-16 NUREG-1520

Completeness refers to the fact that the ISA must address each credible event. 

Consequences refer to the magnitude of the chemical and radiological doses of the

accident and is the basis upon which an accident in classified in 10 CFR 70.61 to be a high

or intermediate consequence event. Likelihood refers to the fact that 10 CFR 70.61

requires that intermediate consequence events be unlikely, and high consequence events

be highly unlikely. Thus, the information provided must address each of these three

elements. 

To be acceptable, the information provided must correspond to the ISA methods,

consequence, and likelihood definitions described in the submittal. The information must

also show the basis for and results of applying these methods to each process. In addition,

the information must show that the methods have been properly applied in each case.

The information showing completeness, consequences, and likelihood for accident

sequences can be presented in various formats, including logic diagrams, fault trees, or

tabular summaries. Appendix A of this chapter provides one example of how this

information could be presented in an application. 

Each of these performance requirements (completeness, consequences, and likelihood) is

discussed below.

Completeness is demonstrated by correctly applying an appropriate accident identification

method, as described in NUREG-1513, AIntegrated Safety Analysis Guidance Document.@

Completeness can be effectively displayed by using an appropriate diagram or description

of the identified accidents. Specific acceptance criteria for completeness are covered in

item 3 above.

Consequence information in the ISA Summary is acceptable for showing compliance with 

10 CFR 70.61, provided that the following conditions are met.

! The information in the ISA Summary for each accident for which the consequences

could exceed the performance requirements of 10 CFR 70.61 includes an estimate of

its quantitative consequences (doses, chemical exposures, criticality) in a form that can

be directly compared with the consequence levels in 10 CFR 70.61, or includes a

reference to a value documented elsewhere in the ISA Summary that applies to or

bounds that accident.

! The consequences were calculated using a method and data consistent with 

NUREG-6410, ANuclear Fuel Cycle Facility Accident Analysis Handbook,@ March 1998,

or another method described and justified in the methods description section of the ISA

Summary.

! All consequences that could result from the accident sequence have been evaluated. 

That is, if an accident can result in a range of consequences, all possibilities must be

considered, including the maximum source term and most adverse weather that could

occur. However, if such conditions are unlikely to occur, credit can be taken for this in

the evaluation of likelihood.

! The ISA Summary correctly assigns each type of accident to one of the consequence

categories of 10 CFR 70.61 (namely, high, or intermediate).
Integrated Safety Analysis 3-17 NUREG-1520

Unshielded nuclear criticality accidents are considered to be high consequence events,

because the radiation exposure that an individual could receive exceeds the acute 1 Sv

(100 rem) dose established by 10 CFR 70.61(b)(1). For processes with effective

engineered shielding, criticalities may actually produce doses below the intermediate

consequences of 10 CFR 70.61. As stated in the regulation, primary reliance must be

on preventing inadvertent nuclear criticalities. This applies, notwithstanding shielding

or other mitigative features. Therefore, regardless of the actual consequences,

shielded criticalities must meet the likelihood criteria described in the following section

of this SRP. If needed, the ANuclear Fuel Cycle Facility Accident Analysis Handbook,@

NUREG/CR-6410, provides methods for estimating the magnitudes of criticality events

that can be applied for workers or members of the public at varying distances from the

event. 

Likelihood information in the ISA Summary is acceptable to show compliance with 

10 CFR 70.61, provided the following conditions are met:

! The ISA Summary specifies the likelihood of each general type of accident sequence

that could exceed the performance requirements of 10 CFR 70.61.

! The likelihoods are derived using an acceptable method described in the ISA

Summary=s methods section.

! The likelihoods comply with acceptable definitions of the terms Aunlikely@ and Ahighly

unlikely,@ as described in this SRP chapter. Note that, when interpreted as required

accident frequencies, these terms refer to long-run average frequencies, not

instantaneous values. That is, a system complies with the performance requirements

of 10 CFR 70.61 as a long-run average. Otherwise, failure of any IROFS, even for a

very short period, would violate the requirement, which is not the intent.

b. Management Measures

10 CFR 70.65(b)(4) requires a description of the management measures to be applied to

IROFS for each accident sequence for which the consequences could exceed the

performance requirements of 10 CFR 70.61. Chapter 11 of this SRP provides detailed

criteria against which the adequacy of such management measures can be evaluated.

c. Criticality Monitoring

10 CFR 70.24 defines specific sensitivity requirements for criticality monitors. To

demonstrate compliance, the application should describe the method for evaluating an

acceptable response of at least two detectors to a nuclear criticality at any location where

SNM may be handled, used, or stored. Locations of all detectors relative to the potential

locations of SNM should be provided as a diagram. The application should also provide

information supporting determination of the gamma and neutron emission characteristics of

the minimum credible accident of concern capable of producing the effects specified in 

10 CFR 70.24. In addition, the application should provide information showing the response

characteristics of the detectors to neutron and gamma doses and rates characteristic of

credible accidents.
Integrated Safety Analysis 3-18 NUREG-1520

10 CFR 70.24 also requires specific emergency preparations. Specifically, the application

should provide information to demonstrate that the applicant’s equipment and procedures

are adequate to ensure that these requirements are met.

d. New Facilities or New Processes at Existing Facilities

10 CFR 70.64 specifies baseline design criteria that must be used, as applicable, for new

facilities and new processes at existing facilities. If the application involves such new

facilities or processes, the ISA Summary should explain how each baseline design criterion

was addressed in the design of the facility. For deterministic design criteria such as doublecontingency, the process-specific information may be provided, along with the other process

information in the ISA Summary. Design basis events and safety parameter limits should

also be given. In addition, the application should provide methods, data, and results of

analysis showing compliance with these design bases for individual processes and facilities.

10 CFR 70.64 states that the design process must be founded on defense-in-depth

principles, and must incorporate, to the extent practicable, preference for engineered

controls over administrative controls, and reduction of challenges to IROFS. Because of

this regulation, new facilities with system safety designs lacking defense-in-depth,

consisting of purely administrative controls, or relying on IROFS that are frequently or

continuously challenged, are not acceptable, unless the application provides justification

showing that alternatives to achieve the design criteria are not feasible.

(5) ISA Team Qualifications and ISA Methods

The ISA teams [10 CFR 70.62(c)(2)] and their qualifications as stated in the ISA Summary are

acceptable if the following criteria are met:

a. The ISA team has a leader who is formally trained and knowledgeable in the ISA

method(s) chosen for the hazard and accident evaluations. In addition, the team leader

should have an adequate understanding of all process operations and hazards under

evaluation, but should not be the responsible, cognizant engineer or expert for that

process.

b. At least one member of the ISA team has thorough, specific, and detailed experience in

the process under evaluation.

c. The team represents a variety of process design and safety experience in those

particular safety disciplines relevant to hazards that could credibly be present in the

process, including, if applicable, radiation safety, nuclear criticality safety, fire protection,

and chemical safety disciplines. 

d. A manager provides overall administrative and technical direction for the ISA.

The description of the ISA method(s) is acceptable if the following criteria are met:

a. Hazard Identification Method. The hazard identification method selected is considered

acceptable if it fulfils the following criteria: 
1

 At a minimum, the inventory list should include the following hazardous materials if present onsite: 

ammonia; fines (uranium oxide dust, beryllium); flammable liquids and gases; fluorine; hydrofluoric

acid; hydrogen; nitric acid; organic solvents; propane; uranium hexafluoride; and Zircalloy.

2

 The release of hazardous chemicals is of regulatory concern to the NRC only to the extent that

such hazardous releases result from the processing of licensed nuclear material or have the potential

to adversely affect radiological safety.

Integrated Safety Analysis 3-19 NUREG-1520

i. Provide a list of materials (radioactive, fissile, flammable, and toxic) and conditions

that could result in hazardous situations (e.g., loss of containment of licensed

nuclear material). The list should include maximum intended inventory amounts and

the location of the hazardous materials at the facility. 1

ii. Determine potential interactions between materials or conditions that could result in

hazardous situations.

b. Process Hazard Analysis Method. The process hazard analysis method is acceptable if

it involves selecting one of the methods described in NUREG-1513 in accordance with

the selection criteria established in that document. Methods not described in 

NUREG-1513 may be acceptable provided that they fulfil the following conditions:

i. Criteria are provided for their use for an individual process, and are consistent with

the principles of the selection criteria in NUREG-1513. 

ii. It adequately addresses all the hazards identified in the hazard identification task. If

an identified hazard is eliminated from further consideration, such action is justified.

iii. The method provides reasonable assurance that the applicant can identify all

significant accident sequences (including the IROFS used to prevent or mitigate the

accidents) that could exceed the performance requirements of 10 CFR 70.61. 2

iv. The method takes into account the interactions of identified hazards and proposed

IROFS, including system interactions that could result in an accident sequence for

which the consequences could exceed the performance requirements of 

10 CFR 70.61. 

v. The method addresses all modes of operation, including startup, normal operation,

shutdown, and maintenance.

vi. The method addresses hazards resulting from process deviations (e.g., high

temperature, and high pressure), initiating events internal to the facility (e.g., fires or

explosions), and hazardous credible external events (e.g., floods, high winds,

earthquakes, and airplane crashes). The applicant provides justification for

determinations that certain events are not credible and, therefore, not subject to the

likelihood requirements of 10 CFR 70.61.

vii. It adequately considers initiation of or contribution to accident sequences by human

error through the use of human-systems interface analysis or other appropriate

methods.
Integrated Safety Analysis 3-20 NUREG-1520

viii. It adequately considers common mode failures and system interactions in evaluating

systems that are to be protected by double-contingency.

ix. The ISA Summary provides justification that the individual method would effectively

accomplish conditions ii through viii, above.

c. Consequence Analysis Method. The methods used for ISA consequence evaluation, as

described in the ISA Summary, are acceptable, provided the following conditions are

met:

i. The methods are consistent with the approaches described in NUREG/CR-6410,

ANuclear Fuel Cycle Facility Accident Analysis Handbook,” March 1998.

ii. The use of generic assumptions and data is reasonably conservative for the types of

accidents analyzed.

d. Likelihood Evaluation Method. The method for evaluating the likelihood of accident

sequences, as described in the ISA Summary, is considered acceptable, provided the

following conditions are met:

i. The method clearly shows how each designated IROFS acts to prevent or mitigate

the consequences (to an acceptable level) of the accident sequence being

evaluated.

ii. When multiple IROFS are designated for an accident sequence, the method

considers the interaction of all such IROFS, as in a logic diagram or tabulation that

accounts for the impact of redundancy, independence, and surveillance on the

likelihood of occurrence of the accident.

iii. The method has objective criteria for evaluating, at least qualitatively, the likelihood

of failure of individual IROFS. When applicable, such likelihood criteria should

include the means to limit potential failure modes, the magnitude of safety margins,

the type of engineered equipment (active or passive) or human action that

constitutes the IROFS, and the types and safety grading (if any) of the management

measures applied to the IROFS. 

 iv. Finally, the method evaluates the likelihood of each accident sequence as unlikely,

highly unlikely, or neither, as defined by the applicant, in accordance with Section

3.4.3.2, Item 9, of this chapter. 

v. For nuclear criticality accident sequences, the method evaluates compliance with 

10 CFR 70.61(d). That is, even in a facility with engineered features to limit the

consequences of nuclear criticalities, preventive control(s) must be in place that are

sufficient to ensure that the likelihood of criticality is controlled to be Ahighly unlikely.@

A moderately higher standard of likelihood may be permitted in preventing such

events, consistent with ANSI/ANS Standard 8.10. In particular, criticality cannot

result from the failure of any single IROFS. In addition, potential criticality accidents

must meet an approved margin of subcriticality for safety. Acceptance criteria for

such margins are reviewed as programmatic commitments, but the ISA methods

must consider and the ISA Summary must document, the actual magnitude of those
Integrated Safety Analysis 3-21 NUREG-1520

margins when they are part of the reason why the postulated accident sequence

resulting in criticality is highly unlikely.

One acceptable method of likelihood evaluation is described in Appendix A of this chapter.

(6) Descriptive List of all IROFS

The Alist describing items relied on for safety@ required by 10 CFR 70.62(c)(vi) is acceptable,

provided the following conditions are met:

a. The list includes all IROFS in the identified high and intermediate consequence accident

sequences

b. The description of the IROFS includes management measures applied to the IROFS

(including the safety grading), characteristics of its preventive, mitigative, or other safety

function, and assumptions and conditions under which the item is relied on to support

compliance with the performance requirements of 10 CFR 70.61. If information on any

safety limits and safety margins associated with an IROFS is not provided in the ISA

Summary, it must be available for review in ISA documentation onsite.

The above acceptance criteria are explained in greater detail below.

a. All Items: The primary function of the list describing each IROFS is to document the safety

basis of all processes in the facility. This list assists in ensuring that the items are not

degraded without a justifying safety review. Thus, the key feature of this list is that all

IROFS are included. To be acceptable, no item, aspect, feature, or property of a process

that is needed to show compliance with the safety performance requirements of the

regulation may be left off this list. IROFS may be hardware with a dedicated safety function

or hardware with a property that is relied on for safety. Thus, IROFS may be the dimension,

shape, capacity, or composition of hardware. The ISA Summary need not provide a

breakdown of hardware IROFS by component or identify all support systems. However, the

ISA documentation maintained onsite, such as system schematics and/or descriptive lists,

should contain sufficient detail about items within a hardware IROFS, such that it is clear to

the reviewer(s) and the applicant, what structure, system, equipment, or component is

included within the hardware IROFS’ boundary and would, therefore, be subject to

management measures specified by the applicant. Some examples of items within a

hardware IROFS are detectors, sensors, electronics, cables, valves, piping, tanks, dykes,

etc. In addition, ISA documentation should also identify essential utilities and support

systems on which the IROFS depends to perform its intended function. Some examples of

these are backup batteries, air supply, steam supply, etc. In some processes, the

frequency of demands made on IROFS must be controlled or limited to comply with 

10 CFR 70.61. In such processes, whatever features are needed to limit the frequency of

demands are themselves IROFS.

b. Description of Items: The essential features of each IROFS should be described. Sufficient

information should be provided about engineered hardware controls to permit an evaluation

that, in principle, controls of this type will have adequate reliability. Because the likelihood

of failure of items often depends on safety margins, the safety parameter controlled by the

item, the safety limit on the parameter, and the margin to true failure should, in general, be

described. For IROFS that are administrative controls, the nature of the action or
Integrated Safety Analysis 3-22 NUREG-1520

prohibition involved must be described sufficiently to permit an understanding that, in

principle, adherence to it should be reliable. Features of the IROFS that affect its

independence from other IROFS, such as reliance on the same power supplies, should be

indicated.

The description of each IROFS should identify its expected function, conditions needed for

the IROFS to reliably perform its function, and the effects of its failure. The description of

each IROFS within an ISA Summary should identify what management measures, such as

maintenance, training, configuration management, etc., are applied to it. If a system of

graded management measures is used, the grade applied to each control should be

determinable from information provided in the ISA Summary. The reliability required for an

IROFS is proportionate to the amount of risk reduction relied on. Thus, the quality of the

management measures applied to an IROFS may be graded commensurate with the

required reliability. The management measures should ensure that IROFS are designed,

implemented, and maintained, as necessary, to be available and reliable to perform their

function when needed. The degree of reliability and availability of IROFS ensured by these

measures should be consistent with the evaluations of accident likelihoods. In particular, for

redundant IROFS, all information necessary to establish the average vulnerable outage time

is required in order to maintain acceptable availability. Otherwise, failures must be

assumed to persist for the life of the facility. In particular, the time interval between

surveillance observations or tests of the item should be stated, since restoration of a safe

state cannot occur until the failure is discovered. 

One example of a tabular description of IROFS meeting these criteria is Table A-13 in

Appendix A to this chapter.

(7) Quantitative Standards for Chemical Consequences

The applicant=s description in the ISA Summary of proposed quantitative standards used to

assess consequences from acute chemical exposure to licensed material or chemicals incident

to the processing licensed material is acceptable, provided the following criteria are met:

a. There are unambiguous quantitative standards for each of the applicable hazardous

chemicals that meet the criteria of 10 CFR 70.65(b)(7) on site, corresponding to, and

consistent with, the quantitative standards in 10 CFR 70.61(b)(4)(i), 70.61(b)(4)(ii),

70.61(c)(4)(i), and 70.61(c)(4)(ii).

b. The quantitative standard of 10 CFR 70.61(b)(4)(i) addresses exposures that could

endanger the life of a worker. The applicant is appropriately conservative in applying the

language Acould endanger,@ so as to include exposures that would result in death,

consistent with the methods used for the U.S. Environmental Protection Agency’s--

“Acute Exposure Guidelines--40 CFR Part 68.”

c. The quantitative standards for 10 CFR 70.61(b)(4)(ii) and 10 CFR 70.61(c)(4)(i) will

correctly categorize all exposures that could lead to irreversible or other serious, longlasting health effects to individuals. As with criterion (b), above, the standard selected

should have appropriate conservatism.

d. The quantitative standard for 10 CFR 70.61(c)(4)(ii) will correctly categorize all exposures

that could cause mild transient health effects to an individual.
Integrated Safety Analysis 3-23 NUREG-1520

The NRC finds the use of the Emergency Response Planning Guidelines (ERPG) established

by the American Industrial Hygiene Association, the Acute Exposure Guideline Levels (AEGL)

established by the National Advisory Committee for Acute Guideline Levels for Hazardous

Substances and exposure limits established by the Occupational Safety and Health

Administration (OSHA) or contained in International Organization for Standardization (ISO)

standards to be acceptable. If the applicant does not use a published exposure standard, or if

a chemical has an unknown exposure standard, the ISA Summary must describe how an

alternative exposure standard was established for use in the ISA. The ISA Summary must list

the actual exposure values for each chemical, specify the source of the data (e.g., ERPG,

AEGL, ISO, etc.), and provide information or a reference justifying that they meet the

acceptance criteria stated above.

(8) List of Sole IROFS

The descriptive list in the ISA Summary that identifies all IROFS that are the sole item for

preventing or mitigating an accident sequence is acceptable if it includes:

a. a descriptive title of the IROFS

b. an unambiguous and clear reference to the process to which the item applies

c. clear and traceable reference to the description of the item as it appears in the full list of

all IROFS

(9) Definitions of “Unlikely”, “Highly Unlikely” and “Credible”

10 CFR 70.65 requires that the applicant=s ISA Summary must define the terms “unlikely,”

“highly unlikely,” and “credible.” The applicant=s definitions of these terms are acceptable if,

when used with the applicant=s method of assessing likelihoods, they provide reasonable

assurance that the performance requirements of 10 CFR 70.61 can be met. The applicant=s

method of likelihood evaluation and the definitions of the likelihood terms are closely related. 

Qualitative methods require qualitative definitions. Such a qualitative definition would identify

the qualities of IROFS, controlling an accident sequence, that would qualify that sequence as

Aunlikely@ or Ahighly unlikely.@

An applicant may use quantitative methods and definitions for evaluating compliance with 

10 CFR 70.61, but nothing in this SRP should be construed as an interpretation that such

methods are required. The reviewer(s) should focus on objective qualities and information

provided concerning accident likelihoods. 

10 CFR 70.61 requires that credible high consequence events be “highly unlikely.@ Thus, the

meaning of the phrase Ahighly unlikely@ is on a per-event basis. The same is true for the terms

Aunlikely@ and Acredible.@ Hence, applicant definitions should be on a per-event basis. The

events referred to are occurrences of consequences, which are herein synonymous with the

phrase Aaccident sequence.@ This is important to recognize, since there may be hundreds of

potential accident sequences identified in an ISA. Thus, the likelihood of each individual

sequence must be quite low.
Integrated Safety Analysis 3-24 NUREG-1520

Acceptance Criteria for the Definition of ACredible@

10 CFR 70.65 requires that the applicant define the term Acredible.@ This term is used in 

10 CFR 70.61, which requires that all credible accident sequences for which the consequences

could exceed the performance requirements of 10 CFR 70.61 must be controlled to be unlikely

or highly unlikely, as appropriate. If an event is not credible, IROFS are not required to prevent

or mitigate the event. Thus, to be Anot credible@ could be used as a criterion for exemption from

use of IROFS. There is a danger of circular reasoning here. In the safety program embodied

in Subpart H to 10 CFR Part 70, the fact that an event is not Acredible@ must not depend on any

facility feature that could credibly fail to function, or be rendered ineffective as a result of a

change to the system. Each facility feature that is needed to ensure that accident events are

sufficiently unlikely is an IROFS. There must be high assurance, provided by management

measures, that such features are not removed or rendered ineffective during system changes. 

One cannot claim that a process does not need IROFS because it is Anot credible@ due to

characteristics provided by IROFS. 

Any one of the following three independent acceptable sets of qualities could define an event as

not credible:

a. an external event for which the frequency of occurrence can conservatively be estimated

as less than once in a million years

b. a process deviation that consists of a sequence of many unlikely human actions or

errors for which there is no reason or motive (In determining that there is no reason for

such actions, a wide range of possible motives, short of intent to cause harm, must be

considered. Necessarily, no such sequence of events can ever have actually happened

in any fuel cycle facility.) 

c. process deviations for which there is a convincing argument, given physical laws, that

they are not possible, or are unquestionably extremely unlikely (The validity of the

argument must not depend on any feature of the design or materials controlled by the

facility’s system of IROFS or management measures.) 

The implication of Acredible@ in 10 CFR 70.61 is that events that are not Acredible@ may be

neglected. For this to be acceptable on a risk basis, unless the event is impossible, it must be

of negligible likelihood. Negligible likelihood means sufficiently low that, considering the

consequences, the addition to total risk is small. Note that consideration must thus be given to

how many such events have, in fact, been neglected. An applicant may demonstrate, by

quantitative reasoning, that a particular event is of negligible frequency. Such a demonstration

must be convincing despite the absence of designated IROFS. Typically, this can only be

achieved for external events known to be extremely unlikely. 

Acceptance Criteria for Qualitative Definitions of Likelihood

If the applicant=s definitions are qualitative, they are acceptable if they meet the following

criteria:

a. are reasonably clear and based on objective criteria
Integrated Safety Analysis 3-25 NUREG-1520

b. can reasonably be expected to consistently distinguish accidents that are highly unlikely

from those that are merely unlikely 

The phrase Aobjective criteria@ means the extent to which the method relies on specific

identifiable characteristics of a process design, rather than subjective judgments of adequacy. 

Objective criteria are needed to achieve consistency. Consistency means the degree to which

the same results are obtained when the method is applied by different analysts. This is

important to maintain an adequate standard of safety because ISAs of future facility

modifications may be performed by individuals not involved in conducting the initial ISA. 

Reliability and Availability Qualities

Qualitative methods of evaluating the likelihood of an accident sequence involve identifying the

reliability and availability qualities of each of the events that constitute the sequence. The

following lists of qualities are not necessarily complete, but contain many of the factors that are

most commonly encountered. Some of these qualities relate to the characteristics of individual

IROFS, such as the following examples:

a. safety margin in the controlled parameter, compared with process variation and

uncertainty

b. whether the IROFS is an active engineered control, a passive engineered control,

an administrative control, or an enhanced administrative control

c. the type and safety grading, if any, of management measures applied to the control 

d. fail-safe, self-announcing, or surveillance measures to limit down time

e. failure modes

f. demand rate

g. failure rate

Other reliability qualities relate characteristics of the IROFS or system of IROFS, protecting

against the following accident sequences as a whole among others:

h. defense-in-depth

i. degree of redundancy

j. degree of independence

k. diversity

l. vulnerability to common-cause failure

Methods of likelihood evaluation and definitions of the likelihood terms “unlikely” and “highly

unlikely” may mix qualitative and quantitative information. Certain types of objective

quantitative information may be available concerning specific processes in a facility. Some

examples of such objective quantitative information include the following:

a. reports of failure modes of equipment or violations of procedures recorded in

maintenance records or corrective action programs

b. the time intervals at which surveillance is conducted to detect failed conditions 

c. the time intervals at which functional tests or configuration audits are held

d. for a fail-safe, monitored, or self-announcing IROFS, the time it takes to render the

system safe

e. demand rates (i.e., how frequent are the demands on an IROFS to perform) (Some

situations amount to effectively continuous demand)
Integrated Safety Analysis 3-26 NUREG-1520

Such items of quantitative information should be considered in evaluating the likelihood of

accident sequences, even in purely qualitative evaluations. For example, knowing the value to

which down time is limited by surveillance can indicate that a system=s availability is extremely

high. For redundant systems, such high availability can virtually preclude concurrent

independent failures of multiple IROFS.

Acceptance Criteria for Likelihood Indexing Methods

One acceptable definition for the likelihood terms Aunlikely@ and Ahighly unlikely@ could be based

on a risk-indexing method. Such a method is described in the example in Appendix A of this

chapter which primarily relies on a qualitative evaluation of reliability and availability factors. In

such methods, qualitative characteristics of the system of IROFS, such as those listed above,

are used to estimate a quantitative likelihood index for each accident sequence. The definition

of Aunlikely@ then is an acceptable limit on this likelihood index.

Acceptance Criteria for Purely Qualitative Methods

A purely qualitative method of defining Aunlikely@ and Ahighly unlikely@ is acceptable if it

incorporates all of the applicable reliability and availability qualities to an appropriate degree. 

For example, one statement of applicable qualities is double-contingency protection, the quality

of a process design that incorporates sufficient factors of safety to require at least two unlikely,

independent, and concurrent changes in process conditions before a criticality accident is

possible. 

Double-contingency explicitly addresses several reliability and availability qualities:

Factors of safety: Safety margins

At least two: Redundancy

Unlikely: Low failure rate, low down time of one of two controls

Concurrent: Low down time

Independent: Independence

Process conditions: Physical events, not virtual human errors

One acceptable definition of “highly unlikely” is a system of IROFS that possesses doublecontingency protection, where each of the applicable qualities is present to an appropriate

degree. For example, as implied by the modifier Aat least,@ sometimes more than just two-fold

redundancy may be appropriate. 

A qualitative method may also be proposed for defining Aunlikely.@ Such a qualitative method

might simply list various combinations of reliability qualities for a system of IROFS that would

qualify as Aunlikely.@ For example, a single high-reliability IROFS, such as an engineered

hardware control with a high grade of applicable management measures, might qualify to be

considered Aunlikely to fail.@ Systems relying on administrative controls would normally have to

make use of enhancing qualities such as large safety margins and redundancy, to qualify as

Aunlikely to fail.@ A single simple administrative control, regularly challenged, without any

special safety margin or enhancement, where a single simple error would lead to an accident,

would not qualify as Aunlikely to fail.@
Integrated Safety Analysis 3-27 NUREG-1520

Acceptance Criteria for Quantitative Definitions of Likelihood

An applicant may choose to provide quantitative definitions of the terms “unlikely” and “highly

unlikely.” Quantitative guidelines are developed below. These guidelines serve two purposes. 

Specifically, (1) they can be used as acceptance criteria for quantitative definitions, if provided,

and (2) they provide guidance to the reviewer(s) when objective quantitative reliability and

availability information exists. 

The goals from which these quantitative guidelines were derived are for specific types of

accidents. Therefore, the guidelines should not be used for accidents that differ significantly

from these specific types. The high consequence guideline, for example, is based on a goal of

no inadvertent nuclear criticalities. Thus, this guideline should be used for accidents that have

consequences similar to a nuclear criticality accident (i.e., one where a few fatal or near fatal

worker doses may occur). For substantially more severe high consequence accidents, more

stringent likelihood criteria would be acceptable. For less severe high consequence accidents,

less stringent criteria may be applied. Quantitative guidelines are derived from goals, not limits,

and have been judged to be the highest values consistent with those goals. 

Quantitative Guidelines

Quantitative definitions of likelihood are based on the NRC’s strategic risk performance goals. 

Quantitative likelihood values are an appropriate fraction of the risks of other industrial accident

risks in the United States, and conform to comparable quantitative values that are already used

in other countries for regulation of nuclear materials facilities. A discussion of quantitative

guidelines here does not imply that quantitative demonstration of compliance with 

10 CFR 70.61 is required.

Highly Unlikely

The guideline for acceptance of the definition of “highly unlikely@ has been derived as the

highest acceptable frequency that is consistent with a goal of having no inadvertent nuclear

criticality accidents, and no accidents of similar consequences, in the industry. To within an

order of magnitude, this is taken to mean a frequency limit of less than one such accident in the

industry every 100 years. This has been translated below into a guideline limiting the frequency

of individual accidents to 10-5 per-event per-year. As the goal is to have no such accidents,

accident frequencies should be reduced substantially below this guideline when feasible.

Unlikely

Intermediate consequence events include significant radiation exposures to workers (those

exceeding 0.25 Sieverts or 25 rem). The NRC's goal is for there to be no increase in the rate of

such significant exposures. This has been translated below into a guideline of 

4 X 10-5 per-event per-year. This guideline may be more generally considered as a range

between 10-4 and 10-5 per-event per-year, since exact frequencies at such levels cannot

accurately be determined.
Integrated Safety Analysis 3-28 NUREG-1520

Quantitative Guidelines for use with Acceptance Criteria

The applicant=s quantitative definitions of the terms “unlikely” and “highly unlikely,” as applied to

individual accident sequences identified in the ISA, are acceptable to show compliance with 

10 CFR 70.61 if they are reasonably consistent with the following quantitative guidelines:

Likelihood term of 10 CFR 70.61 Guideline 

Unlikely Less than 10-4 per-event per-year

Highly Unlikely Less than 10-5 per-event per-year

The stated quantitative guidelines are used to define the largest likelihood values that would be

acceptable limits. Definitions based on lower limits are also acceptable.

3.5 REVIEW PROCEDURES

Organization of the reviews addressed by this chapter of the SRP will differ depending on the

scope of the documents submitted. For a license application, renewal, or amendment

application containing a new or revised chapter addressing the applicant’s safety program and

ISA commitments, there may only be a primary ISA reviewer. However, for an initial ISA

Summary submittal, the primary ISA reviewer will be assisted by specialists in the various

safety disciplines and management measures. An ISA Summary update submitted as part of

an amendment for a process that has hazards in multiple disciplines would also require a team

approach. In general, there will be a primary ISA reviewer who evaluates generic methods,

risk, and reliability criteria used in the ISA, and generic information about individual processes. 

This primary reviewer will be assisted by secondary reviewers who evaluate selected individual

accidents, and advise on the completeness of the accident list for specific safety disciplines.

3.5.1 Acceptance Review

For review of safety program commitments, including commitments pertaining to the ISA and

ISA Summary (a renewal or amendment application), the primary ISA reviewer will conduct a

review to determine if the submittal contains appropriate information addressing each of the

areas of review identified in Section 3.3.1 of this chapter. If the application does not contain

sufficient information addressing the areas of review to permit a safety evaluation, the

application will not be accepted for review. 

For an ISA Summary, the primary ISA reviewer will also conduct an acceptance review to

determine whether the document submitted contains sufficient information addressing the

“Areas of Review” noted in Section 3.3.2, including specifically each of the elements required by

10 CFR 70.65(b), to permit an evaluation of safety for compliance with the regulations. If

sufficient information is not present, the ISA Summary will not be accepted for review. 
Integrated Safety Analysis 3-29 NUREG-1520

3.5.2 Safety Evaluation

3.5.2.1 Evaluation of Safety Program and ISA Commitments

The reviewer(s) examine the descriptions and commitments to program elements in the

application or other documents for the “Areas of Review” described in Section 3.3.1 to

ascertain whether the program elements are sufficient to meet the acceptance criteria of

Section 3.4.3.1. The ISA reviewer must coordinate his or her review, with reviews being

conducted under other chapters of this SRP.

3.5.2.2 Evaluation of ISA Summary 

Evaluation of the ISA Summary to determine if the acceptance criteria of Section 3.4.3.2 have

been met would normally be performed by a team consisting of a primary reviewer together with

specialists in each category of accidents. These categories of accidents depend on the facility

but, in general, are nuclear criticality, fires, chemical accidents, and radiological accidents. If

external event analysis is complex, specialists may be employed to review these separately, as

well. The primary ISA reviewer would normally evaluate the acceptability of the generic

elements of the ISA Summary, such as site and facility descriptions, ISA methods, criteria, and

consequence and likelihood definitions. However, each specialist should also review these

elements to obtain information in support of his or her own evaluations. 

In contrast to these generic ISA elements, process-specific information is needed by, and must

be acceptable to, all of the specialists. Thus, the process descriptions in the ISA Summary

should be evaluated by all of the team members. 

Reviews of accident sequence descriptions and the likelihood and consequence information

showing compliance with 10 CFR 70.61 should be undertaken by separate specialists for each

category of accidents (i.e. nuclear criticalities, fires, radiological releases, and chemical

accidents). As indicated in Appendix A to this chapter, one acceptable format for the ISA

Summary is to separately tabulate or give logic diagrams for accident sequences in each

accident category. 

After a preliminary team review of the ISA Summary, a visit to the facility would normally be

made for familiarization with the 3-D geometry of process equipment, to review components of

the ISA, and to address any issues that arose during review of the ISA Summary.

To select a subset of the accident sequences reported in the ISA Summary for more detailed

review, the reviewer(s) should look at the applicant's tabulation of high and intermediate

consequence accident sequences and the types of IROFS designated for each. High

consequence accident sequences protected by administrative controls should be examined

very carefully, whereas intermediate consequence accident sequences protected by redundant

passive engineered controls warrant a lesser degree of scrutiny. Selection of specific accident

sequences and IROFS for more detailed evaluation should then be made using the following

approach.

The reviewer(s) should evaluate potential accidents using information supplied in the ISA

Summary. The applicant=s method for identifying and establishing the consequences and
Integrated Safety Analysis 3-30 NUREG-1520

likelihood of an accident sequence may provide information sufficient for this purpose. The

NRC reviewer(s) may evaluate the accidents using qualitative screening criteria analogous to

Table A-6 in Appendix A of this chapter. Other more rigorous reliability or consequence

analyses may be performed as deemed necessary. On the basis of this analysis, accidents will

be categorized. Engineered and administrative controls for accidents appearing in the highest

category may be selected for review in greater detail. While onsite, the reviewer(s) should also

select for specific evaluation a small sample of accident sequences determined by the applicant

to either result in less than intermediate consequences or to be not credible.

From the list of the IROFS, the reviewer(s) should categorize IROFS so that items of a similar

nature are grouped together. The reviewer(s) should then ensure that he or she has a full

understanding of one or more prototype IROFS selected from each category. For these

selected prototypes, the reviewer(s) may, if necessary, request additional information to

completely understand a particular IROFS. For complex processes, the reviewer(s) may need

to visit the facility to reach an adequate understanding of how the IROFS work for the process. 

3.5.2.3 Onsite ISA Review

The reviewer(s) should plan on visiting the applicant’s facility at least once as part of the

application review process. This visit should be scheduled after the applicant’s ISA Summary

has received a preliminary review. The visits will enable the reviewer(s) to confirm through

detailed examination of the ISA and ISA documentation that the ISA method(s) were selected

and applied in a reasonable and thorough manner to all facility processes, that all credible high

and intermediate consequence accident sequences were correctly identified, that accident

sequence consequences and likelihoods were reasonably determined, and that appropriate

IROFS and supporting management measures have been proposed. By means of a

“horizontal” review and several “vertical” slice reviews (defined below) of processes selected by

the reviewer(s), the completeness and adequacy of the applicant’s ISA method(s) can be

established. The reviewer(s) may use the ISA documentation to perform independent

evaluations of process hazards and accident sequences using methods selected from 

NUREG-1513, Appendix A to this SRP chapter, or other NRC guidance.

Reviewer(s) should not attempt a comprehensive, all-encompassing review of every facility

process and every accident sequence on the site visit. Rather, the reviewer(s) should use the

site visit to confirm the appropriateness and adequacy of the applicant’s ISA method(s) and the

completeness of the ISA and accuracy of analysis of accident sequences by means of a

“horizontal” review and several “vertical” slice reviews of selected processes. The site visit will

also afford the reviewer(s) an opportunity to seek answers to questions from the applicant (or

possibly the ISA team) that may have arisen in the preliminary review of the ISA Summary.

Each of the three facets of the onsite ISA review are discussed below.

ISA Methods Review

The purpose of the ISA method(s) review is two-fold: (1) to ensure that the applicant selected

appropriate ISA method(s) for each facility process, and (2) to ensure that they were correctly

applied in conducting the ISA. Descriptions of the ISA method(s) and a few example

applications of the ISA method(s) should be provided in the ISA Summary. The ISA method(s)

review should answer any questions that the reviewer(s) may have concerning ISA methods
Integrated Safety Analysis 3-31 NUREG-1520

and procedures after completing a preliminary review of the ISA Summary. In reviewing

process-specific information in the ISA Summary and ISA documentation maintained onsite, a

few processes and accident sequences should be selected to review the adequacy of the

selected ISA method(s) and its/their application. The reviewer(s) should examine any

procedures, checklists, or guidance documents that the applicant may have onsite as guidance

to ISA team members to ensure a complete understanding of the applicant’s ISA methods. The

reviewer(s) should then examine the ISA documentation, including the selected processes and

accident sequences, showing how the ISA methods were applied as part of the horizontal and

vertical slice reviews discussed below.

Horizontal Review

The basic purpose of the horizontal review is to ensure completeness of the ISA of facility

processes. This does not require an absolute checkoff of ISA documentation against the full

list of processes to be covered, but does mean that a substantial fraction of the processes

should receive a brief examination.

Reviewer(s) should consult the ISA and ISA documentation to answer questions or to resolve

outstanding issues resulting from the preliminary review of the ISA Summary. If the ISA

Summary includes sufficiently detailed information for a process, further examination of the

onsite ISA documentation may not be required. In particular, the reviewer(s) should examine

safety information that is not included in the ISA Summary. For example, ISA documentation

related to hardware IROFS, such as system schematics and/or descriptive lists, should contain

sufficient detail about hardware IROFS, such that it is clear to the reviewer(s) what components

(such as cables, detectors, alarms, valves, piping, etc.) are included within the boundary of the

hardware IROFS system and would therefore be subject to management measures specified by

the applicant. In addition, such documentation should also identify support systems (such as

backup batteries, air supply, steam supply, etc.) on which the IROFS depends to perform its

intended function. The reviewer(s) should also examine a few processes to confirm that all

accident sequences were considered and that those having potential consequences exceeding

the performance requirements of 10 CFR 70.61 are included in the ISA Summary.

Vertical Slice Review

The purpose of the vertical slice review is to examine how the ISA method(s) were applied to a

selected subset of facility processes. For this subset of facility processes, the reviewer(s)

should examine the underpinnings of calculations, conclusions, and the design of safety

programs that result from the ISA as well as safety information that is not identified in the ISA

Summary. The reviewer(s) should examine accident sequences for this subset of processes to

determine the adequacy of the applicant’s consequence and likelihood determinations. In

addition, the reviewer(s) should examine the appropriateness and robustness of designated

IROFS and the suitability of proposed management measures.

The ISA Summary review will have categorized accidents according to their consequences,

likelihoods, and IROFS. The subset of processes for vertical slice review should be selected

from these categories. The subset should include accident sequences of relatively high levels

of consequence and likelihood and accident sequences to which IROFS of different types and

relatively low robustness are designated. Vertical slice reviews should be performed on

processes for which less robust IROFS are designated (e.g., greater reliance on administrative

rather than engineered controls). While onsite, the reviewer(s) may confirm the adequacy of
Integrated Safety Analysis 3-32 NUREG-1520

sample accident analyses that the applicant included in the ISA Summary. However, the

reviewer(s) should focus on processes and/or accident sequences that were not included as

sample accident analyses in the ISA Summary to ensure the completeness of the ISA. 

The vertical slice review should address any specific questions the reviewer(s) may have

related to the ISA methods. If the applicant’s methods are evaluated as effective in these

selected cases, there is greater assurance that they will be effective for other processes. If

questions or weaknesses are discovered that may be of a generic nature, the reviewer(s) may

have to perform vertical slice analyses on several additional processes. However, a specific

question on the ISA of one process may not imply that there is a generic question requiring

further examination. The purpose of the vertical slice reviews is not complete verification of ISA

implementation.

The total number of vertical slice reviews to be conducted will depend on the facility’s total

number of accident sequences for which the consequences could exceed the performance

requirements of 10 CFR 70.61, the diversity of the types of processes at the facility, and the

results of initial reviews of the ISA Summary and the horizontal and vertical slice reviews. For

most fuel fabrication facilities, the reviewer(s) should plan on conducting vertical slice reviews

for 5 to 10 NCS-significant processes, 1 to 3 fire-significant processes, and 1 to 3

chemical/radiological/environmental-significant processes. But if the initial reviews of the ISA

Summary and the horizontal and vertical slice reviews identify significant issues, then additional

vertical slice reviews may be warranted.

Another criterion for selecting the process subset is prior accident and precursor experience

showing vulnerability to design weakness. For example, 21 of 22 process criticality accidents

have occurred in solution systems. Exothermic chemical reaction processes have frequently

resulted in accidents. Thus, the reviewer(s) should include these types of processes and

accident sequences in the subset for detailed review. Another criterion for selection is safety

designs where high reliability is inherently difficult to achieve. Examples are (1) designs with

high dependence on correct operator action and (2) complex active engineered control

systems.

Each vertical slice review should include (1) familiarization of the reviewer(s) with the safety

design of the selected process and (2) examination of all onsite documentation related to the

ISA of that process. If the content of the documentation leaves certain issues unclear,

interviews with facility personnel may be necessary. The review should focus on the

information onsite that is not provided in the ISA Summary, but is key to understanding

compliance with 10 CFR 70.61 requirements. 

Following the horizontal and vertical slice reviews, if outstanding questions remain about

compliance with the performance requirements of 10 CFR 70.61, the reviewer(s) may conduct

an independent evaluation using appropriate methods selected from NUREG-1513, Appendix A

to this chapter, or other agency guidance. The purpose of such an independent review is to

identify strengths and weaknesses of the applicant’s ISA methods or implementation practices,

not simply to check compliance in this one case per se. 

The reviewer(s) should take care to document findings and evaluations made during this

process.
Integrated Safety Analysis 3-33 NUREG-1520

3.6 EVALUATION FINDINGS

The reviewer verifies that the information submitted by the applicant is sufficiently complete so

that compliance with the regulations can be evaluated. There should be a finding statement,

following the evaluation of each area of review, stating how and why the information submitted

in that area complies with the related regulatory requirement. Specifically, the reviewer’s(s’)

findings in the SER should state conclusions of the types described in the following paragraphs: 

General conclusion resulting from the reviewer’s(s’) evaluation of Safety Program

commitments:

The NRC staff concludes that the applicant=s safety program, if established and maintained

pursuant to 10 CFR 70.62, is adequate to provide reasonable assurance that IROFS will be

available and reliable to perform their intended safety function(s) when needed and in the

context of the performance requirements of 10 CFR 70.61.

There should be general findings for each of the areas of review that state how the applicant=s

information demonstrates compliance with the acceptance criteria of Section 3.4.3.1. If the

reviewer(s) find(s) that the acceptance criteria are not met, a license condition rectifying the

deficiency should be recommended. If the applicant has submitted an adequate explanation of

an alternative way of complying with the regulations, the NRC’s SER should contain a finding

that the alternative is acceptable to meet the basic regulatory requirement addressed. 

General conclusions resulting from the staff’s evaluation of an ISA Summary:

Many hazards and potential accidents can result in unintended exposure of persons to

radiation, radioactive materials, or toxic chemicals incident to the processing of licensed

materials. The NRC staff finds that the applicant has performed an ISA to identify and

evaluate those hazards and potential accidents as required by the regulations. The NRC

staff has reviewed the ISA Summary and other information, and finds that it provides

reasonable assurance that the applicant has identified IROFS and established engineered

and administrative controls to ensure compliance with the performance requirements of 

10 CFR 70.61. Specifically, the NRC staff finds that the ISA results, as documented in the

ISA Summary, provide reasonable assurance that the IROFS, the management measures,

and the applicant=s programmatic commitments will, if properly implemented, make all

credible intermediate consequence accidents unlikely, and all credible high consequence

accidents highly unlikely.

Findings should be made concerning any specific requirements statements in 10 CFR Part 70

that address the nine elements in the ISA Summary. In particular, these findings should include

statements concerning compliance with the requirements of 10 CFR 70.64 (regarding new

facilities and new processes at existing facilities).

Findings may be made concerning compliance of specific processes with requirements of 

10 CFR 70.61 or other parts of the regulation, for those processes that receive specific detailed

review. However, such findings should be limited to a finding of reasonable assurance that a

process having the IROFS, as described in the ISA Summary, is capable of meeting the

requirements if properly implemented, operated, and maintained. 
Integrated Safety Analysis 3-34 NUREG-1520

3.7 REFERENCES

American Institute of Chemical Engineers (AIChE), AGuidelines for Hazard Evaluation

Procedures, Second Edition with Worked Examples,@ New York, September 1992. 

American National Standards Institute, American Nuclear Society, ANuclear Criticality Safety in

Operations with Fissionable Materials Outside Reactors,A ANSI/ANS-8.1-1983, La Grange Park,

IL, 1983. 

U.S. Code of Federal Regulations, Title 10, Part 70, “Domestic Licensing of Special Nuclear

Material,” U.S. Government Printing Office, Washington, DC.

U.S. Department of Commerce, Bureau of the Census, AStatistical Abstract of the United

States,@ Table No. 688, 1995.

U.S. Nuclear Regulatory Commission, AIntegrated Safety Analysis Guidance Document,@

NUREG-1513, 1995.

U.S. Environmental Protection Agency, “Acute Exposure Guidelines,” 40 CFR Part 68,

Appendix A.

U.S. Nuclear Regulatory Commission, “Nuclear Fuel Cycle Facility Accident Analysis

Handbook, NUREG/CR-6410, March 1998.

U.S. Nuclear Regulatory Commission, “Integrated Safety Analysis Guidance Document,”

NUREG-1513, May 2001.
Appendix A to Chapter 3 3-A-1 NUREG-1520

APPENDIX A

EXAMPLE PROCEDURE FOR ACCIDENT SEQUENCE EVALUATION

The appendix provides the NRC reviewer with a method for reviewing integrated safety

analyses (ISAs). For the applicant, this appendix outlines one approach for performing ISA

analyses of process accident sequences. It employs a semi-quantitative risk index method for

categorizing accident sequences in terms of their likelihood of occurrence and their

consequences of concern. The risk index method framework will enable the applicant to

identify, and the NRC reviewer to confirm, which accident sequences have consequences that

exceed the performance requirements of 10 CFR 70.61 and, therefore, require designation of

items relied on for safety (IROFS) and supporting management measures. Descriptions of

these general types of higher consequence accident sequences need to be reported in the 

ISA Summary.

This appendix works through an example of how the Risk Index Method can be applied to a

uranium powder blender. It describes one method of evaluating compliance with the

consequence and likelihood performance requirements of 10 CFR 70.61. The method is

intended to permit quantitative information to be considered, if available. For consistency, the

NRC reviewer's approach could also include assigning quantitative values to any qualitative

likelihood assessments made by an applicant since likelihoods are inherently quantitative. This

method should not be interpreted as requiring that an applicant use quantitative evaluation. 

However, evaluation of a particular accident should be consistent with any facts available, which

may include quantitative information concerning the availability and reliability of IROFS involved.

This appendix is not a “format and content guide” for either the ISA or the ISA Summary. It

simply presents one method of analysis and categorization of credible accident sequences for

facility processes. The method of this appendix describes both qualitative and quantitative

criteria for evaluating frequency indices of safety controls. These criteria for assigning indices,

particularly the descriptive criteria provided in some tables of this appendix, are intended to be

examples, not universal criteria. It is preferable that each applicant develop such criteria, based

on the particular types of IROFS and management measure programs. The applicant should

modify and improve such criteria as insights are gained during performance of the ISA.

If the applicant evaluates accidents using a different method, the method should produce

similar results in terms of how accidents are categorized. The method should be regarded as a

screening method, not as a definitive method of proving the adequacy or inadequacy of the

IROFS for any particular accident. Because methods can rarely be universally valid, individual

accidents for which this method does not appear applicable may be justified by an evaluation

using other methods. The method does have the benefit that it evaluates, in a consistent

manner, the characteristics of IROFS used to limit accident sequences. This will permit

identification of accident sequences with defects in the combination of IROFS used. Such

IROFS can then be further evaluated or improved to establish adequacy. The procedure also

ensures the consistent evaluation of similar IROFS by different ISA teams. Sequences or

IROFS that have risk significance and are evaluated as marginally acceptable are good

candidates for more detailed evaluation by the applicant and the reviewer.

The tabular accident summary resulting from the ISA should identify, for each sequence, what

engineered or administrative IROFS must fail to allow the occurrence of consequences that
Appendix A to Chapter 3 3-A-2 NUREG-1520

exceed the levels identified in 10 CFR 70.61. Chapter 3 of this Standard Review Plan (SRP)

specifies acceptance criteria for these IROFS and for meeting the performance requirements of 

10 CFR 70.61. These criteria require that IROFS be sufficiently unlikely to fail. However, the

acceptance criteria do not explicitly mandate any particular method for assessing likelihood. 

The purpose of this appendix is to provide an example of an acceptable method to perform this

evaluation of likelihood.

A.1 RISK MATRIX DEVELOPMENT

Consequences 

10 CFR 70.61 specifies two categories for accident sequence consequences: “high

consequences” and “intermediate consequences.” Implicitly there is a third category for

accidents that produce consequences less than “intermediate.” These will be referred to as

“low consequence” accident sequences. The primary purpose of process hazard analysis

(PHA) is to identify all uncontrolled and unmitigated accident sequences. These accident

sequences can then be categorized into one of these three consequence categories (high,

intermediate, low) based on their forecast radiological, chemical, and/or environmental impacts. 

Although the subsequent ISA analysis focuses only on those accident sequences having high

or intermediate consequences, by identifying and tabulating “low consequence” events in the

ISA, the reviewer can evaluate the completeness of the PHA and ISA analyses. Table A-1

presents the radiological and chemical consequence severity limits of 10 CFR 70.61 for each of

the three accident consequence categories.

Table A-1: Consequence Severity Categories Based on 10 CFR 70.61

Workers Offsite Public Environment

Category 3

High

Consequence

*RD> 1 Sievert (Sv)

(100 rem)

**CD = endanger life

RD> 0.25 Sv (25 rem)

30 mg sol U intake

CD = long-lasting

health effects

Category 2

Intermediate 

Consequence 

0.25 Sv (25 rem)

<RD# 1 Sv (100 rem)

CD = long-lasting

health effects

0.05 Sv (5 rem) <RD#

0.25 Sv (25 rem)

CD = mild transient

health effects

Radioactive release

>5000 x Table 2 of 

10 CFR Part 20,

Appendix B

Category 1 

Low 

Consequence 

Accidents of lower

radiological and

chemical exposures

than those above in

this column 

Accidents of lower

radiological and

chemical exposures

than those above in

this column

Radioactive releases

producing lower

effects than those

referenced above in

this column

 * RD = Radiological Dose

** CD = Chemical Dose

Note: > = greater than, < = less than, and # = less than or equal to
Appendix A to Chapter 3 3-A-3 NUREG-1520

Likelihood 

10 CFR 70.61 also specifies the permissible likelihood of occurrence of accident sequences of

different consequences. “High consequence” accident sequences must be “highly unlikely” and

“intermediate consequence” accident sequences must be “unlikely.” Implicitly, accidents in the

“low consequence” category can have a likelihood of occurrence less than “unlikely” or simply

“not unlikely.” Table A-2 shows the likelihood of occurrence limits of 10 CFR 70.61 for each of

the three likelihood categories.

Table A-2: Likelihood Categories Based on 10 CFR 70.61

Qualitative Description

Likelihood Category 1 Consequence Category 3 accidents must be

“highly unlikely”

Likelihood Category 2 Consequence Category 2 accidents must be

“unlikely”

Likelihood Category 3 “Not unlikely”

Risk Matrix 

The three categories of consequence and likelihood can be displayed as a 3 x 3 risk index

matrix. By assigning a number to each category of consequence and likelihood, a qualitative

risk index can be calculated for each combination of consequence and likelihood. The risk

index equals the product of the integers assigned to the respective consequence and likelihood

categories. The risk index matrix, along with computed risk index values, is illustrated in Table

A-3. The shaded blocks identify accidents of which the consequences and likelihoods yield an

unacceptable risk index and for which IROFS must be applied.

Table A-3: Risk Matrix With Risk Index Values

Severity of

Consequences

Likelihood of Occurrence

Likelihood Category 1

Highly Unlikely 

(1)

Likelihood Category 2

Unlikely 

(2)

Likelihood Category 3

Not Unlikely

(3)

Consequence

Category 3 High

 (3)

Acceptable Risk

3

Unacceptable Risk

6

Unacceptable Risk

9

Consequence 

Category 2

Intermediate (2)

Acceptable Risk

2

Acceptable Risk

4

Unacceptable Risk

6

Consequence 

Category 1 Low

(1)

Acceptable Risk

1

Acceptable Risk

2

Acceptable Risk

3
Appendix A to Chapter 3 3-A-4 NUREG-1520

The risk indices can initially be used to examine whether the consequences of an uncontrolled

and unmitigated accident sequence (i.e., without any IROFS) could exceed the performance

requirements of 10 CFR 70.61. If the performance requirements could be exceeded, the

applicant must designate IROFS to prevent the accident or to mitigate its consequences to an

acceptable level. A risk index value less than or equal to 4 means the accident sequence is

acceptably protected and/or mitigated. If the applicant provides this risk index in the ISA and

ISA Summary, the reviewer can quickly scan these data to confirm that each accident

sequence meets the performance requirements of 10 CFR 70.61.

If the risk index of an uncontrolled and unmitigated accident sequence exceeds 4, the likelihood

of the accident must be reduced through designation of IROFS. In this risk index method the

likelihood index for the uncontrolled and unmitigated accident sequence is adjusted by

subtracting a score corresponding to the type and number of IROFS that have been

designated. Table A-4 lists the qualitative scores assigned to the four types of IROFS.

Reviewers should note that the qualitative scores assigned in Table A-4 are for illustrative

purposes only. IROFS meeting the criteria for a particular score in Table A-4 could have a wide

range of availability or reliability. Such coarse criteria are useful for screening purposes, but

when the total evaluated likelihood score for an accident sequence lies near the acceptance

guideline value, a more careful evaluation should be done. Such evaluations should consider

the management measures applied to all the reliability and availability qualities of the IROFS, or

system of IROFS, protecting against the accident, as explained in the likelihood acceptance

criteria of this chapter in Subsections 5 and 7 of Section 3.4.3.2.

Table A-4: Qualitative Categorization of IROFS

Numeric Value Description of IROFS

1 Protection by a single trained operator with adequate response time

(Administrative lROFS)

2 Protection by a single active engineered IROFS, functionally tested on a

regular basis

(Active Engineered lROFS)

3 Protection by a single passive-engineered IROFS, functionally tested on a

regular basis, or by an active engineered IROFS with a to trained operator for

back-up

(Passive Engineered IROFS or Combined Engineered and

Administrative IROFS)

4 Protection by two independent and redundant engineered IROFS, as

appropriate, functionally tested on a regular basis

(Combination of Two Active or Passive Engineered IROFS)

To demonstrate compliance with the performance requirements of 10 CFR 70.61, the ISA

should assign a consequence category to each identified accident sequence. The likelihood of

occurrence of those accident sequences identified as high or intermediate consequence events

must then be assigned to one of the three likelihood categories. To be acceptable, the

controlled and/or mitigated accident consequences and likelihoods must have valid bases, and
Appendix A to Chapter 3 3-A-5 NUREG-1520

the applicant must include the bases for all general types of high and intermediate

consequence accident sequences in the ISA Summary.

A.2 CONSEQUENCE CATEGORY ASSIGNMENT

Categorization of an accident sequence as a high-consequence event or an intermediateconsequence event, or neither, is based on the estimated consequences of prototype

accidents. Although accident consequences can be determined by actual calculations,

calculations need not be performed for each individual accident sequence listed for a process. 

Accident consequences may also be estimated by comparison to similar events for which

reasonably bounding conservative calculations have been made. Categorization also requires

consideration of acute chemical exposures that an individual could receive from licensed

material or hazardous chemicals incident to the processing of licensed material. The applicant

must select appropriate acute chemical exposure data and relate these data to the performance

requirements of 10 CFR 70.61(b)(4) and (c)(4). In this appendix, the Acute Exposure Guideline

Level (AEGL) and Emergency Response Planning Guideline (ERPG) are used. AEGL-3 and

ERPG-3 levels are life-threatening.

Consequence Category 3 (High-Consequences) includes accidents resulting in any

consequence specified in 10 CFR 70.61(b). These include (1) acute worker exposures of 

(a) radiation doses greater than 1 Sievert (100 rem) total effective dose equivalent (TEDE), and

(b) chemical exposures that could endanger life (above AEGL-3 or ERPG-3), and (2) acute

exposures to members of the public outside the controlled area to (a) radiation doses greater

than 0.25 Sievert (25 rem) TEDE, (b) soluble uranium intakes greater than 30 milligram, and 

(c) chemical exposures that could lead to irreversible or other serious long-lasting health effects

(exceeding AEGL-2 or ERPG-2). An unshielded nuclear criticality would normally be

considered a “high consequence” event because of the potential for producing a high radiation

dose to a worker.

Consequence Category 2 (Intermediate-Consequences) includes accidents resulting in any

consequence specified in 10 CFR 70.61(c). These include (1) acute exposures of workers to

(a) radiation doses between 0.25 Sievert (25 rem) and 1 Sievert (100 rem) TEDE, and 

(b) chemical exposures that could lead to irreversible or other serious long-lasting health effects

above AEGL-2 or ERPG-2), and (2) acute exposures of members of the public outside the

controlled area to (a) radiation doses between 0.05 Sievert (5 rem) and 0.25 Sievert (25 rem)

TEDE, (b) chemical exposures that could cause mild transient health effects (exceeding AEGL

or ERPG-1), and (3) release of radioactive material outside the restricted area that would, if

averaged over a 24-hour period, exceed 5000 times the values specified in Table 2 of 

Appendix B to 10 CFR Part 20.

Consequence Category 1 (Low-Consequences) includes accidents with potential adverse

radiological or chemical consequences, but at exposures less than Categories 3 and 2.

This system of consequence categories is shown in Table A-5.
Appendix A to Chapter 3 3-A-6 NUREG-1520

Table A-5: Consequence Severity Categories Based on 10 CFR 70.61

Workers Offsite Public Environment

Category 3

High 

Consequence 

*RD>1 Sievert (Sv) 

(100 rem)

**CD>AEGL-3,

ERPG-3

RD>0.25 Sv (25 rem)

30 mg sol U intake

CD>AEGL-2, ERPG-2

Category 2

Intermediate 

Consequence 

0.25 Sv (25 rem)

<RD# 1 Sv (100 rem)

AEGL-2, ERGP-2

<CD# AEGL-3,

ERPG-3

0.05 Sv(5 rem) < RD#

0.25 Sv (25 rem)

AEGL-1, ERGP-1

<CD# AEGL-2,

ERPG-2

Radioactive release >

5000 x Table 2

Appendix B of

10 CFR Part 20

Category 1

Low 

Consequence 

Accidents of lower

radiological and

chemical exposures

than those above in

this column

Accidents of lower

radiological and

chemical exposures

than those above in

this column

Radioactive releases

with lower effects

than those referenced

above in this column

* RD - Radiological Dose

 **CD - Chemical Dose

Note: > = greater than, < = less than, and # = less than or equal to

The applicant should document the bases for bounding calculations of the consequence

assignment in the ISA Summary submittal. NUREG/CR-6410, “Nuclear Fuel Cycle Facility

Accident Analysis Handbook,” March 1998, describes valid methods and data that may be used

by the applicant or staff for confirmatory evaluations.

A.3 LIKELIHOOD CATEGORY ASSIGNMENT

An assignment of an accident sequence to a likelihood category is acceptable if it is based on

the record of occurrences at the facility, the record of failures of IROFS at the facility, or on

other methods that have objective validity. Because sequences leading to accidents often

involve multiple failures, the likelihood of the whole sequence will depend on the frequencies of

initiating events and failure likelihoods of engineered and administrative IROFS. The method of

likelihood assignment used in this appendix relies on the expert engineering judgment of the

analyst and includes assessment of the number, type, independence, and observed failure

history of designated IROFS. Engineered and administrative IROFS, even those of the same

types, have a wide range of reliability. By requiring explicit consideration of most of the

underlying events and factors that significantly affect the likelihood of the accident and explicit

criteria for assigning likelihood, greater consistency in assigning likelihood to accident

sequences across different systems within a facility and among different applicants should 

be possible.

Quantitative measures of likelihood are based on the NRC's determinations reported in Item 9

of Section 3.4.3.2 of SRP Chapter 3: “highly unlikely” means a frequency of less than 
Appendix A to Chapter 3 3-A-7 NUREG-1520

10-5 per-event per-year, and “unlikely” means a frequency within the range of 10-4 and 

10-5 per-event per-year. The numerical scores assigned to each likelihood of occurrence are

presented in Table A-6.

Table A-6: Event Likelihood

Likelihood Category Probability of Occurrence*

Not Unlikely 3 more than 10-4 per-event per-year

Unlikely 2 between 10-4 and 10-5 per-event peryear

Highly Unlikely 1 less than 10-5 per-event per-year

*Based on approximate order-of-magnitude ranges

In assessing the adequacy of engineered and administrative IROFS, individual accident

frequencies greater than 10-5 per-year may not be evaluated as “highly unlikely.” Similarly,

accident sequences having frequencies more than 10-4 per-year may not be considered 

as “unlikely.”

The accident evaluation method described below does not preclude the need to comply with the

double-contingency principle for sequences leading to criticality. Although exceptions are

permitted with compensatory measures, double contingency should, in general, be applied. 

Double contingency is needed as there are usually insufficient firm data as to the reliability of

the IROFS equipment and administrative IROFS procedures used in criticality safety. If only

one item were relied on to prevent a criticality, and it proved to be less reliable than expected,

then the first time it failed, a criticality accident could result. For this reason, at least two

independent IROFS should be used. Inadequate IROFS can then be determined by observing

their failures without also suffering the consequences of a criticality accident. Even with double

contingency, each IROFS should be sufficiently unlikely to fail, for if one of the two items that

establish double contingency is actually ineffective, criticality will still be unlikely.

A.4 ASSESSING EFFECTIVENESS OF IROFS

The risk of an accident sequence is reduced through application of different numbers and types

of IROFS. By either reducing the likelihood of occurrence or by mitigating the consequences,

IROFS can reduce the overall resulting risk. The designation of IROFS should generally be

made to reduce the likelihood (i.e., prevent an accident), but the consequences may also be

reduced by minimizing the potential hazards (e.g., quantity) if practical. Based on hazards

identification and accident sequence analyses for which the resulting unmitigated or

uncontrolled risks are unacceptable, key safety controls (administrative and/or engineered

IROFS) may be designated as IROFS to reduce the likelihood of occurrence and/or mitigate the

consequence severity.
Appendix A to Chapter 3 3-A-8 NUREG-1520

A.5 RISK INDEX EVALUATION SUMMARY

As previously mentioned, an acceptable way for the applicant to present the results of the ISA

is a tabular summary of the identified accident sequences. Table A-7 is an acceptable format

for such a table. This table lists several example accident sequences for a powder blender at a

typical facility. Table A-7 summarizes two sets of information: (1) the accident sequences

identified in the ISA; and (2) a risk index, calculated for each sequence, to show compliance

with the regulation. The risk index calculation is summarized below.

Accident sequences result from initiating events, followed by failure of one or more IROFS. 

Thus, in Table A-7 there are columns for the initiating event and for IROFS. IROFS may be

mitigative or preventive. Mitigative IROFS are measures that reduce the consequences of an

accident. The phrase “uncontrolled and/or unmitigated consequences” describes the results

when the system of preventive IROFS fails and mitigation also fails. Mitigated consequences

result when the preventive IROFS fail, but mitigative measures succeed. These are

abbreviated in the table as “unmit.” and “mitig.,” respectively. Index numbers are assigned to

initiating events, IROFS failure events, and mitigation failure events, based on the reliability

characteristics of these items.

With redundant IROFS and in certain other cases, there are sequences in which an initiating

event places the system in a vulnerable state. While the system is in this vulnerable state, an

IROFS must fail for the accident to result. Thus, the frequency of the accident depends on the

frequency of the first event, the duration of vulnerability, and the frequency of the second

IROFS failure. For this reason, the duration of the vulnerable state should be considered, and

a duration index should be assigned. The values of all index numbers for a sequence,

depending on the number of events involved, are added to obtain a total likelihood index, T. 

Accident sequences are then assigned to one of the three likelihood categories of the risk

matrix, depending on the value of this index in accordance with Table A-8.

The values of index numbers in accident sequences are assigned considering the criteria in

Tables A-9 through A-11. Each table applies to a different type of event. Table A-9 applies to

events that have frequencies of occurrence, such as initiating events and certain IROFS

failures. When failure probabilities are required for an event, Table A-10 provides the index

values. Table A-11 provides index numbers for durations of failure. These are used in certain

accident sequences where two IROFS must simultaneously be in a failed state. In this case,

one of the two controlled parameters will fail first. It is then necessary to consider the duration

that the system remains vulnerable to failure of the second. This period of vulnerability can be

terminated in several ways. The first failure may be “fail-safe” or be continuously monitored,

thus alerting the operator when it fails so that the system may be quickly placed in a safe state. 

Or the IROFS may be subject to periodic surveillance tests for hidden failures. When hidden

failures are possible, these surveillance intervals limit the duration that the system is in a

vulnerable state. The reverse sequences, where the second IROFS fails first, should be

considered as a separate accident sequence. This is necessary because the failure frequency

and the duration of outage of the first and the second IROFS may differ. The values of these

duration indices are not merely judgmental. They are directly related to the time intervals used

for surveillance and the time needed to render the system safe.
Appendix A to Chapter 3 3-A-9 NUREG-1520

Table A-7: Example Accident Sequence Summary and Risk Index Assignment

Process: uranium dioxide(UO2) powder preparation (PP); Unit Process: additive blending; Node: blender hopper node (PPB2)

Accident

identifier

Initiating

Event

(a)

Preventive Safety

Parameter 1

 or IROFS 1

Failure/Success 

(b)

Preventive

Safety

Parameter 2

or IROFS 2

Failure/Success

(c)

Mitigation

IROFS

Failure/

Success

(d)

Likelihood*

Index T

uncontrolled/

controlled

(e)

Likelihood

Category

(f)

Conseque

nce

Evaluation

Reference

Consequence

Category 

(g)

Risk Index

(h=f x g)

uncontrolled/

controlled

(h)

Comments 

&

Recommendations

PPB2-1A

(Criticality

from

blender

leak of

UO2)

See IROFS 1

(Note 1)

PPB2-C1: Mass

Control

Failure:

Blender leaks

UO2 onto floor,

critical mass

exceeded

Frq1 = -1 

Dur1 = -4

PPB2-C2: 

Moderation

Failure:

Suffic. Water

for criticality

introduced

while UO2 on

floor: 

Frq2 = -2

N/A Unc T = -1 

Con T = -7

Unc 3 

Con I

Rad 35 3 

(Crit: 3,

rad: 0)

9 

3

Criticality,

consequences = 3

IROFS 2 fails while

IROFS 1 is in failed

state.

T = -1-4-2 = -7

PPB2-1B 

(Rad.

release

from

blender

leak of

UO2)

Blender leaks

UO2

Frqi = -1

PPB2-C1: Mass

Control

 Success: leaked

UO2

 below critical

mass

PPB2-C2: 

Moderation

Success: no

moderator

Ventilation

Failure:

Ventilated

blender

enclosure

Prf = -3

Unc T = -1

Con T = -4 

Con T = -1

Unc 3

Unmit 2 

Mitig 3

Rad 36 Unc 2

Unmit 2

Mitig 1

6

 Unmit 4

Mitig 3

Rad consequences, no

criticality unmitigated

sequence: IROFS 1 &

mitigation fail.

T= -1-3 = -4

Mitig: IROFS 1 fails,

mitig IROFS does not

fail. T = -1

PPB2-1C See 

IROFS 1 

(Note 1)

PPB2-C2: 

Moderation

Failure:

Suffic. water for

criticality

on floor under

UO2 blender

Frq1 = -2 Dur1 =

-3

PPB2-C1: 

Mass Control

Failure:

Blender leaks

UO2 on floor

while water

present 

Frq2 = -2

N/A Unc T = -2

Con T = -6

Unc 2

Con 1

Rad 35 3

(Crit: 3, rad: 0)

6 

3

Criticality by reverse

sequence of PPB2-1A.

Moderation fails first. 

Note different

likelihood. T = -6

PPB2-2 Fire in

Blender Room

Frqi = -2

Fire Suppression

Failure:

Fails on demand:

Prf1 = -2

N/A N/A Unc T = -2 

Con T = -4

Unc 2

Con 2

Rad 37 2

(rad)

1

4

2

Event sequence is just

initiating event plus one

IROFS failure on

demand

*Likelihood index T is a sum; uncontrolled: T=frqi or frq1; controlled: includes all indices

T=a+b+c+d.

Note 1: For these sequences the initiating event is failure of one of the IROFS, hence the

frequency is assigned under that IROFS.

Table A-8: Determination of Likelihood Category

Likelihood Category Likelihood Index T* (= sum of index numbers)

1 T # -5

2 -5 < T # -4

3 -4 < T

*The likelihood category is determined by calculating the likelihood index, T, then using

this table. The term T is calculated as the sum of the indices for the events in the

accident sequence.
Appendix A to Chapter 3 3-A-10 NUREG-1520

Table A-9: Failure Frequency Index Numbers

Frequency

Index No.

Based on

Evidence

Based on Type of IROFS** Comments

-6 * External event

with freq. < 10-6 /yr

If initiating event, no

IROFS needed.

-4 * No failures in 30

years for

hundreds of

similar IROFS in

industry

Exceptionally robust passive

engineered IROFS (PEC), or an

inherently safe process, or two

independent active engineered

IROFS (AECs), PECs, or enhanced

admin. IROFS

Rarely can be justified by

evidence. Further, most

types of single IROFS

have been observed to fail.

-3 * No failures in 30

years for tens of

similar IROFS in

industry

A single IROFS with redundant

parts, each a PEC or AEC

-2 * No failure of this

type in this facility

in 30 years 

A single PEC

-1 A few failures may

occur during

facility lifetime

A single AEC, an enhanced admin.

IROFS, an admin. IROFS with

large margin, or a redundant

admin. IROFS

0 Failures occur

every 1 to 3 years

A single administrative IROFS

1 Several

occurrences per

year

Frequent event, inadequate IROFS Not for IROFS, just

initiating events

2 Occurs every

week or more

often

Very frequent event, inadequate

IROFS

Not for IROFS, just

initiating events

* Indices less than (more negative than) -1 should not be assigned to IROFS unless the

configuration management, auditing, and other management measures are of high quality, because,

without these measures, the IROFS may be changed or not maintained. 

** The index value assigned to an IROFS of a given type in column 3 may be one value higher or

lower than the value given in column 1. Criteria justifying assignment of the lower (more negative)

value should be given in the narrative describing ISA methods. Exceptions require individual

justification.
Appendix A to Chapter 3 3-A-11 NUREG-1520

Table A-10: Failure Probability Index Numbers

Probability

Index No.

Probability of

Failure on

Demand

Based on Type of IROFS Comments

-6* 10-6 If initiating event, no

IROFS needed.

-4 or -5* 10-4 - 10-5 Exceptionally robust passive engineered

IROFS (PEC), or an inherently safe process,

or two redundant IROFS more robust than

simple admin. IROFS (AEC, PEC, or

enhanced admin.)

Can rarely be justified by

evidence. Most types of

single IROFS have been

observed to fail.

-3 or -4* 10-3 - 10-4 A single passive engineered IROFS (PEC) or

an active engineered IROFS (AEC) with high

availability

-2 or -3* 10-2 - 10-3 A single active engineered IROFS, or an

enhanced admin. IROFS, or an admin.

IROFS for routine planned operations

-1 or -2 10-1 - 10-2 An admin. IROFS that must be performed in

response to a rare unplanned demand

*Indices less than (more negative than) -1 should not be assigned to IROFS unless the

configuration management, auditing, and other management measures are of high quality,

because, without these measures, the IROFS may be changed or not maintained.

Table A-11: Failure Duration Index Numbers

Duration

Index

No.

Avg. Failure Duration Duration in Years Comments

1 More than 3 years 10

0 1 year 1

-1 1 month 0.1 Formal monitoring to justify

indices less than -1

-2 A few days 0.01

-3 8 hours 0.001

-4 1 hour 10-4

-5 5 minutes 10-5
Appendix A to Chapter 3 3-A-12 NUREG-1520

As shown in Table A-11, the duration of failure is accounted for in establishing the overall

likelihood that an accident sequence will continue to the defined consequence. Thus, the time

to discover and repair the failure is accounted for in establishing the risk of the postulated

accident. Accordingly, as long as the actual undiscovered failures and repair times in service

are conservatively described by the applicant's chosen duration of failure index and as long as

the defined risks (reported in the ISA Summary) associated with the consequences are

acceptable pursuant to 10 CFR 70.61, then such failures do not imply a violation of the

approved license.

For all these index numbers, the more negative the number is the less likely the failure. 

Accident sequences may consist of varying numbers of events, starting with an initiating event. 

The total likelihood index is the sum of the indices for all the events in the sequence, including

those for duration.

Consequences are assigned to one of the three consequence categories of the risk matrix,

based on calculations or estimates of the actual consequences of the accident sequence. The

consequence categories are based on the levels identified in 10 CFR 70.61. Multiple types of

consequences can result from the same event. The consequence category is chosen for the

most severe consequence.

As shown in the first row of Table A-7, the failure duration index can make a large contribution

to the total likelihood index. Therefore, the reviewer should verify that there is adequate

justification that the failure will be corrected in the time ascribed to the duration index. In

general, duration indices with values less than minus one (-1), corresponding to 36 days,

should be based on intentional monitoring of the process. The duration of failure for an

unmonitored process should be conservatively estimated.

Table A-7 provides two risk indices for each accident sequence to permit evaluation of the risk

significance of the IROFS involved. To measure whether an IROFS has high risk significance,

the table provides an “uncontrolled risk index,” determined by modeling the sequence with all

IROFS as failed (i.e., not contributing to a lower likelihood). In addition, a “controlled risk index”

is also calculated, taking credit for the low likelihood and duration of IROFS failures. When an

accident sequence has an uncontrolled risk index exceeding 4 but a controlled index of less

than 4, the IROFS involved have a high risk significance because they are relied on to achieve

acceptable safety performance. Thus, use of these indices permits evaluation of the possible

benefit of improving IROFS and also whether a relaxation may be acceptable.

Table A-12 provides a more detailed description of the accident sequences used in the example

of Table A-7. The reviewer needs the information in Table A-12 to understand the nature of the

accident sequences listed in Table A-7. Table A-7 lacks room to explain any but the simplest

failure events.

Table A-13 explains the IROFS and external initiating events that appear in the accident

sequences in Table A-7. The reviewer needs the information in Table A-13 to understand why

the initiating events and IROFS listed in Table A-7 have the low likelihood indices assigned. 

Thus, Table A-13 should contain such information as (1) the margins to safety limits, (2) the

redundancy of an IROFS, and (3) the measures taken to ensure adequate reliability of an

IROFS. Table A-13 must also justify why external events, which are not obviously extremely

unlikely, have the low likelihoods that are being relied on for safety. The applicant should

provide separate tables to list the IROFS for criticality, chemical, fire, radiological, and
Appendix A to Chapter 3 3-A-13 NUREG-1520

environmental accidents. If an applicant chooses to classify IROFS by applying different levels

or grades of quality assurance, then the applicant should also provide the appropriate quality

assurance grade for the IROFS.

Table A-12: Accident Sequence Descriptions

Process: uranium dioxide (UO2) powder preparation (PP) Unit: additive blending

Node: blender hopper node (PPB2)

Accident

(see Table A-6)

Description

PPB2-1A

Blender UO2

leak criticality

The initial failure is a blender leak of UO2 that results in a mass sufficient for criticality on the

floor. (This event is not a small leak.) Before the UO2 can be removed, moderator sufficient to

cause criticality is introduced. Duration of critical mass UO2 on floor estimated to be 1 hour.

PPB2-1B

Blender UO2

leak, rad.

release

The initial failure is a blender leak of UO2 that results in a mass insufficient for criticality on the

floor or a mass sufficient for criticality but moderation failure does not occur. Consequences

are radiological, not a criticality. A ventilated enclosure should mitigate the radiological release

of UO2. If It fails during cleanup or is not working, unmitigated consequences occur.

PPB2-1C The events of PPB2-1A occur in reverse sequence. The Initial failure is introduction of water

onto the floor under the blender. Duration of this flooded condition is 8 hours. During this time,

the blender leaks a critical mass of UO2 onto the floor. Criticality occurs.

PPB2-2 Initiating event is a fire in the blender room. Fire is not extinguished in time. UO2 is released

from process equipment. Offsite dose estimated to exceed 1 mSv (100 mrem).

Table A-13: Descriptive List of Items Relied on for Safety

Process: uranium dioxide (UO2) powder preparation (PP) Unit: additive blending

Node: blender hopper node (PPB2)

IROFS

Identifier

Safety

Parameter

and Limits

IROFS Description Max Value of

Other

Parameters

Reliability

Management

Measures

Quality

Assurance

Grade

PPB2-C1 Mass outside

hopper: zero

Mass outside hopper: Hopper

and outlet design prevent UO2

leaks, double gasket at outlet

Full water

reflection,

enrichment 5%

Surveillance for

leaked UO2

each shift

A

PPB2-C2 Moderation:

in UO2 < 1.5

wt. %

External

water in area: 

zero

Moderation in UO2: Two sample

measurements by two persons

before transfer to hopper

External water: Posting

excluding water, double piping in

room, floor drains, roof integrity

Full water

reflection,

enrichment 5%

Drain, roof, and

piping under 

safety-grade

maintenance

A

Note: In addition to engineered IROFS, this table should include descriptions of external

initiating events of which the low likelihood is relied on to achieve acceptable risk, especially

those which are assigned frequency indices lower than -4. The descriptions of these initiating

events should contain information supporting the frequency index value selected by 

the applicant.
Appendix A to Chapter 3 3-A-14 NUREG-1520

A.6 ACCIDENT SUMMARY AND RISK INDEX ASSIGNMENT FOR TABLE A-7

The definitions for the contents of each column in the accident summary tabulation, Table A-7,

are provided below.

Accident Identifier

This column identifies the accident sequence being analyzed. The ISA will have all accident

sequences for uniquely identified facility processes, referred to here as “nodes.” Symbols,

names, or numbers of these nodes permit them to be uniquely identified. For example, the

“blender hopper” node described in Table A-7 has the unique identifying symbol PPB2. PPB2-

1is the first accident sequence identified in that node. By reviewing example accident

sequences presented in the ISA Summary and the selected accident sequences contained in

the ISA, the reviewer(s) can evaluate and confirm (1) the adequacy of the IROFS for preventing

accidents, and (2) the bases for making the consequence and likelihood assignments in the

table.

Initiating Event (Column a)

This column lists initiating events or IROFS failures that are typically identified in the PHA

phase of the ISA and that may lead to consequences exceeding those identified in 10 CFR

70.61. Initiating events are of several distinct types: (1) external events, such as hurricanes

and earthquakes; (2) facility events external to the node being analyzed (e.g., fires, explosions,

failures of other equipment, flooding from facility water sources); (3) deviations from normal

operations of the process in the node (i.e., credible abnormal events); and (4) failures of IROFS

of the node. The tabulated initiating events should only consist of those that involve an actual

or threatened failure of IROFS or that cause a demand requiring IROFS to function to prevent

consequences exceeding 10 CFR 70.61 levels. The frequency index number for initiating

events is referred to in the table by the symbol “frqi.” Table A-9 provides criteria for assigning a

value to frqi. Usually, there is insufficient room in a tabular presentation like Table A-7 to

describe events accurately. Consequently, the applicant should provide supplementary

narrative information to adequately describe each general type of accident sequence in 

Table A-7. Cross-referencing between this information and the table should be adequate (e.g.,

the unique symbolic accident sequence identifiers can be used). Table A-12 is an example of a

list of supplementary accident sequence descriptions corresponding to Table A-7. 

Preventive IROFS 1 Failure/Success (Column b)

This column addresses the failure or success of the safety parameter designated to prevent

consequences exceeding 10 CFR 70.61 levels. Specific IROFS that may be needed to

maintain the safety parameter should be included in this table. If separate parameters or

IROFS are used to prevent different consequences, separate rows in the table should be

defined corresponding to each type of consequence. Table A-7 contains an example of a set of

related sequences so separated. Accident sequences where two IROFS must simultaneously

be in a failed state require assignment of three index numbers (1) the failure frequency of the

first IROFS, frq1, (2) the duration of this failure, dur1, and (3) the failure frequency of the

second IROFS, frq2. For such accident sequences, the initiating event is failure of the first

IROFS. In these cases, frq1 is assigned using Table A-9. The failure duration of the first

IROFS is assigned using Table A-11. Other accident sequences may be more easily described

as a failure of the IROFS on demand after the occurrence of an initiating event. In these cases,

the failure probability index number, prf1, is assigned using Table A-10.
Appendix A to Chapter 3 3-A-15 NUREG-1520

Preventive IROFS 2 Failure/Success (Column c)

This column is provided in case a second preventive IROFS is designated. The failure

frequency or failure probability on demand is assigned in the same manner as for preventive

IROFS 1.

Mitigation IROFS Failure/Success (Column d)

This column is provided in case IROFS are available to mitigate the consequences of the

accident sequence. That is, they reduce, but do not eliminate, the consequences of an

accident sequence. An IROFS that eliminates all adverse consequences should be 

considered preventive.

Likelihood Index (Column e) and Likelihood Category (Column f)

This column lists the likelihood category number for the risk matrix, which is based on the total

likelihood index for an accident sequence. The total likelihood index, T, is the sum of the

indices for those events that comprise an accident sequence which normally consists of the

initiating event and failure of one or more IROFS, including any failure duration indices. 

However, accident sequences may consist of varying numbers and types of undesired events. 

Methods for deciding what frequencies and failure durations need to be considered will be

described later in this appendix. Based on the sum of these indices, the likelihood category

number for the risk matrix is assigned using Table A-8.

Consequence Evaluation Reference

This column permits identification of the consequence calculations that relate to this accident

sequence. Multiple references may be required to refer to calculations of the different types of

consequences (e.g., radiological, chemical, etc.).

Consequence Category (Column g)

This column is provided to assign the consequence category numbers from the risk matrix

based on estimating the consequences of all types (i.e., radiological, criticality, chemical, and

environmental) of accident sequences that may occur. Accident sequences having IROFS to

mitigate consequences must be divided into two cases, one where the mitigation succeeds, and

one where it fails, each with different consequences. The two cases may be tabulated in one

row of Table A-7, but the mitigated and unmitigated consequences should be separately

indicated. Unless the mitigated case results in consequences below those levels identified in 

10 CFR 70.61, both cases must satisfy the likelihood requirements as shown by the risk matrix.

Risk Index (Column h)

This column is provided to list the risk index, which is calculated as the product of the likelihood

category and consequence category numbers. This is shown in the column heading by the

formula “h = f x g.” Sequences with values of h less than or equal to 4 are acceptable. Another

risk index can also be calculated as the product of the consequence category number times the

likelihood category associated with only the failure frequency index for the initiating event. The

resulting product can be referred to as the “unmitigated” risk index. It is unmitigated in the

sense that no credit is taken for the functioning of any subsequent IROFS. For example, in the

first three cases in Table A-7, the initiating event is failure of preventive IROFS 1. In these

cases, the failure frequency of preventive IROFS 1 is used to determine the likelihood category

when calculating the unmitigated risk index.
Appendix A to Chapter 3 3-A-16 NUREG-1520

Comments and Recommendations

This column records ISA team recommendations. It is especially useful when the existing

system of IROFS is evaluated as being deficient. This may happen because a newly identified

accident sequence is not addressed by existing IROFS or because an unacceptable

performance deficiency has been found in the existing IROFS.

A.7 DETERMINATION OF LIKELIHOOD CATEGORY IN TABLE A-8

The likelihood category is determined by calculating the likelihood index, T, which equals the

sum of the indices for the events in the accident sequence. Based on the calculated value of T,

the likelihood category of each accident sequence can be determined from Table A-8.

A.8 DETERMINATION OF FAILURE FREQUENCY INDEX NUMBERS IN TABLE A-9

Table A-9 is used to assign frequency index numbers to plant initiating events and IROFS failures

as found in the columns of Table A-7. The term “failure” must be understood to mean not merely

failure of the IROFS but also a violation of the process safety. In the example in Table A-7, accident

sequence PPB2-1A involves loss of mass control over uranium dioxide (UO2) in a blender. If

criticality is the concern, failure does not occur unless UO2 accumulates to a critical mass before the

leak is stopped. For radiological consequences, any amount leaked may cause exposure. In

assessing the frequency index, this factor should be considered because many IROFS failures do

not cause safety limits to be exceeded.

Table A-9 provides two columns with two sets of criteria for assigning an index value, one based on

type of IROFS, the other on observed failure frequencies. Since IROFS of a given type have a wide

range of failure frequencies, assignment of index values based on this table should be done with

caution. Due consideration should be given to whether the IROFS will actually achieve the

corresponding failure frequency in the next column. Based on operational experience, more refined

criteria for judging failure frequencies may be developed by an individual applicant. In the column

labeled “Based on Type of IROFS,” references to redundancy allow for IROFS that may themselves

have internal redundancy to achieve a necessary level of reliability.

Another objective basis for assignment of an index value is actual observations of failure events. 

These actual events may have occurred in the applicant’s facility or in a comparable process

elsewhere. Justification for specific assignments may be noted in the Comments column 

of Table A-7. 

As previously noted, the definition of “failure” of an IROFS to be used in assigning indices is, for

nonredundant IROFS, a failure severe enough to cause an accident with consequences exceeding

those of 10 CFR 70.61. For redundant IROFS, it is a failure such that, if no credit is taken for

functionality of the IROFS, an accident with consequences exceeding the performance

requirements of 10 CFR 70.61 could result. If most IROFS malfunctions qualify as such failures, the

index assignments of this table are appropriate. If true failure is substantially less frequent, credit

should be taken and adequate justification provided.

Note that indices less than (more negative than) -1 should not be assigned to IROFS unless the

configuration management, auditing, and other required management measures are of high quality,

because, without these measures, the IROFS may be changed or inadequately maintained. The

reviewer should be able to determine this from the tabular summary of IROFS provided in the

application. This summary should identify the process parameters to be controlled and their safety

limits and include a thorough description of the IROFS and the applied management measures.
Appendix A to Chapter 3 3-A-17 NUREG-1520

A.9 DETERMINATION OF FAILURE PROBABILITY INDEX NUMBERS IN TABLE A-10

Occasionally, information concerning the reliability of an IROFS may be available as a probability on

demand. That is, there may be a history of tests or incidents where the system in question is

demanded to function. To quantify such accident sequences, the demand frequency, the initiating

event, and the demand failure probability of the IROFS must be known. This table provides an

assignment of index numbers for such IROFS in a way that is consistent with Table A-9. The

probability of failure on demand may be the likelihood that it is in a failed state when demanded

(availability) or that it fails to remain functional for a sufficient time to complete its function.

A.10 DETERMINING MANAGEMENT MEASURES FOR IROFS

Table A-13 is an acceptable way of listing IROFS in all the general types of accident sequences

having consequences exceeding those identified in 10 CFR 70.61. The items listed should

include all IROFS and all external events whose low likelihood of occurrence is relied upon to

meet the performance requirements of 10 CFR 70.61. The reviewer(s) examine this list to

determine whether adequate management measures have been applied to each IROFS to

ensure its continual availability and reliability, in conformance to 10 CFR 70.62(d). The types of

management measures are maintenance, training, configuration management, audits and

assessments, quality assurance, etc. Certain criteria for management measures are indicated

in the baseline design criteria; others are described in greater detail in SRP Chapters 4 through

7 and Chapter 11. IROFS meeting all the provisions of these chapters have acceptable

management measures. IROFS may, with justification, have lesser management measures

than those described. However, every IROFS in accident sequences leading to Consequence

Categories 2 or 3 should be assigned at least a minimal set of management measures. 

Specifically, to defend against common mode failure of all IROFS on a process, this minimal set

of measures must include (a) adequate configuration management, (b) regular auditing for the

continued effectiveness of the IROFS, (c) adequate labeling, training, or written procedures to

ensure that the operating staff is aware of the safety function, (d) adequate surveillance and

corrective maintenance, and (e) adequate preventive maintenance.

If lesser or graded management measures are applied to some IROFS, Tables A-7 and A-13

and the narratives associated with them must identify to which IROFS these lesser measures

are applied. In addition, information indicating that acceptable reliability can be achieved with

these lesser measures must be presented. The specifics of how each management measure,

such as the surveillance interval, type of maintenance, or type of testing, is applied to each

IROFS need not be provided, for the NRC recognizes that such specific measures must be

applied differently to each IROFS to achieve adequate reliability. The formality, documentation,

and quality assurance requirements applied to these direct management measures that may be

graded generically in a risk-informed manner must be documented.

The following paragraphs describe the application of management measures to IROFS based

on the risk importance of the item in an accident sequence, as defined by (1) the uncontrolled 

risk index shown in Table A-7, and (2) the accident likelihood index, T, also described 

in Table A-7:

! For a particular accident sequence that would have unmitigated consequences in the

two highest categories identified in 10 CFR 70.61, IROFS should reduce the risk from

initially high risk (an uncontrolled risk index of 6 or 9 from Table A-3) to an acceptable

risk (controlled risk index of less than or equal to 4).
Appendix A to Chapter 3 3-A-18 NUREG-1520

! For those sequences that are initially evaluated as being in an acceptable risk category

(an uncontrolled risk index of less than or equal to 4), a more detailed discussion is

necessary. Some such accidents could have a relatively high uncontrolled likelihood

(see discussion under B below), yet be of low consequence so that the risk is

acceptable without IROFS. However, if the accident consequence of interest is a

nuclear criticality, the accident’s likelihood must be “highly unlikely,” irrespective of the

expected magnitude of consequence. Further, for accident sequences resulting in

nuclear criticality, double contingency should be achieved, thus requiring at least one

more IROFS and an initiating event of low probability. With this exception for criticality

sequences, the following three cases apply:

A. If the initiating event is not an IROFS failure, assurances for IROFS are not

necessary. No additional risk reduction is required. However, for accident sequences

claimed to be highly unlikely, the assessment that the initiating event has such a low

probability must be adequately justified in the application.

B. If the initiating event is an IROFS failure, and if the likelihood of that failure is taken

to be more than a few times per facility lifetime (T is greater than -2), assurances for

that IROFS may be less. Any subsequent IROFS in the accident sequence will be

unregulated.

The rationale is that since T is greater than -2, the likelihood category is 3. Therefore,

the consequence category should not be greater than 1, to limit the uncontrolled risk

index to, at most, 4. As the consequence category is low, the assurance level can be

reduced.

C. If the initiating event is an IROFS failure, and if the likelihood of that failure is taken

to be less than a few times per facility lifetime (T is less than or equal to -2), assurance

for this IROFS must satisfy the full baseline design requirements. No regulation of

subsequent IROFS in the sequence is necessary.

The rationale is that as T is less than or equal to -2, the likelihood category must be 1

or 2. Therefore, the consequence category must be no greater than 2 to limit the

uncontrolled risk index to 4 at most. In this case, the uncertainty in determining a low

failure likelihood requires compensatory measures in the form of increased assurances

(high-level criteria) that the IROFS is indeed kept at a low failure likelihood.
Appendix A to Chapter 3 3-A-19 NUREG-1520

A.11 RISK-INFORMED REVIEW OF IROFS

Column (h) in Table A-7 gives the risk indices for each accident sequence that was identified in

the ISA. There are two indices, uncontrolled and controlled. The controlled index is a measure

of risk without credit for the IROFS. If the uncontrolled risk index is a 6 or 9, while the

controlled index is an acceptable value (4 or less), the set of IROFS involved are significant in

achieving acceptable risk. That is, these IROFS have high risk significance. The uncontrolled

risk index will be used by the reviewer(s) to identify all risk-significant systems of IROFS. 

These systems of IROFS will be reviewed more closely than IROFS established to prevent or

mitigate accident sequences of low risk.

Accession #: ML013370311
Radiation Protection 4-1 NUREG-1520

4 RADIATION PROTECTION

4.1 PURPOSE OF REVIEW

The purpose of this review is to determine whether the applicant’s radiation protection program

is adequate to protect the radiological health and safety of workers and to comply with the

regulatory requirements in 10 CFR Parts 19, 20, and 70.

The content and level of detail in this chapter are more detailed because this chapter provides

acceptance criteria for evaluating compliance with 10 CFR Part 20, which has very specific

requirements. Nevertheless the applicant is expected and the NRC reviewer should accept

insights gained from the conduct of the Integrated Safety Analysis (ISA) and information

contained in the ISA Summary in developing and reviewing the acceptability of the applicant’s

radiation protection program. Review procedures and acceptance criteria for the applicant’s

program for protecting members of the public and controlling effluent releases are presented in

Chapter 9, “Environmental Protection,” of this SRP. In Chapter 3, ”Integrated Safety Analysis

(ISA) and ISA Summary,” there are criteria for performing a comprehensive ISA at a fuel cycle

facility. In performing an ISA, an applicant will evaluate and rank the radiological risks posed by

potential accident sequences throughout the facility and assess the adequacy of the items

relied on for safety (IROFS) to ensure that the radiation exposure performance criteria of 

10 CFR 70.61(b) and (c) are met.

4.2 RESPONSIBILITY FOR REVIEW

Primary: Health Physicist

Secondary: Licensing Project Manager, Environmental Reviewer 

Supporting: Fuel Cycle Facility Inspector

4.3 AREAS OF REVIEW

The radiation protection program must address the occupational radiation protection measures

in 10 CFR Parts 19, 20, and 70. Specifically, licensees must develop, document, and

implement a radiation protection program in accordance with 10 CFR 20.1101. Additionally, 

10 CFR 20.2102 requires licensees to keep records of the radiation protection program,

including a description of the program components, audits, and other aspects of program

implementation. The reviewer should also review the ISA Summary to identify those facility

operations, analyzed in the ISA, that have radiological consequences, and the IROFS and the

management measures implemented to prevent or mitigate such radiological risks. 

The staff will review an applicant’s commitments to:

(1) establish, maintain, and implement a radiation protection program

(2) keep occupational exposures to radiation as low as reasonably achievable (ALARA)

(3) appoint radiological protection staff who are suitably qualified and trained in radiation

protection procedures

(4) prepare written radiation protection procedures and radiation work permits (RWPs)
Radiation Protection 4-2 NUREG-1520

(5) train employees in radiation protection, including the health protection problems

associated with exposure to radiation, precautions and procedures to minimize exposure,

and the purposes and functions of protective devices employed

(6) design and implement programs to control airborne concentrations of radioactive material

by using ventilation systems, containment systems, and respirators

(7) conduct radiation surveys and monitoring programs to document radiation levels,

concentrations of radioactive materials in the facility, and occupational exposures to

radiation by workers

(8) maintain additional programs including (a) a records maintenance program, (b) a

corrective action program, and (c) a program for reporting to the NRC in accordance with

requirements in 10 CFR Parts 20 and 70

4.4 ACCEPTANCE CRITERIA

4.4.1 Commitment to Radiation Protection Program Implementation

4.4.1.1 Regulatory Requirements

Regulations applicable to establishment of a radiation protection program are presented in 

10 CFR 20.110, Subpart B, “Radiation Protection Programs.”

4.4.1.2 Regulatory Guidance

The NRC regulatory guide applicable to the commitment to design and implement a radiation

protection program is Regulatory Guide 8.2, "Guide for Administrative Practice in Radiation

Monitoring,” February 2, 1973.

4.4.1.3 Regulatory Acceptance Criteria

The applicant’s radiation protection program is acceptable if the applicant provides data and

information in the license application that meets each of the following commitments to: 

(1) design and implement a radiation protection program that meets the regulatory

requirements of 10 CFR Part 20, Subpart B

(2) outline the radiation protection program structure and define the responsibilities of key

program personnel

(3) staff the radiation protection program with suitably trained people, provide sufficient

resources, and implement the program

(4) commit to the independence of the radiation protection function from the facility’s

operations

(5) review, at least annually, the content and implementation of the radiation protection

program as required by 10 CFR 20.1101(c) 
1

 The ALARA Committee should meet at least annually and the membership should

include management, radiation protection, environmental safety, industrial safety, production,

etc. personnel or representatives. The ALARA Committee will review the ALARA program and

the review should include an evaluation of the results of audits made by the radiation protection

organization, reports of radiation levels in the facility, contamination levels, employee

exposures, and effluent releases, etc. The review should determine if there are any upward

trends in personnel exposure for identified categories of workers and types of operations. The

review should identify any upward trends in effluent releases and contamination levels. Finally,

the review should determine if exposures, releases and contamination levels are in accordance

Radiation Protection 4-3 NUREG-1520

The review should consider facility changes, new technologies, or other process enhancements

that could improve the overall program effectiveness.

4.4.2 Commitment to an ALARA Program

4.4.2.1 Regulatory Requirements

Regulations applicable to the ALARA program are presented in 10 CFR 20.1101, “Radiation

Protection Programs.”

4.4.2.2 Regulatory Guidance

NRC regulatory guides applicable to the ALARA program are:

• Regulatory Guide 8.2, “Guide for Administrative Practice in Radiation

 February 2, 1973 Monitoring”

• Regulatory Guide 8.13, Rev. 3, “Instructions Concerning Prenatal Radiation 

 Draft DG 8014, October 1994 Exposure”

• Regulatory Guide 8.29, “Instructions Concerning Risks From Occupational

 February 1996 Radiation Exposure”

4.4.2.3 Regulatory Acceptance Criteria

The applicant’s ALARA program is acceptable if the applicant provides data and information in

the license application that meet each of the following commitments to:

(1) establish a comprehensive, effective, and written ALARA program

(2) prepare policies and procedures to ensure occupational radiation exposures are

maintained ALARA and that such exposures are consistent with the requirements of 

10 CFR 20.1101

(3) outline specific ALARA program goals, establish an ALARA program organization and

structure, and have written procedures for its implementation in the facility design and

operations

(4) establish an ALARA Committee, or equivalent organization, with sufficient staff,

resources, and clear responsibilities to ensure that the occupational radiation exposure

dose limits of 10 CFR Part 20 are not exceeded under normal operations1
with the ALARA concept. Recommendations of the ALARA Committee should be documented

and tracked to completion.

Radiation Protection 4-4 NUREG-1520

(5) use the ALARA program as a mechanism to facilitate interaction between radiation

protection and operations personnel

(6) regularly review and revise, when appropriate, the ALARA program goals and objectives

and to incorporate, when appropriate, new approaches, technologies, operating

procedures or changes that could reduce potential radiation exposures at a reasonable

cost

4.4.3 Organization and Personnel Qualifications

 

4.4.3.1 Regulatory Requirements

Regulations applicable to the organization and qualifications of the radiological protection staff

are presented in 10 CFR 70.22, “Contents of Applications.”

4.4.3.2 Regulatory Guidance

NRC regulatory guides applicable to the organization and personnel qualifications of radiation

protection program staff are:

• Regulatory Guide 8.2, "Guide for Administrative Practice in Radiation

 February 1973 Monitoring"

• Regulatory Guide 8.10, "Operating Philosophy for Maintaining Occupational 

 Rev. 1-R, May 1977 Radiation Exposures As Low As Is Reasonably 

Achievable"

4.4.3.3 Regulatory Acceptance Criteria 

The applicant’s commitment to organize and staff a radiation protection program is acceptable if

the applicant provides data and information in the license application that meet each of the

following commitments to:

(1) appoint suitably trained radiation protection personnel and to identify their authority and

responsibilities

(2) establish clear organizational relationships among the individual positions responsible for

the radiation protection program and other line managers

(3) appoint a suitably trained radiation protection program director (typically referred to as the

radiation safety officer) who has direct access to the facility manager, who is skilled in the

interpretation of data and regulations pertinent to radiation protection, who is familiar with

the operation of the facility and radiation protection concerns of the site, who is used as a

resource in radiation safety management decisions, and who will be responsible for

establishing and implementing the radiation protection program
2

 These procedures should be reviewed and revised as necessary, to incorporate any

facility or operational changes or changes to the facility’s ISA. Approval of the procedures

should be made by the radiation safety officer, or an individual who has the qualifications of the

radiation safety officer.

3

 RWPs should define the authorized activities, the level of approval required (a

radiation specialist as a minimum), information requirements, period of validity, expiration and

termination times, and the recordkeeping requirements.

Radiation Protection 4-5 NUREG-1520

(4) assign responsibility to the radiation protection program staff for implementation of the

radiation program functions

(5) describe the minimum training requirements and qualifications for the radiation protection

staff

4.4.4 Commitment to Written Procedures 

4.4.4.1 Regulatory Requirements

The regulations applicable to radiation protection procedures and RWPs are presented in 

10 CFR 70.22(8), “Contents of Applications.”

4.4.4.2 Regulatory Guidance

Regulatory guidance applicable to procedures and RWPs is Regulatory Guide 8.10, Rev. 1-R,

“Operating Philosophy for Maintaining Occupational Radiation Exposures As Low As Is

Reasonably Achievable” May 1977.

4.4.4.3 Regulatory Acceptance Criteria

The applicant’s commitment to prepare written radiation protection procedures and RWPs is

acceptable if the applicant provides data and information in the license application that meet

each of the following commitments to: 

(1) prepare written, approved radiation protection procedures to carry out activities related to

the radiation protection program

(2) specify how the radiation protection procedures will be prepared, authorized, approved,

and distributed2

(3) specify written, approved RWPs for activities involving licensed material that are not

covered by written radiation protection procedures3

4.4.5 Training

An applicant’s commitments to employee training are addressed in SRP Chapters 4 and 11. 

Chapter 4 addresses corporate radiation protection training programs. Chapter 11 addresses

training which serves as a management control for ensuring that an administrative control

IROFS is available and reliable when required.
Radiation Protection 4-6 NUREG-1520

4.4.5.1 Regulatory Requirements

The following regulations apply to the radiation safety training program:

1. 10 CFR 19.12 "Instructions to workers"

2. 10 CFR 20.2110 "Form of records”

4.4.5.2 Regulatory Guidance

NRC regulatory guides and American Society for Testing and Materials (ASTM) standards

pertaining to radiation protection training are:

• Regulatory Guide 8.10, "Operating Philosophy for Maintaining Occupational

 Rev. 1-R, May 1977 Radiation Exposures As Low As Is Reasonably

 Achievable"

• Regulatory Guide 8.13, "Instructions Concerning Prenatal Radiation 

 Draft DG-801 proposed Exposure"

 R-3, October 1994

• Regulatory Guide 8.29, "Instructions Concerning Risks From Occupational 

 Draft DG-8012 proposed Radiation Exposure"

 R-1, December 1994

• ASTM C986-89, "Developing Training Programs in the Nuclear Fuel

 reapproved 1995 Cycle"

• ASTM E1168-95 "Radiological Protection Training for Nuclear Facility 

 Workers"

4.4.5.3 Regulatory Acceptance Criteria

The applicant’s commitment to train its employees in radiation protection is acceptable if the

applicant provides data and information in the license application that meets each of the

following commitments to:

(1) design and implement an employee radiation protection training program that complies

with the requirements of 10 CFR Parts 19 and 20

(2) provide training, to all personnel and visitors entering restricted areas, that is

commensurate with the health risk to which they may be exposed, or to provide trained

escorts who have received training

(3) provide a level of training based on the potential radiological health risks associated with

that employee’s work responsibilities

(4) incorporate, in the radiation protection training program, the provisions in 10 CFR 19.12

and topics such as:

 
Radiation Protection 4-7 NUREG-1520

! correct handling of radioactive materials 

! minimization of exposures to radiation and/or radioactive materials

! access and egress controls and escort procedures 

! radiation safety principles, policies, and procedures 

! monitoring for internal and external exposures 

! monitoring instruments 

! contamination control, including protective clothing and equipment 

! ALARA and exposure limits 

! radiation hazards and health risks 

! emergency response 

(5) review the radiation protection training programs at least every 3 years and to conduct

refresher training at least every 3 years to address changes in policies, procedures,

requirements, and the facility ISA

(6) evaluate the effectiveness and adequacy of the training program curriculum and

instructors 

4.4.6 Ventilation and Respiratory Protection Programs

4.4.6.1 Regulatory Requirements

Regulations applicable to the ventilation and respiratory protection programs are presented in 

10 CFR Part 20, Subpart H, “Respiratory Protection and Controls to Restrict Internal Exposure

in Restricted Areas.”

4.4.6.2 Regulatory Guidance

The following NRC regulatory guides, American National Standard Institute (ANSI) standards,

and National Council on Radiation Protection and Measurements (NCRP) report are applicable

to the design of the ventilation and respiratory protection programs:

• Regulatory Guide 8.24, "Health Physics Surveys During Enriched 

Rev. 1, October 1979 Uranium-235 Processing and Fuel Fabrication"

• ANSI N510-1980 "Testing of Nuclear Air Cleaning Systems"

• ERDA 76-21 “Nuclear Air Cleaning Handbook” by C. A. Burchsted,

A. B. Fuller, and J. E. Kahn

• NCRP Report No. 59, "Operational Radiation Safety Program"

December 15, 1978

• Regulatory Guide 8.15 “Acceptable Programs for Respiratory Protection”

• ANSI Z88.2-1992 “Practices for Respiratory Protection”
Radiation Protection 4-8 NUREG-1520

4.4.6.3 Regulatory Acceptance Criteria

The applicant’s commitment to have ventilation and respiratory protection programs is

acceptable if the applicant provides data and information in the license application that meet

each of the following commitments to:

(1) install appropriately sized ventilation and containment systems in areas of the facility

identified in the ISA Summary as having potential airborne concentrations of radionuclides

that could exceed the occupational, derived air concentration values specified in 10 CFR

Part 20, Appendix B, during normal operations

(2) describe management measures, including preventive and corrective maintenance and

performance testing, to ensure that the ventilation and containment systems designated as

IROFS operate when required, and are within their design specifications

(3) describe the design criteria for the ventilation and containment systems, including minimum

flow velocity at openings in these systems, maximum differential pressure across filters

and types of filters to be used

(4) describe the frequency and types of tests to measure ventilation and containment system

performance, the acceptance criteria, and the actions to be taken when the acceptance

criteria are not satisfied

(5) establish a respiratory protection program that meets the requirements of 

10 CFR Part 20, Subpart H

(6) prepare written procedures for the selection, fitting, issuance, maintenance, testing,

training of personnel, monitoring, and recordkeeping for individual respiratory protection

equipment, and for specifying when such equipment is to be used

(7) revise the written procedures for use of individual respiratory protection equipment as

applicable, when processing, facility, or equipment changes are made

(8) maintain records of the respiratory protection program, including training for respirator use,

and maintenance

4.4.7 Radiation Surveys and Monitoring Programs 

Radiation surveys are conducted for two purposes: (1) to ascertain radiation levels,

concentrations of radioactive materials, and potential radiological hazards that could be present

in the facility; and (2) to detect releases of radioactive material from facility equipment and

operations. Radiation surveys will focus on those areas of the facility identified in the ISA

Summary where the occupational radiation dose limits could potentially be exceeded. 

Measurements of airborne radioactive material and/or bioassays are used to determine that

internal occupational exposures to radiation do not exceed the dose limits specified in 

10 CFR Part 20, Subpart C.

4.4.7.1 Regulatory Requirements

NRC regulations in 10 CFR Part 20 are applicable to radiation surveys and monitoring

programs:
Radiation Protection 4-9 NUREG-1520

1. Subpart F “Surveys and Monitoring”

2. Subpart C “Occupational Dose Limits”

3. Subpart L “Records”

4. Subpart M “Reports”

4.4.7.2 Regulatory Guidance

The following NRC regulatory guides, NUREGs, and ANSI standards are applicable to radiation

surveys and monitoring programs:

• Regulatory Guide 8.2, "Guide for Administrative Practice in Radiation

February 1973 Monitoring"

• Regulatory Guide 8.4, “Direct-Reading and Indirect-Reading Pocket 

February 1973 Dosimeters”

• Regulatory Guide 8.7, “Instructions for Recording and Reporting 

Rev. 1, June 1992 Occupational Radiation Exposure Data”

• Regulatory Guide 8.9, “Acceptable Concepts, Models, Equations, and

Rev. 1, July 1993 Assumptions for a Bioassay Program”

• Regulatory Guide 8.24, "Health Physics Surveys During Enriched 

 Rev. 1, October 1979 Uranium-235 Processing and Fuel Fabrication"

• Regulatory Guide 8.25, Rev. 1, "Air Sampling in the Workplace"

June 1992

• Regulatory Guide 8.34, “Monitoring Criteria and Methods To Calculate

July 1992 Occupational Radiation Doses”

• NUREG-1400, "Air Sampling in the Workplace"

September 1993

• ANSI N13.1-1969, "Guide to Sampling Airborne Radioactive Materials 

reaffirmed 1993 in Nuclear Facilities”

• ANSI N328-1978 “Radiation Protection Instrumentation Test and 

Calibration”

• ANSI N13.11-1983 “Dosimetry-Personnel Dosimetry PerformanceCriteria for Testing”

• ANSI N13.15-1985 “Radiation Detectors-Personnel 

 Thermoluminescence Dosimetry Systems- 

 Performance”

• ANSI.HPSN 13.22, 1995 “Bioassay Program for Uranium”

• ANSI N13.27-1981 “ “Performance Requirements for Pocket-Sized Alarm 
Radiation Protection 4-10 NUREG-1520

 Dosimeters and Alarm Ratemeters”

• ANSI.HPSN 13.30, 1996 “Performance Criteria for Radiobioassay”

• ANSI N13.6-1966, "Practice for Occupational Radiation Exposure

reaffirmed 1989 Records Systems”

4.4.7.3 Regulatory Acceptance Criteria 

The applicant’s commitment to implement radiation surveys and monitoring programs is

acceptable if the applicant provides data and information in the license application that meet

each of the following commitments to: 

(1) have radiation surveys and monitoring programs consistent with the requirements of 10

CFR Part 20, Subpart F

(2) prepare written procedures for the radiation survey and monitoring program that include

an outline of the program objectives, sampling procedures, data analysis methods, types

of equipment and instrumentation to be used, frequency of measurements, recordkeeping

and reporting requirements, and actions to be taken when measurements exceed 10 CFR

Part 20 occupational dose limits or administrative levels established by the applicant

(3) design and implement a personnel monitoring program for external occupational radiation

exposures that outlines methods or procedures to:

• identify the criteria for worker participation in the program

• identify the types of radiation to be monitored

• specify how exposures will be measured, assessed, and recorded

• identify the type and sensitivity of personal dosimeters to be used, when they will be

used, and how the collected data will be processed and evaluated

• identify the facility’s administrative exposure levels or action levels at which actions 

 are taken to investigate the cause of exposures exceeding these levels

(4) design and implement a personnel monitoring program, for internal occupational radiation

exposures, based on the requirements of 10 CFR 20.1201, 20.1204, and 20.1502(b), that

outlines methods or procedures to:

• identify the criteria for worker participation in the program

• identify the type of sampling to be used, the frequency of collection and

measurement, and the minimum detection levels 

• specify how worker intakes will be measured, assessed, and recorded 

• specify how the data will be processed, evaluated, and interpreted

• identify the facility’s administrative exposure levels or the levels at which actions are 

taken to investigate the causes of exposures exceeding these levels

(5) comply with the requirements of 10 CFR 20.1202 for summation of external and internal

occupational radiation exposures through the use of procedures such as those outlined in

Regulatory Guide 8.7 or 8.34

(6) design and implement an air sampling program in areas of the facility identified as

potential airborne radioactivity areas, to conduct air surveys, and to calibrate and maintain
Radiation Protection 4-11 NUREG-1520

the airborne sampling equipment in accordance with the manufacturers’

recommendations 

(7) implement additional procedures, as may be required by 10 CFR Part 20 and the ISA

Summary, to control the concentration of airborne radioactive material (e.g., control of

access, limitation of exposure times to licensed materials, and use of respiratory

protection equipment) 

(8) conduct a contamination survey program in areas of the facility identified in the ISA

Summary most likely to be radiologically contaminated (the program must include the

types and frequencies of surveys for various areas of the facility and the action levels and

actions to be taken when contamination levels are exceeded)

(9) implement the facility’s corrective action program when the results of personnel monitoring

or contamination surveys exceed the applicant’s administrative personnel contamination

levels

(10) implement the facility’s corrective action program when any incident results in airborne

occupational exposures to radiation exceeding the facility’s administrative limits, or the

dose limits in 10 CFR Part 20, Appendix B, or 10 CFR 70.61

(11) use equipment and instrumentation with sufficient sensitivity for the type or types of

radiation being measured and to calibrate and maintain equipment and instrumentation in

accordance with the manufacturers’ recommendations

(12) establish policies to ensure equipment and materials removed from restricted areas to

unrestricted areas are not contaminated above the specified release levels in NRC Branch

Technical Position, "Guidelines for Decontamination of Facilities and Equipment Prior to

Release for Unrestricted Use or Termination of Licenses for Byproduct, Source, or Special

Nuclear Material," April 1993

(13) leak-test all sealed sources in accordance with the following NRC Branch Technical

Positions: (1) "License Condition for Leak-Testing Sealed Byproduct Material Sources,"

April 1993, (2) "License Condition for Leak-Testing Sealed Plutonium Sources," April

1993, (3) "License Condition for Plutonium Alpha Sources," April 1993, (4) "License

Condition for Leak-Testing Sealed Source Which Contains Alpha and/or Beta-Gamma

Emitters," April 1993, and (5) "License Condition for Leak-Testing Sealed Uranium

Sources," April 1993

(14) establish and implement an access control program that ensures that (a) signs, labels,

and other access controls are properly posted and operative, (b) restricted areas are

established to prevent the spread of contamination and are identified with appropriate

signs, and (c) step-off pads, change facilities, protective clothing facilities, and personnelmonitoring instruments are provided in sufficient quantities and locations

(15) have a radiation reporting program consistent with the requirements of 10 CFR Parts 19

and 20
Radiation Protection 4-12 NUREG-1520

4.4.8 Additional Program Commitments

4.4.8.1 Regulatory Requirements

The following 10 CFR Part 20 regulations are applicable to the additional program

commitments:

1. Subpart L “Records”

2. Subpart M “Reports”

3. Section 70.61 ”Performance Requirements”

4. Section 70.74 ”Additional Reporting Requirements”

4.4.8.2 Regulatory Guidance

There are no NRC regulatory guidelines applicable to these additional program commitments.

4.4.8.3 Acceptance Criteria

The applicant’s commitment to implement additional program features is acceptable if the

applicant provides data and information in the license application that meet each of the

following commitments to: 

(1) maintain records of the radiation protection program (including program provisions,

audits, and reviews of the program content and implementation), radiation survey results

(air sampling, bioassays, external-exposure data from monitoring of individuals, internal

intakes of radioactive material), and results of corrective action program referrals,

RWPs and planned special exposures

(2) establish a program to report to the NRC, within the time specified in 10 CFR 20.2202

and 10 CFR 70.74, any event that results in an occupational exposure to radiation

exceeding the dose limits in 10 CFR Part 20 

(3) prepare and submit to the NRC an annual report of the results of individual monitoring,

as required by 10 CFR 20.2206(b)

(4) refer to the facility’s corrective action program any radiation incident that results in an

occupational exposure that exceeds the dose limits in 10 CFR Part 20, Appendix B, or is

required to be reported per 10 CFR 70.74, and to report to the NRC both the corrective

action taken (or planned) to protect against a recurrence and the proposed schedule to

achieve compliance with the applicable license condition or conditions

4.5 REVIEW PROCEDURES

4.5.1 Acceptance Review

The primary reviewer should evaluate the license application to determine whether it addresses

the areas of review discussed in Section 4.3. If significant deficiencies are identified, the

applicant should be requested to submit additional material before the reviewer(s) start the

safety evaluation. 
Radiation Protection 4-13 NUREG-1520

4.5.2 Safety Evaluation

The primary reviewer will perform a safety evaluation with respect to the acceptance criteria in

Section 4.4. For existing facilities, the reviewer will consult with the cognizant radiation

protection NRC inspector to identify and resolve any issues of concern related to the licensing

review. The primary reviewer will prepare an SER on the licensing action for the Licensing

Project Manager. 

4.6 EVALUATION FINDINGS

The reviewer will write an SER that addresses each topic reviewed and explains why the NRC

staff has reasonable assurance that the radiation protection part of the application is acceptable

and that the health and safety of the workers is adequately protected. License conditions may

be proposed to impose requirements where the application is deficient. The following kinds of

statements and conclusions will be included in the staff's SER:

The applicant has committed to an acceptable radiation protection program that

includes (1) an effective documented program to ensure that occupational

radiological exposures are ALARA; (2) an organization with adequate qualification

requirements for the radiation protection personnel; (3) approved written radiation

protection procedures and RWPs for radiation protection activities; (4) radiation

protection training for all personnel who have access to restricted areas; (5) a

program to control airborne concentrations of radioactive material with engineering

controls and respiratory protection; (6) a radiation survey and monitoring program

that includes requirements for controlling radiological contamination within the facility

and monitoring of external and internal radiation exposures; and (7) other programs

to maintain records, report to the NRC in accordance with 10 CFR Parts 20 and 70,

and correct for upsets at the facility. 

The NRC staff concludes that the applicant's radiation protection program is

adequate and meets the requirements of 10 CFR Parts 19, 20, and 70. 

Conformance to the license application and license conditions will ensure safe

operation.

4.7 REFERENCES

U.S. Code of Federal Regulations, Title 10, Part 70, "Domestic Licensing of Special Nuclear

Material," U.S. Government Printing Office, Washington, DC.

U.S. Code of Federal Regulations, Title 10, Part 20, “Standards for Protection Against

Radiation,” U.S. Government Printing Office, Washington, DC. 

U.S. Nuclear Regulatory Commission, Branch Technical Position, "License Condition for LeakTesting Sealed Plutonium Sources," April 1993.

U.S. Nuclear Regulatory Commission, Branch Technical Position, "License Condition for

Plutonium Alpha Sources," April 1993.

U.S. Nuclear Regulatory Commission, Branch Technical Position, "License Condition for LeakTesting Sealed Source Which Contains Alpha and/or Beta-Gamma Emitters," April 1993.
Radiation Protection 4-14 NUREG-1520

U.S. Nuclear Regulatory Commission, Branch Technical Position, "License Condition for LeakTesting Sealed Uranium Sources," April 1993.

U.S. Nuclear Regulatory Commission, Branch Technical Position, "Guidelines for

Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or

Termination of Licenses for Byproduct, Source, or Special Nuclear Material," April 1993.

Accession #: ML013370325
Nuclear Criticality Safety 5-1 NUREG-1520

5 NUCLEAR CRITICALITY SAFETY

5.1 PURPOSE OF REVIEW 

The purpose of this review is to determine whether the applicant’s nuclear criticality safety

(NCS) program is adequate to support safe operation of the facility, as required by 

10 CFR Part 70. The reviewers should examine the controls and barriers that are relied on to

prevent inadvertent nuclear criticalities and that are designated as IROFS in the applicant’s ISA

Summary. The NCS review should be coordinated with the onsite ISA review (see Section

3.5.2.3 of this SRP) to examine the NCS evaluations used to prepare the facility’s ISA.

5.2 RESPONSIBILITY FOR REVIEW 

Primary: Nuclear Process Engineer (NCS Reviewer) 

Secondary: None 

Supporting: Project Manager and Fuel Cycle Inspector (As needed.) 

5.3 AREAS OF REVIEW 

The staff should review the application to determine whether (1) the applicant has provided for

the appropriate management of the NCS program; (2) the applicant has identified and

committed to the responsibilities and authorities of individuals for developing and implementing

the NCS program; (3) the facility management measures described in 10 CFR 70.62 have been

committed to and will support implementing and maintaining the NCS program; and (4) an

adequate NCS program is described that includes identifying and committing to the

methodologies and technical practices used to ensure the safe operation of the facility, as

required by 10 CFR Part 70.

The specific areas for review are as follows: 

5.3.1 Management of the NCS Program 

The primary reviewer should review the application to determine whether the applicant has

committed to and implemented effective management of the NCS program and has provided

enough resources for an effective NCS program. The objectives of an effective NCS program

should include: 

(a) preventing an inadvertent nuclear criticality

(b) protecting against the occurrence of an identified accident sequence in the ISA

Summary that could lead to an inadvertent nuclear criticality

(c) complying with the NCS performance requirements of 10 CFR 70.61

(d) establishing and maintaining NCS safety parameters and procedures

(e) establishing and maintaining NCS safety limits and NCS operating limits for IROFS
Nuclear Criticality Safety 5-2 NUREG-1520

(f) conducting NCS evaluations to assure that under normal and credible abnormal

conditions, all nuclear processes are remain subcritical, and maintain an approved

margin of subcriticality for safety

(g) establishing and maintaining NCS IROFS, based on current NCS determinations 

(h) providing training in emergency procedures in response to an inadvertent nuclear

criticality

(i) complying with NCS baseline design criteria requirements in 10 CFR 70.64(a)

(j) complying with the NCS ISA Summary requirements in 10 CFR 70.65(b)

(k) complying with the NCS ISA Summary change process requirements

in 10 CFR 70.72

5.3.2 Organization and Administration

The reviewer(s) should determine whether the applicant has identified and committed to the

responsibilities and authorities of individuals to develop, organize, implement and administer the

NCS program. The following matters related to the applicant’s organization and administration

should be reviewed: 

(a) for familiarity, the general organization and administration methods used by the

applicant (see SRP Chapter 2)

(b) the areas of review listed in SRP Section 2.3 as they relate to NCS, including the

experience, educational requirements, responsibilities, and authorities of NCS

management and staff 

5.3.3 Management Measures 

The reviewer(s) should determine whether the applicant has committed to the facility

management measures in 10 CFR 70.62 and whether the commitments demonstrate the

applicant’s ability to implement and maintain the NCS program. The following matters related

to the applicant’s management measures should be reviewed in the application:

 

(1) For familiarity, the general configuration management, maintenance, training and

qualifications, procedures, audits and assessments, incident investigations, records

management, and other quality assurance elements used by the applicant (see SRP

Sections 11.3.1 through 11.3.8); and 

(2) Management provisions for the following:

(a) training and qualifications of NCS management and staff 

(b) auditing, assessing, and upgrading the NCS program 

(c) revising the ISA Summary as it relates to NCS 
Nuclear Criticality Safety 5-3 NUREG-1520

(d) recommending modifications to operating and maintenance procedures, to reduce the

likelihood of occurrence of an inadvertent nuclear criticality

(e) designing and installing a criticality accident alarm system (CAAS) to provide

immediate detection and annunciation of an inadvertent nuclear criticality 

(f) referring to the corrective action function any unacceptable performance deficiencies

that might result in an inadvertent nuclear criticality 

(g) referring to the corrective action function any unacceptable performance deficiencies

that did result in an inadvertent nuclear criticality

(h) retaining records of the NCS program, including documentation of corrective actions

taken 

5.3.4 Methodologies and Technical Practices 

The reviewer(s) should determine whether the applicant has identified NCS methodologies and

NCS technical practices used to make NCS determinations, as required by 10 CFR 70.24,

“Criticality Accident Requirements” (for criticality accident alarm systems), 10 CFR 70.61(d),

“Performance Requirements” (which require that under normal and credible abnormal

conditions, all nuclear processes remain subcritical, and maintain an approved margin of

subcriticality for safety), 10 CFR 70.64(a), “Requirements for New Facilities or New Processes

at Existing Facilities” (baseline design criteria), and 10 CFR 70.65(b), “Additional Content of

Applications”, (the ISA Summary). The following matters related to the applicant’s NCS

methodologies and NCS technical practices should be reviewed in the application: 

(1) the commitment to use the NCS methodologies identified in the applicant’s NCS program

(2) the commitment to use the NCS technical practices identified in the applicant’s NCS

program 

(3) the commitment to fulfill the requirements of 10 CFR 70.24 and to have a CAAS that has

been incorporated into the facility management measures 

(4) the commitment to detect an inadvertent nuclear criticality and promptly notify personnel,

which should ensure that the radiation exposure to workers will be minimized

(5) the commitment to the requirements of 10 CFR 70.61(d) 

(6) the commitment to the requirements in 10 CFR 70.64 as they relate to NCS 

(7) the areas of review listed in SRP Section 3.3 as they relate to NCS, including accident

sequences, consequences, likelihoods and IROFS 

(8) identification and use of appropriate NCS methodologies and NCS technical practices 
Nuclear Criticality Safety 5-4 NUREG-1520

5.4 ACCEPTANCE CRITERIA 

The applicant should provide NCS commitments and describe how the commitments will be

met. Commitments and descriptions are expected only when the acceptance criteria are

relevant to the operations and materials to be licensed. 

The applicant's use of standards should be considered acceptable if the applicant has met the

following acceptance criteria: 

If an applicant intends to conduct activities to which a standard applies and the standard has

been endorsed by an NRC Regulatory Guide, then a commitment to comply with all the

requirements (i.e., “shalls”) of the standard is necessary but may not be sufficient to meet the

acceptance criteria. Notwithstanding a general commitment to a standard, the applicant should

clarify its intended compliance with those requirements in the standard that are expressed only

as general principles by more specific commitments and descriptions in the application. Any

variations from the requirements of the standard should be identified and justified in the

application. The commitments and descriptions should be considered acceptable if the

applicant has met the acceptance criteria described below. 

5.4.1 Regulatory Requirements 

The regulatory basis for the review should be the general and “additional content” of an

application, as required by 10 CFR 70.22 and 70.65, respectively. In addition, the NCS review

should verify compliance with 10 CFR 70.24, 70.61, 70.62, 70.64, 70.72, and Appendix A of 10

CFR Part 70. 

5.4.2 Regulatory Guidance 

NRC Regulatory Guide 3.71, “Nuclear Criticality Safety Standards for Fuels and Materials

Facilities,” August 1998, endorses the American National Standards Institute (ANSI)/American

Nuclear Society (ANS)-8 national standards listed below in part or in full. 

• ANSI/ANS-8.1, “Nuclear Criticality Safety in Operations with Fissionable Materials Outside

Reactors,” 1983 (Reaffirmed in 1988). 

• ANSI/ANS-8.3, “Criticality Accident Alarm System,” 1997. 

• ANSI/ANS-8.5, “Use of Borosilicate-Glass Raschig Rings as a Neutron Absorber in Solutions

of Fissile Material,” 1996. 

• ANSI/ANS-8.6, “Safety in Conducting Subcritical Neutron-Multiplication Measurements In

Situ,” 1983 (Reaffirmed in 1995). 

• ANSI/ANS-8.7, “Guide for Nuclear Criticality Safety in the Storage of Fissile Materials,” 1975

(Reaffirmed in 1987). 

• ANSI/ANS-8.9, “Nuclear Criticality Safety Criteria for Steel-Pipe Intersections Containing

Aqueous Solutions of Fissile Materials,” 1987 (Reaffirmed in 1995). 

• ANSI/ANS-8.10, “Criteria for Nuclear Criticality Safety Controls in Operations With Shielding

and Confinement,” 1983 (Reaffirmed in 1988). 
Nuclear Criticality Safety 5-5 NUREG-1520

• ANSI/ANS-8.12, “Nuclear Criticality Control and Safety of Plutonium-Uranium Fuel Mixtures

Outside Reactors,” 1987 (Reaffirmed in 1993). 

• ANSI/ANS-8.15, “Nuclear Criticality Control of Special Actinide Elements,” 1981 (Reaffirmed

in 1995). 

• ANSI/ANS-8.17, “Criticality Safety Criteria for the Handling, Storage, and Transportation of

LWR Fuel Outside Reactors,” 1984 (Reaffirmed in 1997). 

• ANSI/ANS-8.19, “Administrative Practices for Nuclear Criticality Safety,” 1996. 

• ANSI/ANS-8.20, “Nuclear Criticality Safety Training,” 1991. 

• ANSI/ANS-8.21, “Use of Fixed Neutron Absorbers in Nuclear Facilities Outside Reactors,”

1995. 

• ANSI/ANS-8.22, “Nuclear Criticality Safety Based on Limiting and Controlling Moderators,”

1997. 

• ANSI/ANS-8.23, “Nuclear Criticality Accident Emergency Planning and 

Response,” 1997. 

NRC endorsement of these standards means that they provide procedures and methodology

generally acceptable to NRC staff for the prevention and mitigation of nuclear criticality

accidents. However, application of a standard is not a substitute for detailed nuclear criticality

safety analyses for specific operations. Applicants should generally use the most current

revision of the aforementioned standards. If an applicant commits to an unendorsed standard,

then the applicant needs to demonstrate in the application why the unendorsed standard should

be acceptable to NRC. 

5.4.3 Regulatory Acceptance Criteria 

5.4.3.1 Management of the NCS Program 

The applicant's management of the NCS program should be considered acceptable if the

applicant has met the following acceptance criteria or has identified and justified an alternative

in the application: 

(1) The applicant commits to develop, implement, and maintain an NCS program to meet the

regulatory requirements of 10 CFR Part 70.

(2) The applicant states the NCS program objectives, which should include those objectives

listed in SRP Section 5.3.1.

(3) The applicant establishes NCS safety parameters and procedures.

(4) The applicant outlines an NCS program structure and defines the responsibilities and

authorities of key program personnel.
Nuclear Criticality Safety 5-6 NUREG-1520

(5) The applicant commits to keep NCS methodologies and NCS technical practices applicable

to current configuration by means of the configuration management function.

(6) The applicant commits to use the NCS program to establish and maintain NCS safety limits

and NCS operating limits for IROFS in nuclear processes and commits to maintain

adequate management measures to ensure the availability and reliability of the IROFS.

(7) The applicant commits to preparation of NCS postings, to NCS training, and to NCS

emergency procedure training.

(8) The applicant commits to adhere to the NCS baseline design criteria requirements in 

10 CFR 70.64(a) for new facilities and new processes at existing facilities that require a

license amendment under 10 CFR 70.72.

(9) The applicant commits to use the NCS program to evaluate modifications to operations, to

recommend process parameter changes necessary to maintain the safe operation of the

facility, and to select appropriate IROFS and management measures. 

5.4.3.2 Organization and Administration 

Information related to NCS organization and administration acceptance criteria may be located

in the organization and administration part of the application (SRP Chapter 2). The applicant's

NCS organization and administration should be considered acceptable if the applicant has met

the following acceptance criteria or has identified and justified an alternative in the application: 

(1) The applicant meets the acceptance criteria in SRP Section 2.4 as they relate to NCS,

including organizational positions, functional responsibilities, experience, and qualifications

of personnel responsible for NCS.

(2) The applicant commits to ANSI/ANS-8.1-1983 and ANSI/ANS-8.19-1996, as they relate to

organization and administration.

(3) The applicant commits to the intent of Section 4.11 of ANSI/ANS-8.1-1983, which is to use

personnel, skilled in the interpretation of data pertinent to NCS and familiar with the

operation of the facility, as a resource in NCS management decisions. These specialists

should be independent of operations supervision.

(4) The applicant commits to provide NCS postings in areas, operations, work stations, and

storage locations.

(5) The applicant commits to the following policy: Personnel shall report defective NCS

conditions to the NCS function and perform actions only in accordance with written,

approved procedures. Unless a specific procedure deals with the situation, personnel shall

report defective NCS conditions to the NCS function and take no action until the NCS

function has evaluated the situation and provided recovery procedures.

(6) The applicant commits to describe organizational positions, experience of personnel,

qualifications of personnel, and functional responsibilities, and commits also to outline

organizational relations among the individual positions. 
Nuclear Criticality Safety 5-7 NUREG-1520

(7) The applicant commits to designate an NCS program director who will be responsible for

implementation of the NCS program.

(8) The applicant commits to staff the NCS program with suitably trained personnel and to

provide sufficient resources for its operation. 

5.4.3.3 Management Measures 

Information related to NCS management measures acceptance criteria may be located in the

management measures part of the application. The applicant's NCS management measures

(required by 10 CFR 70.62) should be considered acceptable if the applicant has met the

following acceptance criteria or has identified and justified an alternative in the application: 

(1) Training (see SRP Section 11.4.3.3) 

(a) The applicant commits to ANSI/ANS-8.19-1996 and ANSI/ANS-8.20-1991 as they

relate to training.

(b) The applicant commits to provide training to all personnel to recognize the CAAS

signal and to evacuate promptly to a safe area. 

(c) The applicant commits to provide instruction training regarding the policy in SRP

Section 5.4.3.2(5). 

(2) Procedures (see SRP Section 11.4.3.4) 

(a) The applicant commits to ANSI/ANS-8.19-1996 as it relates to procedures and to the

policy that no single, inadvertent departure from a procedure could cause an

inadvertent nuclear criticality. 

(3) Audits and assessments (see SRP Section 11.4.3.5) 

(a) The applicant commits to ANSI/ANS-8.19-1996 as it relates to audits 

and assessments.

(b) The applicant commits to conduct and document weekly NCS walkthroughs (e.g.,

using checklists) of all operating SNM process areas such that all operating SNM

process areas will be reviewed at least every 2 weeks. Identified weaknesses should

be referred to the facility corrective action function and should be promptly and

effectively resolved. A graded approach may be used to justify an alternate NCS

walkthrough schedule if it is based on the ISA and included in the ISA Summary.

(c) The applicant commits to conduct and document quarterly NCS audits such that all

NCS aspects of management measures (see SRP Chapter 11) will be audited at least

every 2 years. A graded approach may be used to justify an alternate NCS audit

schedule if it is based on the ISA and included in the ISA Summary. 
Nuclear Criticality Safety 5-8 NUREG-1520

5.4.3.4 Methodologies and Technical Practices 

The applicant's methodologies and technical practices should be considered acceptable if the

applicant has met the following acceptance criteria or has identified and justified an alternative

in the application: 

(1) NCS controlled parameters will be appropriately applied.

(2) NCS limits on IROFS will be appropriately determined. 

5.4.3.4.1 Methodologies 

The applicant's commitment to NCS methodologies should be considered acceptable if the

applicant has met the following acceptance criteria or has identified and justified an alternative

in the application: 

(1) NCS determinations will be performed using acceptable methodologies. 

(2) NCS limits on controls and controlled parameters will be established to ensure an

adequate margin of subcriticality for safety. 

(3) Methods used to develop NCS limits will be validated to ensure that they are within

acceptable ranges, and that the applicant utilized both appropriate assumptions, and

acceptable computer codes. 

(4) An inadvertent nuclear criticality will be detected promptly to ensure that radiation

exposures to workers are minimized.

(5) The applicant commits to ANSI/ANS-8.1-1983 as it relates to methodologies. 

(6) The applicant commits to the intent of the validation report statement in NRC Regulatory

Guide 3.71, August 1998, which states that the applicant should demonstrate: (1) the

adequacy of the margin of safety for subcriticality by assuring that the margin is large

compared to the uncertainty in the calculated value of k-eff, (2) that the calculation of k-eff

is based on a set of variables whose values lie in a range for which the methodology used

to determine k-eff has been validated, and (3) that trends in the bias support the extension

of the methodology to areas outside the area or areas of applicability.

(7) The applicant includes a reference to (including the date and revision number), and

summary description of, either a manual or a documented, reviewed, and approved

validation report (by NCS function and management) for each methodology that will be

used to make an NCS determination (e.g., experimental data, reference books, hand

calculations, deterministic computer codes, probabilistic computer codes, consensus

standards). When there are changes to either a reference manual or validation report, the

change should be reported to NRC by letter. After reviewing the change notification letter,

NRC will determine if a review of a reference manual or validation report is necessary. The

summary description of a reference manual or validation report should have: 

(a) A summary of the theory of the methodology that is sufficiently detailed and clear to

allow understanding of the methodology.
Nuclear Criticality Safety 5-9 NUREG-1520

(b) A summary of the area or areas to which the reference manual or validation report

applies.

(c) A commitment to apply the methodology only in the area or areas of applicability or

provide justifications for applying the methodology outside the area or areas of

applicability.

(d) A commitment to use pertinent computer codes, assumptions, and techniques in the

methodology.

(e) A commitment to properly perform the mathematical operations in the methodology.

(f) A commitment to use data based upon reliable and reproducible experimental

measurements.

(g) A commitment to use plant-specific benchmark experiments and data derived

therefrom to validate the methodology.

(h) A commitment to determine the bias, the uncertainty in the bias, the uncertainty in the

methodology, the uncertainty in the data, the uncertainty in the benchmark

experiments, and the margin of subcriticality for safety, when using the methodology.

(i) A commitment to use controlled software and hardware, when using the methodology.

(j) A commitment to use a verification process when using the methodology. 

(8) The applicant commits to have, at the facility, the reference manual or documented,

reviewed, and approved validation report (by NCS function and management) for each

methodology used to make an NCS determination. The manual or validation report should

have:

(a) A description of the theory of the methodology that is sufficiently detailed and clear to

allow understanding of the methodology and independent duplication of results.

(b) A description of the area or areas of applicability that identifies the range of values for

which valid results have been obtained for the parameters used in the methodology. 

In accordance with the provisions in ANSI/ANS-8.1-1983, any extrapolation beyond

the area or areas of applicability should be supported by an established mathematical

methodology.

(c) A description of the use of pertinent computer codes, assumptions, and techniques in

the methodology.

(d) A description of the proper functioning of the mathematical operations in the

methodology (e.g., a description of mathematical testing). 

(e) A description of the data used in the methodology, showing that the data were based

on reliable experimental measurements.

(f) A description of the plant-specific benchmark experiments and the data derived

therefrom that were used for validating the methodology.
Nuclear Criticality Safety 5-10 NUREG-1520

(g) A description of the bias, uncertainty in the bias, uncertainty in the methodology,

uncertainty in the data, uncertainty in the benchmark experiments, and margin of

subcriticality for safety, as well as the basis for these items, as they are used in the

methodology. If the bias is determined to be advantageous to the applicant, the

applicant shall use a bias of 0.0 (e.g., in a critical experiment where the k-eff is known

to be 1.00 and the code calculates 1.02, the applicant cannot use a bias of 0.02 to

allow calculations to be made above the value of 1.00). 

(h) A description of the software and hardware that will use the methodology. 

(i) A description of the verification process and results. 

(9) The applicant commits to incorporate each reference manual or documented, reviewed,

and approved validation report (by NCS function and management) for a methodology, as

well as the assumptions used, into the facility configuration management program. 

(10) The applicant commits to performing NCS determinations in accordance with specified

methods incorporated in the facility’s management measures and in accordance with

the following principles. 

(a) NCS safety limits, NCS operating limits, and limits on NCS controlled parameters will

be established assuming credible optimum conditions (i.e., most reactive conditions

physically possible or limited by written commitments to regulatory agencies) unless

specified controls are implemented to control the limit to a certain range of values.

(b) NCS safety limits, NCS operating limits, and limits on NCS controlled parameters will

be derived from the NCS determinations.

(c) NCS safety limits, NCS operating limits, and limits on NCS controlled parameters will

be based on the proper application of the NCS methodology to the process under

study.

(d) NCS operating limits will be derived from NCS safety limits by taking into consideration

changes in operating parameters to ensure processes will remain subcritical under

both normal and credible abnormal conditions.

(e) NCS operating limits will establish sufficient margins of safety for processes and take

into consideration the variability and uncertainty in processes and the NCS subcritical

limits.

(f) NCS safety limits will establish sufficient margins of safety for processes and take into

consideration the variability and uncertainty in processes and the NCS operating limits.

The margin of subcriticality for safety for a process should be relative compared to the

calculated value of k-eff. 

(g) K-eff is calculated from a set of variables whose values lie in a range for which the

validity of the NCS methodology has been demonstrated. 
Nuclear Criticality Safety 5-11 NUREG-1520

5.4.3.4.2 Technical Practices 

Controlled parameters available for NCS control include the following: mass, geometry,

density, enrichment, reflection, moderation, concentration, interaction, neutron absorption, and

volume. The applicant's commitment to NCS technical practices should be considered

acceptable if the applicant has met the following acceptance criteria or has identified and

justified an alternative in the application: 

(1) The applicant’s use of a single NCS control to maintain the values of two or more

controlled parameters constitutes only one component necessary to meet double

contingency protection. 

(2) Based on 10 CFR 70.61, the applicant commits to the policy that no single credible event

or failure can result in a criticality accident.

(3) The applicant commits to the preferred use of passive engineered controls to ensure NCS. 

In general, the applicant should commit to the following order of preference, for NCS

controls: (a) passive engineered; (b) active engineered; (c) augmented administrative; and

(d) simple administrative. When using a control, the choice of the type and manner should

be justified. 

(4) When evaluating a controlled parameter, the applicant should consider heterogeneous

effects. Heterogeneous effects are particularly relevant for low-enriched uranium

processes, where, all other parameters being equal, heterogeneous systems are more

reactive than homogeneous systems. 

(5) The applicant commits to perform an evaluation for all controlled parameters that shows

that during both normal and credible abnormal conditions, the controlled parameter will be

maintained. 

(6) When controlled parameters are controlled by measurement, reliable methods and

instruments should be used. Where there is significant susceptibility to human error, the

applicant may commit to representative sampling, reliable measurement instruments and

methods, and dual independent measurements. 

(7) The use of mass as a controlled parameter should be considered acceptable if: 

(a) When a given mass of material has been determined, a percentage factor is used to

determine the mass percentage of SNM in that material. 

(b) When fixed geometric devices are used to limit the mass of SNM, a conservative

process density is used. 

(c) When the mass is measured, instrumentation is used. 

(d) When using double-batching of SNM as a single parameter limit control from

experimental data, and double-batching of SNM is possible, the mass of SNM is

limited to no more than 45 percent of the minimum critical mass, based on spherical

geometry. 
Nuclear Criticality Safety 5-12 NUREG-1520

(e) When using double-batching of SNM as a single parameter limit control from

experimental data and double-batching of SNM is not possible, the mass of SNM is

limited to no more than 75 percent of the critical mass, based on spherical geometry.

(8) The use of geometry as a controlled parameter should be considered acceptable if: 

(a) Before beginning operations, all dimensions and nuclear properties that use geometry

control are verified. The facility configuration management program should be used to

maintain these dimensions and nuclear properties. 

(b) When using large single units as a single parameter control from experimental data,

the margins of safety are 90 percent of the minimum critical cylinder diameter, 85

percent of the minimum critical slab thickness, and 75 percent of the minimum critical

sphere volume. 

(9) The use of density as a controlled parameter should be considered acceptable if: 

(a) When process variables can affect the density, the process variables are shown in the

ISA Summary to be controlled by IROFS. 

(b) When the density is measured, the measurement is obtained by the use of

instrumentation. 

(10) The use of enrichment as a controlled parameter should be considered acceptable if: 

(a) A method of segregating enrichments is used to ensure differing enrichments will not

be interchanged, or else the most limiting enrichment is applied to all material. 

(b) When the enrichment needs to be measured, the measurement is obtained by using

instrumentation. 

(11) The use of reflection as a controlled parameter should be considered acceptable if: 

(a) When investigating an individual unit, the wall thickness of the unit and all reflecting

adjacent materials of the unit are considered. The adjacent materials should be

farther than 30.48 cm (12 inches) away from the unit. 

(b) After identifying potential reflectors, the controls to prevent the presence of the

potential reflectors are identified as IROFS in the ISA Summary. 

(12) The use of moderation (e.g., exclusion of moderators) as a controlled parameter

should be considered acceptable if: 

(a) When using moderation, the applicant commits to ANSI/ANS-8.22-1997. 

(b) When process variables can affect the moderation, the process variables are shown in

the ISA Summary to be controlled by IROFS. 

(c) When the moderation is measured, the measurement is obtained by using

instrumentation. 

(d) When designing physical structures, the design precludes the ingress of moderation.
Nuclear Criticality Safety 5-13 NUREG-1520

(e) When moderation is needed to be sampled, dual independent sampling methods are

used. 

(f) When developing firefighting procedures for use in a moderation controlled area,

restrictions are placed on the use of moderator material.

(g) After evaluating all credible sources of moderation for the potential for intrusion into a

moderation-controlled area, the ingress of moderation is precluded or controlled. 

(13) The use of concentration as a controlled parameter should be considered acceptable

if: 

(a) When process variables can affect the concentration, the process variables are shown

in the ISA Summary to be controlled by IROFS. 

(b) High concentrations of SNM in a process are precluded unless the process is

analyzed to be safe at any credible concentration. 

(c) When using a tank containing-concentration controlled solution, the tank is normally

closed.

(d) When concentration needs to be sampled, dual independent sampling methods are

used.

(e) After identifying possible precipitating agents, precautions are taken to ensure that

such agents will not be inadvertently introduced. 

(14) The use of interaction as a controlled parameter should be considered acceptable if: 

(a) When maintaining a physical separation between units, engineered controls to ensure

a minimum spacing or augmented administrative controls are used. The structural

integrity of the spacers or racks should be sufficient for normal and credible abnormal

conditions. 

(15) The use of neutron absorption as a controlled parameter should be considered

acceptable if: 

(a) When using borosilicate-glass raschig rings, the applicant commits to 

ANSI/ANS-8.5-1996. 

(b) When using fixed neutron absorbers, the applicant commits to ANSI/ANS-8.21-1995. 

(c) When evaluating absorber effectiveness, neutron spectra are considered (e.g.,

cadmium is an effective absorber for thermal neutrons, but ineffective for fast

neutrons). 

(16) The use of volume as a controlled parameter should be considered acceptable if: 

(a) When using volume control, fixed geometry is used to restrict the volume of SNM with

engineered devices to limit the accumulation of SNM. 

(b) When the volume is measured, instrumentation is used. 
Nuclear Criticality Safety 5-14 NUREG-1520

5.4.3.4.3 Requirements in 10 CFR 70.24 

The applicant's commitment to the CAAS requirements in 10 CFR 70.24 should be considered

acceptable if the applicant has met the following acceptance criteria or has identified and

justified an alternative in the application: 

(1) The applicant documents that the facility CAAS meets the requirements of 10 CFR 70.24. 

(2) The applicant commits to ANSI/ANS-8.3-1997, as modified by Regulatory Guide 3.71,

August 1998. 

(a) At or above the 10 CFR 70.24 mass limits, CAAS coverage should be required in each

area in which SNM is handled, stored, or used. 

(b) A requirement of 10 CFR 70.24 is that each area that needs CAAS coverage be

covered by two detectors. 

(c) A requirement of 10 CFR 70.24 is that a CAAS be capable of detecting a nuclear

criticality that produces an absorbed dose in soft tissue of 20 rads of combined

neutron and gamma radiation at an unshielded distance of 2 meters from the reacting

material within 1 minute. 

(3) The applicant commits to having a CAAS that is uniform throughout the facility for the type

of radiation detected, the mode of detection, the alarm signal, and the system

dependability. 

(4) The applicant commits to having a CAAS that is designed to remain operational during

credible events such as a seismic shock equivalent to the site-specific design-basis

earthquake or the equivalent value specified by the Uniform Building Code. 

(5) The applicant commits to having a CAAS that is designed to remain operational during

credible events such as a fire, an explosion, a corrosive atmosphere, and other credible

conditions. 

(6) The applicant commits to having a CAAS alarm that is clearly audible in areas that must be

evacuated or provides alternate notification methods that are documented to be effective in

notifying personnel that evacuation is necessary. 

(7) The applicant commits to rendering operations safe, by shutdown and quarantine if

necessary, in any area where CAAS coverage has been lost and not restored within a

specified number of hours. The number of hours should be determined on a process-byprocess basis, because shutting down certain processes, even to make them safe, may

carry a larger risk than being without a CAAS for a short time. The applicant should

commit to compensatory measures (e.g., limit access, halt SNM movement) when the

CAAS system is not functional. 

(8) Emergency management (see SRP Chapter 8) 

(a) The applicant commits to the requirements in ANSI/ANS-8.23-1997 as they relate to

NCS.
Nuclear Criticality Safety 5-15 NUREG-1520

(b) The applicant either has an emergency plan or satisfies the alternate requirements in

10 CFR 70.22.(h)(1)(i). 

(c) The applicant commits to provide fixed and personnel accident dosimeters in areas

that require a CAAS. These dosimeters should be readily available to personnel

responding to an emergency and there should be a method for prompt onsite

dosimeter readouts. 

(d) The applicant commits to provide emergency power for the CAAS or provide

justification for the use of continuous monitoring with portable instruments. 

5.4.3.4.4 Requirements in 10 CFR 70.61(d) 

The applicant's commitment to the requirements in 10 CFR 70.61(d) that under normal and

credible abnormal conditions, all nuclear processes are subcritical, including use of an

approved margin of subcriticality for safety should be considered acceptable if the applicant has

met the following acceptance criteria or has identified and justified an alternative in the

application: 

(1) The applicant commits to the use of NCS controls and controlled parameters to assure that

under normal and credible abnormal conditions, all nuclear processes are subcritical,

including use of an approved margin of subcriticality for safety. 

(2) As stated in ANSI/ANS-8.1-1983, the applicant commits to the following policy: Process

specifications shall incorporate margins to protect against uncertainties in process

variables and against a limit being accidentally exceeded. 

(3) The applicant commits to the following national standards, as they relate to these

requirements: ANSI/ANS-8.7-1975, ANSI/ANS-8.9-1987, ANSI/ANS-8.10-1983,

ANSI/ANS-8.12-1987, ANSI/ANS-8.15-1981, and ANSI/ANS-8.17-1984. 

(4) If the applicant intends to use administrative k-eff margins for normal and credible

abnormal conditions, the applicant commits to NRC pre-approval of the administrative

margins. 

(5) The applicant commits to determining subcritical limits for k-eff calculations such that: ksubcritical = 1.0 - bias margin, where the margin includes adequate allowance for

uncertainty in the methodology, data, and bias to assure subcriticality. 

(6) The applicant commits to performing studies to correlate the change in a value of a

controlled parameter and its k-eff value. The studies should also include changing the

value of one controlled parameter and determining its effect on another controlled

parameter and k-eff. 

(7) The applicant commits to implement an NCS program that ensures double contingency

protection, when practicable. When evaluating double contingency protection, the

following should be considered with respect to the contents of both ANSI/ANS-8.1-1983

and the likelihood discussion in SRP Chapter 3:
Nuclear Criticality Safety 5-16 NUREG-1520

(a) Adherence to double contingency protection: Each process that has accident

sequences that could result in an inadvertent nuclear criticality should have double

contingency protection. Double contingency protection may be provided by either: 

(i) at least two-parameter control (the control of at least two independent process

parameters) or (ii) single-parameter control (a system of multiple independent controls

on a single process parameter). The first method is the preferred approach because

of the difficulty of preventing common-mode failure when controlling only one

parameter. 

(b) As used in double contingency protection, the term “concurrent” means that the effect

of the first process change persists until a second change occurs, at which point the

process could have an inadvertent nuclear criticality. It does not mean that the two

events initiating the change must occur simultaneously. The possibility of an

inadvertent nuclear criticality can be markedly reduced if failures of NCS controls are

rapidly detected and the processes rendered safe. If not, processes can remain

vulnerable to a second failure for extended periods of time.

(c) Exceptions to double contingency protection: There may be processes where double

contingency protection is not practicable. In those processes, the facility should

implement sufficient redundancy and diversity in controlled parameters such that at

least two unlikely and concurrent events, errors, accidents, or equipment malfunctions

are necessary before an inadvertent nuclear criticality is possible. The applicant

should commit in the license application, to identify in the ISA Summary any process

that could lead to an inadvertent nuclear criticality and for which double contingency

was not applied. The applicant must provide adequate justification for this decision in

the ISA Summary.

(8) The applicant meets the acceptance criteria in SRP Section 3.4, as they relate to

subcriticality of operations and margin of subcriticality for safety. 

Note: This is the acceptance criterion for reviewing the application and evaluating the

high-risk accident sequences and a selected sampling of other than high-risk

accident sequences. 

5.4.3.4.5 Requirements in 10 CFR 70.64(a) 

The applicant's commitment to the baseline design criteria requirements in 10 CFR 70.64(a)

should be considered acceptable if the applicant has met the following acceptance criteria or

has identified and justified an alternative in the application: 

(1) The applicant commits to the double contingency principle in determining NCS controls

and IROFS in the design of new facilities or new processes at existing facilities that

require a license amendment under 10 CFR 70.72.

(2) The applicant commits to double contingency protection as discussed in SRP Section

5.4.3.4.4(9). 

5.4.3.4.6 Requirements in 10 CFR 70.65(b) (ISA Summary) 

The applicant is required to meet the performance requirements in 10 CFR 70.61(b) and (c), as

well as the performance requirements in 10 CFR 70.61(d), which include the requirement to
Nuclear Criticality Safety 5-17 NUREG-1520

limit the risk of an inadvertent nuclear criticality by assuring that all nuclear processes remain

subcritical. The applicant’s evaluation of NCS accident sequences should be performed in a

manner consistent with the applicant’s evaluation of non-NCS accident sequences used to

meet 10 CFR 70.61(b) and (c); however 10 CFR 70.61(d) requires the applicant to use

prevention methods as the primary means to meet the performance requirements of 

10 CFR 70.61(b) and (c). 

The applicant's commitment to the requirements in 10 CFR 70.65(b) should be considered

acceptable if the applicant has met the following acceptance criteria or has identified and

justified an alternative in the application: 

(1) The applicant meets the acceptance criteria in SRP Section 3.4, as they relate to the

following: identification of NCS accident sequences, consequences of NCS accident

sequences, likelihoods of NCS accident sequences, and descriptions of IROFS for NCS

accident sequences. 

(2) The applicant commits to use Appendix A of ANSI/ANS-8.1-1983 in determining NCS

accident sequences. 

(3) The applicant commits to ANSI/ANS-8.10-1983, as modified by Regulatory Guide 3.71,

August 1998, in determining the consequences of NCS accident sequences. 

5.4.3.4.7 Additional NCS Program Commitments 

The applicant’s description of additional commitments regarding the NCS program should be

considered acceptable if the applicant has met the following acceptance criteria or has

identified and justified an alternative in the application: 

(1) The applicant commits to use the NCS program to promptly detect any NCS deficiencies

by means of operational inspections, audits, or investigations, and to refer to the

facility’s corrective action function any unacceptable performance deficiencies in IROFS,

NCS function, or management measures, so as to prevent recurrence. 

(2) The applicant commits to support the facility change mechanism process by performing

NCS determinations to evaluate changes to processes, operating procedures, IROFS,

and management measures. 

(3) The applicant commits to upgrade the NCS program to reflect changes in the ISA or

new NCS methodologies and to modify operating and maintenance procedures in ways

that could reduce the likelihood of occurrence of an inadvertent nuclear criticality. 

(4) The applicant commits to retain records of NCS programs and to document any

corrective actions taken. 

(5) The applicant commits to use the NCS methodologies and technical practices in SRP

Section 5.4.3.4 to evaluate NCS accident sequences in operations and processes. 

(6) The applicant’s description of measures to implement the facility change process

requirements in 10 CFR 70.72 should be considered acceptable if the applicant has met

the following acceptance criteria or has identified and justified an alternative in 

the application: 
Nuclear Criticality Safety 5-18 NUREG-1520

(a) The applicant commits to a change control process that is sufficient to ensure that the

safety basis of the facility will be maintained during the lifetime of the facility. The

change process should be documented in written procedures and should ensure that

all potentially affected SNM processes are evaluated to determine the effect of the

change on the safety basis of the process, including the effect on bounding process

assumptions, on the reliability and availability of NCS controls, and on the NCS of

connected processes. The change control process should have procedures for the

review and approval of facility changes by the NCS function to determine the potential

effects on NCS.

(b) The change control process should be connected to the facility’s configuration

management system to ensure that changes to the NCS basis are incorporated into

procedures, evaluations, postings, drawings, other safety basis documentation, and

the ISA Summary. 

(c) The applicant commits to a program to determine whether facility changes require

NRC approval in accordance with the 10 CFR 70.72(c). This program should be

documented in written procedures and must involve individuals qualified to determine

the incremental effect of changes to the safety basis as documented in the ISA

Summary. All proposed changes should be compared to the approved ISA Summary.

(7) The applicant’s description of measures to implement the reporting requirements in 

10 CFR 70 Appendix A should be considered acceptable if the applicant has met the

following acceptance criteria or has identified and justified an alternative in 

the application: 

(a) The applicant has a program for evaluating the criticality significance of NCS events

and an apparatus in place for making the required notification to the NRC Operations

Center. The determination of significance of NCS events should be made by qualified

individuals. The determination of loss or degradation of double contingency protection

should be made against the license and 10 CFR 70 Appendix A.

(b) The applicant incorporates the reporting criteria of Appendix A and the report content

requirements of 10 CFR 70.50 into the facility emergency procedures. 

(c) The applicant commits to issue the necessary report based on whether the IROFS

credited were lost, irrespective of whether the safety limits of the associated

parameters were actually exceeded. 

(d) The applicant commits to the following: If the licensee cannot ascertain within one

hour of whether the criteria of 10 CFR 70 Appendix A Paragraph (a) or (b) apply, the

event should be treated as a one-hour reportable event. 

5.5 REVIEW PROCEDURES 

The reviewer should use the regulatory guidance of this chapter; references in this chapter; and

the applicant’s reports to the NRC (e.g., NRC Bulletin 91-01, 10 CFR 70.50, and 70.74).
Nuclear Criticality Safety 5-19 NUREG-1520

5.5.1 Acceptance Review 

The primary reviewer should review the applicant's NCS information for completeness with

respect to the requirements in 10 CFR 70.22, 70.24, 70.61, 70.62, 70.64, and 70.65, and the

acceptance criteria in Section 5.4. If deficiencies are identified, then either the applicant should

be requested to submit additional material before the start of the safety evaluation or the

application should be denied. 

5.5.2 Safety Evaluation 

When an acceptable application is received from the applicant, the primary reviewer should

conduct a complete review of the application and determine its acceptability. The primary

reviewer should consult with the supporting reviewers to identify and resolve any issues of

concern related to the licensing review. The primary reviewer should also coordinate with other

primary reviewers of SRP Chapters 2, 3, 8 and 11 to confirm that all acceptance criteria

pertinent to NCS have been met. 

The primary reviewer for Chapter 5 should determine whether the acceptance criteria in SRP

Section 5.4 have been met and should prepare the SER NCS chapter in accordance with SRP

Section 5.6.

5.6 EVALUATION FINDINGS 

Note: The evaluation finding for the ISA Summary requirements for 10 CFR 70.65 should be in

SRP Section 3.6.

If the staff’s review verifies that sufficient information has been provided in the safety program

description to satisfy the acceptance criteria in SRP Section 5.4, the staff should document its

review as follows: 

The staff has reviewed the Nuclear Criticality Safety (NCS) program for [name of facility]

according to SRP Chapter 5. The staff has reasonable assurance that: 

(1) The applicant will have in place a staff of managers, supervisors, engineers, process

operators, and other support personnel who are qualified to develop, implement, and

maintain the NCS program in accordance with the facility organization and

administration and management measures.

(2) The applicant's conduct of operations will be based on NCS methodologies and NCS

technical practices, which will ensure that the fissile material will be possessed, stored,

and used safely according to the requirements in 10 CFR Part 70.

(3) The applicant will develop, implement, and maintain a criticality accident alarm system in

accordance with both the requirements in 10 CFR 70.24 and the facility emergency

management program. 

(4) The applicant will have in place an NCS program in accordance with the subcriticality of

operations and margin of subcriticality for safety requirements in 10 CFR 70.61(d) and

baseline design criteria requirements in 10 CFR 70.64(a). 
Nuclear Criticality Safety 5-20 NUREG-1520

(5) Based on this review, the staff concludes that the applicant’s NCS program meets the

requirements of 10 CFR Part 70 and provides reasonable assurance for the protection

of public health and safety, including workers and the environment. 

5.7 REFERENCES 

H. K. Clark,, "Maximum Safe Limits for Slightly Enriched Uranium and Uranium Oxide," Du Pont

de Nemours and Co. DP-1014, Aiken, SC, 1966. 

R. A. Knief, “Nuclear Criticality Safety --Theory and Practice,” American Nuclear Society, La

Grange Park, IL, 1985. 

H. C. Paxton and N. L. Pruvost, LA-10860-MS, “Critical Dimensions of Systems Containing

235U, 239Pu, and 233U,” Los Alamos National Laboratory, Los Alamos, NM, 1987. 

N. L. Pruvost and H. C. Paxton, LA-12808/UC-714, “Nuclear Criticality Safety Guide,” Los

Alamos National Laboratory, Los Alamos, NM, 1996. 

W. R. Stratton (D. R. Smith Revisor), DOE/NCT-04, “A Review of Criticality Accidents,” U.S.

Department of Energy, March 1989. 

U.S. Code of Federal Regulations, Title 10, Part 70, ‘Domestic Licensing of Special Nuclear

Material,’ U.S. Government Printing Office, Washington, DC. 

U.S. Department of Energy, DOE Order 420.1 (Change 2), “Facility Safety,” October 24, 1996.

Accession #: ML013370337
Chemical Process Safety 6-1 NUREG-1520

6 CHEMICAL PROCESS SAFETY

6.1 PURPOSE OF REVIEW

The primary purpose of the review is to determine with reasonable assurance that the applicant

has designed a facility that will provide adequate protection against chemical hazards related to

the storage, handling, and processing of licensed materials. The facility design must

adequately protect the health and safety of workers and the public during normal operations

and credible accident conditions from the chemical risks of licensed material and from

hazardous chemicals produced from licensed material. It must also protect against facility

conditions that could affect the safety of licensed materials and thus present an increased

radiation risk (e.g., release of a chemical that could incapacitate operators and preclude their

entry to an area of the facility where licensed materials are handled).

Chemical safety issues are initially evaluated as part of the applicant’s ISA Summary. The ISA

Summary must evaluate credible accident sequences at the facility; identify IROFS to prevent

the occurrence or to mitigate the consequences of accidents; and include the management

measures that provide reasonable assurance of the availability and reliability of IROFS when

needed. Before assessing the applicant’s facility design to protect against chemical hazards,

the reviewer should first review the license application, “Facility and Process Description” (SRP

Section 1.1), and the ISA Summary (Chapter 3), to gain familiarity with: 

• process information and accident sequences leading to conditions that could pose

chemical hazards

• specific IROFS, to prevent or mitigate such chemical hazards and

• recommended management measures for ensuring that the IROFS will be available and

reliable when required

6.2 RESPONSIBILITY FOR REVIEW

Primary: Chemical Process Safety Reviewer (all sections of this

chapter)

Supporting: Licensing Project Manager 

Fuel Cycle Facility Inspection Staff (as needed)

Health Physicist (for 10 CFR Part 20 uranium and

transuranic toxicity issues)

Primary Reviewers of Chapters 1, 3, 8 and 11 of this SRP

6.3 AREAS OF REVIEW

An applicant is required by 10 CFR 70.62(a) to establish and maintain a safety program that will

adequately protect worker, public health and safety, and the environment from the chemical

hazards from licensed material. The applicant is not necessarily required to establish a

separate chemical process safety program, but the applicant must demonstrate that chemical

hazards and accident sequences that affect licensed material be considered and adequately

prevented or mitigated. Applicants are required to conduct an ISA and provide an ISA

Summary that meets the requirements of 10 CFR 70.65.
Chemical Process Safety 6-2 NUREG-1520

The staff’s chemical process safety review should focus on the chemical safety-related accident

sequences described in the ISA Summary (SRP Chapter 3) and the corresponding

management measures (SRP Chapter 11) to confirm that the applicant’s equipment, facilities,

and management measures are adequate to protect against releases and chemical exposures

of licensed material, hazardous chemicals produced from licensed material, and chemical risks

of plant conditions that affect the safety of licensed material. The review must verify that any

grading of IROFS or management measures proposed by the applicant in accordance with 

10 CFR 70.62(a) is commensurate with the accident risk that the IROFS are designed to

reduce.

The Memorandum of Understanding (MOU) between the NRC and the Occupational Safety and

Health Administration (OSHA) directs the NRC to oversee chemical safety issues related to 

(a) radiation risks of licensed materials, (b) chemical risks of licensed materials, and (c) plant

conditions that affect or may affect the safety of licensed materials and thus increase radiation

risk to workers, the public, and the environment. The NRC does not oversee plant conditions

that absolutely do not affect or involve the safety of licensed materials.

The staff’s review should cover the following specifications:

(1) the chemical process description, (the narrative description of the site, facility, and

processes with respect to chemical safety for normal operations) including process

chemistry, flow diagrams, major process steps, and major pieces of equipment

(2) chemical accident sequences, including unmitigated accident sequences involving

hazardous chemicals and licensed materials and interpretation of the quantitative

chemical risk levels

(3) chemical accident consequences, identified in the ISA Summary, including the

applicant’s interpretation of the qualitative chemical risk levels and the assumptions,

bases, and methods the applicant used to forecast the consequences to workers and

the public of accidents that involve hazardous chemicals and licensed materials

(4) chemical process IROFS, including a list of the adequacy of items relied on for chemical

safety and justification of their adequacy 

(5) chemical process management measures, including management measures to assure

the reliability and availability of IROFS (chemical process safety)

(6) safety grading, including, if applicable, grading of IROFS and their associated

management measures

(7) the coordination of chemical process safety and emergency management (to be verified

by contacting the reviewer of chapters)

(8) the applicant’s commitment to retain records for chemical process safety compliance

and reporting commitments for chemical releases

(9) the applicant’s commitment to adhere to the 10 CFR 70.64 chemical baseline design

criteria for new facilities or for new processes at an existing facility requiring a license

amendment under 10 CFR 70.72 (as applicable)
Chemical Process Safety 6-3 NUREG-1520

(10) the applicant’s commitment to refer any unacceptable performance deficiency to the

facility’s corrective action function

6.4 ACCEPTANCE CRITERIA

6.4.1 Regulatory Requirements

The regulatory basis for the review should be the general and additional contents of an

application for chemical process safety, as required by 10 CFR 70.22, and 70.65. In addition,

the chemical process safety review should be conducted to provide reasonable assurance of

compliance with 10 CFR 70.61, 70.62, and 10 CFR 70.64, for new facilities or new processes,

at an existing facility, requiring a license amendment under 10 CFR 70.72.

6.4.2 Regulatory Guidance

The following regulatory guidance is relevant to chemical process safety:

• NUREG-1513, “Integrated Safety Analysis Guidance Document,” 2001.

• NUREG-1601, “Chemical Process Safety at Fuel Cycle Facilities,” 1997.

• NUREG/CR-6410, “Nuclear Fuel Cycle Facility Accident Analysis Handbook,” 1998.

6.4.3 Regulatory Acceptance Criteria

The reviewer should find the applicant’s chemical process safety information acceptable if it

provides reasonable assurance that the following acceptance criteria are adequately addressed

and satisfied. The applicant may elect to incorporate some or all the requested chemical

process information in Facility and Process Description (SRP Section 1.1), or the ISA

Summary, rather than in this section. Either approach is acceptable, so long as the information

is adequately cross-referenced.

The applicant should also describe commitments to maintain chemical process safety records,

and describe applicable commitments to “audits and assessments” and “incident investigation”

for detecting and correcting of any unacceptable performance deficiencies in accordance with

the SRP chapters on “Management Measures” (Chapter 11).

6.4.3.1 Process Chemical Risk and Accident Sequences

The applicant’s descriptions of facility processes and chemical accident sequences are

acceptable if:

(1) Process descriptions are sufficiently detailed to allow an understanding of the chemical

process hazards (including radiological hazards caused by, or involving chemical

accidents) and to allow development of potential accident sequences.

(2) The applicant provides an adequate list of the consequences and likelihoods of accident

sequences identified in the ISA Summary as involving hazardous chemicals produced

from licensed material and chemical risks of plant conditions that affect the safety of

licensed materials. Each accident sequence should include a chemical hazard
Chemical Process Safety 6-4 NUREG-1520

evaluation of potential interactions of process chemicals with confinement vessels,

process equipment, and facility personnel. The hazard evaluation should use

appropriate accepted methods. The applicant provides reasonable assurance that

measures to mitigate the consequences of accident sequences identified in the ISA

Summary are consistent with actions described in “Emergency Management” (SRP

Chapter 8).

(3) The applicant identifies and uses appropriate techniques and valid assumptions in

estimating the concentrations or in predicting the “toxic” footprint for releases of

hazardous chemicals produced from licensed material or by abnormal plant conditions

that could affect the safety of licensed materials. The applicant uses the performance

requirements criteria as described in 10 CFR 70.61(b) and 70.61(c).

(4) Source term and vapor dispersion models used to calculate the concentration of

uranium hexafluoride (UF6) and its reaction products conform to the guidance on the

applicability of models in NUREG/CR-6481, “Review of Models Used for Determining

Consequences of UF6 Release.”

(5) If dispersion models are used to determine whether a release of chemicals might affect

worker or public health and safety, the applicant provides evidence that the models used

are appropriate to the application and that the assumed input data lead to a

conservative estimate of potential consequences. Consequence analyses conform to

the guidance on atmospheric and consequence modeling in NUREG/CR-6410, “Nuclear

Fuel Cycle Facility Accident Analysis Handbook,” 1998.

(6) The applicant has proposed appropriate chemical exposure standards to assess

chemical consequences. Acceptable exposure standards include, but are not limited to,

The Emergency Response Planning Guidelines (ERPGs) established by the American

Industrial Hygiene Association, the Acute Exposure Guideline Levels (AEGLs)

established by the National Advisory Committee for Acute Guideline Levels for

Hazardous Substances, the exposure limits established by OSHA, and the exposure

limits contained in International Organization for Standardization (ISO) standards. If the

applicant does not use a published exposure standard or knows of no exposure

standard for a chemical, the applicant may propose an alternative exposure standard

accompanied by supporting documentation to justify the alternative. Note: 10 CFR

70.61, “Performance Requirements,” are for “acute chemical exposures,” and OSHA

permissible exposure limits (PELs) are typically time-weighted average (TWA) values. 

Consequently, for ISA purposes, acute chemical release limits may not be adjusted by

the TWA calculation (which involves concentration and duration of exposure) unless a

rational basis is provided in the ISA Summary.

6.4.3.2 IROFS and Management Measures

The license application should identify the design basis for chemical process safety for normal

operation and describe features such as materials of construction, sizing, system fabrication,

and process control schemes. Based on a comparison of the unmitigated chemical

consequences determined in Section 6.4.3.1, with the performance criteria of 10 CFR 70.61,

the applicant should identify (in the ISA Summary) chemical process safety controls (i.e.,

IROFS) suitable to prevent or mitigate potential accidents. IROFS also should be identified for

those accident sequences containing a chemical system or process failure that may ultimately
Chemical Process Safety 6-5 NUREG-1520

lead to radiological consequences that exceed the performance requirements. If the applicant

takes a graded approach to safety in accordance with 10 CFR 70.62(a), the reviewer should

establish that the grading of IROFS is appropriate and sufficient to protect against chemical

process risks. For example, the applicant should consider reliance on passive controls of active

systems and defense-in-depth in accordance with 10 CFR 70.64(b). To reduce common mode

failures, the applicant should favor design features that use independent sources of motive

force for items such as control actuators, jet pumps, eductors, and ejectors. Fail-safe controls

are preferred unless safety concerns preclude this approach.

The applicant must also review management measures to assure the availability and reliability

of such IROFS, when they are required to perform their safety functions. Management

measures may be graded commensurate with risk.

The application must contain other information:

(1) the application should describe the engineering approach, basis, or schemes employed

for maintaining safety in normal operations.

(2) the ISA Summary must identify the administrative and engineered controls to prevent or

mitigate a chemical process risk, the hazard being mitigated, and the risk category. The

applicant should also explain how any safety grading of IROFS and management

measures has been made and how such grading is commensurate with the reduction in

risk that the IROFS are designed to achieve.

(3) the application should demonstrate the management measures proposed to assure

IROFS are available and reliable when required by briefly describing: 

(a) its procedures to ensure the reliable operation of engineered controls (e.g.,

inspection and testing procedures and frequencies, calibration programs,

functional tests, corrective and preventive maintenance programs, criteria for

acceptable test results).

(b) its procedures to ensure that administrative controls will be correctly implemented,

when required (e.g., employee training and qualification in operating procedures,

refresher training, safe work practices, development of standard operating

procedures, training program evaluation).

6.4.3.3 Requirements for New Facilities or New Processes at Existing Facilities

The application should address the baseline design criteria (BDC) for new facilities or new

processes at existing facilities that require a license amendment under 10 CFR 70.72. 

NUREG-1601, Section 2.4, “Design Basis,” contains a list of items that should be considered in

an adequate facility design. With respect to chemical process safety, the application should be

considered acceptable if it includes the information listed below (or references other sections of

the application):

(1) the applicant briefly describes how the ISA was performed for the new process, and

from the ISA satisfies principles to the BDC and the performance requirements in 10

CFR 70.61, and applies defense-in-depth to higher-risk accident sequences. 

Acceptable principles for defense-in-depth of the chemical process safety design are
Chemical Process Safety 6-6 NUREG-1520

those that support a hierarchy of controls: prevention, mitigation, and operator

intervention in order of preference. For example, limiting inventory of onsite chemicals

is preventive and is the preferred a preferential, preventive practice for avoiding

chemical safety-related accidents

(2) the applicant describes proposed facility-specific or process-specific relaxations or

additions to BDC, along with justifications for relaxations

(3) the ISA Summary contains a description of how the chemical safety BDC were applied

in establishing the design principles, features, and control systems of the new process

6.5 PROCEDURES FOR REVIEW

6.5.1 Acceptance Review 

The primary reviewer should evaluate the application to determine whether it addresses the

topics in Section 6.3, “Areas of Review.” If significant deficiencies are identified, the applicant

should be requested to submit additional material before the start of the safety evaluation.

6.5.2 Safety Evaluation

After determining that the application is acceptable for review in accordance with Section 6.5.1,

the primary reviewer will perform a safety evaluation against the “Acceptance Criteria”

described in Section 6.4. If, during the course of the safety evaluation, the primary reviewer

determines a need for additional information, the primary reviewer coordinates a request for

additional information with the licensing project manager. The reviewer should ascertain that

the chemical process safety approach is consistent with other sections of the application,

including the “ISA Summary” (SRP Chapter 3), “Radiation Safety” (SRP Chapter 4) “Emergency

Management” (SRP Chapter 8), and “Management Measures” (SRP Chapter 11). For

example, the reviewer should determine that the chemical safety program will not have

unacceptable or adverse impacts on the radiological safety at the facility.

For an existing facility the reviewer may consult cognizant NRC inspectors to identify and

resolve any issues related to the licensing review. For a planned facility the reviewers may wish

to consult with the facility design team to gain a better understanding of the process, its

potential hazards, and safety approaches.

The primary reviewer will prepare a safety evaluation report (SER) for the Licensing Project

Manager in support of the licensing action.

6.5.2.1 Process Chemical Risks and Accident Sequences

The results of the ISA are the basis for the chemical process safety evaluation. The reviewer

should review the chemical risks identified in the ISA Summary and ensure that the level of

safety is reflected in the design and the operational plans for the facility. The reviewer should

establish that the applicant’s facility design, operations, and IROFS for to chemical safety

provide reasonable assurance that they will function as intended and provide for the safe

handling of licensed material at the facility. The reviewer should review the mechanisms that

will allow the applicant to identify and correct potential problems. 
Chemical Process Safety 6-7 NUREG-1520

In return, to validate the criteria used by the applicant in reporting sequences in the ISA

Summary, the reviewer will make an independent judgment of the comparative risks assigned

by the applicant to accident sequences identified in the ISA Summary. The judgement is based

on risk relative to other sequences (competing risks), the complexity of the sequence, facility

operating history, and general industry performance. The review may cover a selected number

of lower risk chemical safety-related accident sequences not identified in the ISA Summary,

6.5.2.2 IROFS and Management Measures

The staff reviews the chemical process safety IROFS to ensure their adequacy in protecting

against all unmitigated sequences identified in the ISA Summary.

If the applicant has applied a graded approach to safety, the reviewer should establish that the

grading of IROFS or management measures, is appropriate and sufficient to protect against

chemical process risks.

6.5.2.3 Requirements for New Facilities or New Processes at Existing Facilities

The staff reviews the applicant’s commitments to adhere to the BDC, in 10 CFR 70.64(a), for

the design of new facilities or new processes at an existing facility that require a license

amendment under 10 CFR 70.72.

6.6 EVALUATION FINDINGS

The reviewer writes an SER addressing each topic reviewed and explains why the NRC staff

has reasonable assurance that the chemical safety part of the application is acceptable. 

License conditions may be proposed to impose requirements where the application is deficient. 

In cases where the SER is drafted in advance of resolving all outstanding chemical process

safety issues, the reviewer documents the review as described below and includes a list of

open issues that require resolution before the staff finding of reasonable assurance. For partial

reviews, revisions, and process changes, the reviewer uses applicable sections of the

acceptance criteria, and the SER notes areas that were not reviewed and the chemical process

safety significance, if any. On completion of the review, NRC staff may impose temporary

license conditions to authorize short-duration activities. For certain functions and requirements

that concern safety or regulatory issues, a license condition may be imposed and remain in

effect until removed by an amendment or license renewal.

The SER should include a summary statement of what was evaluated and the basis for the

reviewer’s conclusions. The SER should include statements like the following:

The staff has evaluated the application using the criteria listed previously. Based on the

review of the license application, the NRC staff has concluded that the applicant has

adequately described and assessed accident consequences that could result from the

handling, storage, or processing of licensed materials and that could have potentially

significant chemical consequences and effects. The applicant has constructed hazard

analysis that identified and evaluated those chemical process hazards and potential

accidents and established safety controls to provide reasonable assurance of safe facility

operation. To ensure that the performance requirements in 10 CFR Part 70 are met, the

applicant has provided reasonable assurance that controls are maintained available and

reliable when required to perform their safety functions. The staff has reviewed these
Chemical Process Safety 6-8 NUREG-1520

safety controls and the applicant's plan for managing chemical process safety and finds

them acceptable. 

The staff concludes that the applicant's plan for managing chemical process safety and

the chemical process safety controls meet the requirements of 

10 CFR Part 70, and provides reasonable assurance that the health and safety of the

public will be protected. 

6.7 REFERENCES

Center for Chemical Process Safety, “Guidelines for the Technical Management of Chemical

Process Safety,” American Institute of Chemical Engineers, New York, 1989, Chapter 11, as

revised.

Chemical Manufacturers Association, “Responsible Care®, Process Safety Code of

Management Practices,” Washington, 1990.

U.S. Code of Federal Regulations, Title 10, Part 70, “Domestic Licensing of Special Nuclear

Material,” U.S. Government Printing Office, Washington, D.C., as revised.

U.S. Code of Federal Regulations, Title 29, Part 1910.119, “Process Safety Management of

Highly Hazardous Chemicals,” U.S. Government Printing Office, Washington, D.C., as revised.

U.S. Nuclear Regulatory Commission, Manual Chapter 2603, “Inspection of the Nuclear

Chemical Process Safety Program at Fuel Cycle Facilities,” as revised.

U.S. Nuclear Regulatory Commission/Occupational Safety and Health Administration,

“Memorandum of Understanding between the Nuclear Regulatory Commission and the

Occupational Safety and Health Administration, ‘Worker Protection at NRC-Licensed Facilities,’”

Federal Register No. 53, October 31, 1988.

U.S. Nuclear Regulatory Commission, “Chemical Process Safety at Fuel Cycle Facilities,”

NUREG-1601, 1997.

U.S. Nuclear Regulatory Commission, “Integrated Safety Analysis Guidance Document,”

NUREG-1513, 2001.

U.S. Nuclear Regulatory Commission, “Nuclear Fuel Cycle Facility Accident Analysis

Handbook,” NUREG/CR-6410, 1998.

U.S. Nuclear Regulatory Commission, “Review of Models Used for Determining Consequences

of UF6 Release,” NUREG/CR-6481, as revised.

Accession #: ML013370346
Fire Safety 7-1 NUREG-1520

7 FIRE SAFETY

7.1 PURPOSE OF REVIEW

The purpose of this review is to determine with reasonable assurance that the applicant has

designed a facility that provides adequate protection against fires and explosions that could

affect the safety of licensed materials and thus present an increased radiological risk. The

review should also establish that the application has considered radiological consequences of

the fires and will institute suitable safety controls to protect workers, the public, and the

environment. 

7.2 RESPONSIBILITY FOR REVIEW

Primary: Fire Safety Specialist

Secondary: Criticality Safety Specialist

Environmental Specialist 

Chemical Safety Specialist

Physical Security Specialist

Supporting: Regional, Resident, and Fuel Cycle Inspection Staff

7.3 AREAS OF REVIEW

The regulation established in 10 CFR 70.62(a) requires an applicant to develop, implement, and

maintain a safety program that will reasonably protect health and safety of the public and the

environment from the fire and explosive hazards associated with processing, handling, and

storing licensed materials during normal operations, anticipated operational occurrences, and

credible accidents. The reviewers should first consult the ISA Summary (SRP Chapter 3) to

identify analyzed operations that have the potential to result in fire or explosion, and should gain

familiarity with the IROFS and complementary management measures that are proposed to

prevent or mitigate any resulting chemical or radiological risks. The fire protection program

must address these process-specific risks, as well as general fire prevention, protection, and

management issues. Although 10 CFR Part 70 does not require a separate fire safety

program, an applicant should provide commitments pertaining to fire safety in the following

areas:

! Fire safety management includes safety organization, engineering review, fire prevention;

inspection, testing, and maintenance; pre-fire plans; and personnel qualifications, drills,

and training.

! Fire risk identification includes the Fire Hazards Analysis (FHA) and the ISA Summary.

! Facility design includes information on building construction, fire areas, life safety,

ventilation, and electrical system design. The facility design should also consider

competing requirements among fire safety and security, criticality, and environmental

concerns.

! Process fire safety involves design considerations to prevent an accident or to mitigate the

consequences of an accident resulting from using process chemicals, combustible metals,
Fire Safety 7-2 NUREG-1520

flammable and combustible liquids and gases, high-temperature equipment, hot cells and

glove boxes, and laboratories.

! Fire protection systems include fire detection, alarm, and suppression systems; portable

extinguishers; water supplies; and emergency response organizations.

7.4 ACCEPTANCE CRITERIA

An applicant that meets the acceptance criteria defined in this section, or has provided an

acceptable alternative, should be considered to have an acceptable fire safety program.

7.4.1 Regulatory Requirements

The regulatory basis for the fire safety review should be the general and additional contents of

application as required by 10 CFR 70.22 and 70.65. In addition, the fire safety review should

focus on to providing reasonable assurance of compliance with 10 CFR 70.61, 70.62, and

70.64.

7.4.2 Regulatory Guidance

The relevant regulatory guidance for fire safety includes the following NRC and industrial

standards:

• National Fire Protection Association, “Standards for Facilities Handling Radioactive Material,”

NFPA Standard 801, latest edition. 

• U.S. Nuclear Regulatory Commission, “Integrated Safety Analysis Guidance Document,”

NUREG-1513, 2001.

• U.S. Nuclear Regulatory Commission, ”Nuclear Fuel Cycle Facility Accident Analysis

Handbook,” NUREG/CR-6410, 1998.

7.4.3 Regulatory Acceptance Criteria

Partial acceptability of the application and the ISA Summary will be contingent on the NRC

staff’s review of the applicant’s commitments to control and mitigate fire hazards. The staff will

focus on whether the application is risk-informed, addresses the applicant’s procedures for 

maintaining an acceptable level of fire safety, and demonstrates that the applicant is prepared

to react quickly and safely to extinguish fires. An applicant may use a graded approach to

define fire safety, but must provide sufficient documentation and commitments to ensure that 

workers, the public, and the environment will be adequately protected from fire events.

These acceptance criteria may be incorporated in the information supplied to satisfy SRP

Section 3 (ISA) or other SRP sections so long as clear references are provided (information

need not be repeated). The staff’s fire safety specialist will review the application, ISA

Summary, and other documentation, as needed, regarding these acceptance criteria. 

The reviewer(s) will use nationally recognized codes and standards are used by the reviewer to

measure reasonable assurance of fire safety. These include, but are not limited to, the NFPA

National Fire Codes, Factory Mutual (FM) Data Sheets and Approval Guide, Underwriters

Laboratories (UL) Standards and Building Material Directory, ANSI Standards; and ASTM
Fire Safety 7-3 NUREG-1520

Standards. Commitments to specified standards will normally be considered an acceptable

means of meeting the acceptance criteria. 

The NRC staff will review the application regarding the following acceptance criteria: 

7.4.3.1 Fire Safety Management Measures

An adequate application documents how the applicant will administer and ensure fire safety at

the licensed facility. The application should reflect a commitment to ensure that the IROFS, as

identified in the ISA Summary, are available and reliable, and the facility maintains fire safety

awareness among employees, controls transient ignition sources and combustibles, and

maintains a readiness to extinguish or limit the consequences of fire. These measures are

unique to fire safety and, therefore are not included in the acceptance criteria for SRP 

Section 11, “Management Measures.”

An adequate application identifies a senior-level manager who has the authority and staff to

ensure that fire safety receives appropriate priority. A facility safety committee or fire safety

review committee staffed by managers of different disciplines should integrate facility

modifications. (The facility safety committee can do the work of a fire safety review committee.) 

As described in the application, day-to-day fire safety should be supervised by an individual with

sufficient practical fire safety experience in nuclear facilities. 

The “Standard for Fire Protection for Facilities Handling Radioactive Materials,” NFPA 801,

specifies the following fire safety management measures, including fire prevention; inspection,

testing, and maintenance of fire protection systems; emergency response organization

qualifications, drills, and training; and pre-fire plans. An adequate application documents the

fire safety management measures in sufficient detail to identify their relationship to, and

functions for, normal operations; anticipated (off-normal) events; and accident safety (i.e.,

IROFS). The staff recognizes NFPA 801 as one acceptable standard for fire safety

management measures, however the applicant may use other nationally recognized codes and

standards if appropriate. The staff’s fire safety specialist will review the adequacy of the fire

safety management measures presented in the application.

7.4.3.2 Fire Hazards Analysis

Knowing the fire risk allows an applicant to apply the appropriate level of fire protection to

ensure the safety of workers, the public and the environment from fire-induced radiological

hazards. To be risk-informed, a licensee should conduct an FHA for each facility, or part

thereof, that, if totally consumed by fire, could release SNM in quantity and form that could

cause at least an intermediate consequence, as defined in 10 CFR 70.61. The FHA should

develop bounding credible fire scenarios for each fire area containing significant fire loading,

and then assess the consequences of an unmitigated fire. The staff recognizes NFPA 801 as

one standard that provides guidance for conducting FHAs, however, the applicant may use

other nationally recognized codes and standards if appropriate. The FHA should include a

description, by fire area, of the fuel loading, fire scenarios, methods of consequence analysis,

the potential consequences, and a description of the mitigative controls. 

The FHA is used to identify possible fire initiators and accident sequences leading to

radiological consequences or toxic chemical consequences resulting from interaction with SNM. 

In developing accident sequences that will be reported in the ISA Summary, the ISA team will

consider the FHA results and assign likelihoods of the various events in the accident

sequences. With respect to fire safety, the ISA Summary is acceptable if the credible fire
Fire Safety 7-4 NUREG-1520

hazards (e.g., from the FHA) are identified for each process fire are, and information is

provided to detail how each fire hazard was considered and addressed (i.e., the management

measures and/or IROFS) for each process accident sequence whose consequence could

exceed the performance requirements in 10 CFR 70.61. Thus, the FHA is a fundamental tool

for evaluating fire hazards as input to the ISA evaluation. The staff’s fire safety specialist will

review the adequacy fire safety aspects of the ISA Summary. 

7.4.3.3 Facility Design

Building construction, fire area determination, electrical installation, life safety, ventilation,

drainage, and lightning protection are all facility design features that affect fire safety. An

adequate application documents the fire safety considerations used in the general design of the

facilities containing licensed material or facilities that impose an exposure threat to radiological

facilities. The staff recognizes NFPA 801 as one standard that specifies acceptable facility fire

safety design criteria; however the applicant may use other nationally recognized codes and

standards if appropriate. The staff’s fire safety specialist will review the adequacy facility’s fire

safety design.

The following are other specific areas of concern:

• Criticality concerns may exclude water extinguishing systems from process areas. 

However, during major fire events, the fire may easily overcome the extinguishing

capability of portable extinguishers, and hose lines may be needed to extinguish the fire. 

Consequently, applicants should consider using total flooding gaseous systems in waterexclusion areas with significant fire risks. An adequate application addresses the

methodology for extinguishing fires in water-exclusion areas. The staff’s fire safety and

criticality specialist will review for adequacy.

• Environmental concerns include the potential for thousands of gallons of fire water to be

contaminated with nuclear material during a fire event. Consequently, diked areas and

drainage of process facilities may be needed. NFPA 801 provides guidance on how to

calculate the potential amount of runoff to properly size drainage and containment

systems. An adequate application documents any measures used to control fire water

runoff. The staff’s fire safety and environmental specialists will review the applicant’s fire

water retention system for adequacy.

• Physical Security Concerns include the need to design buildings and facilities to provide

safe egress in case of fire or chemical events that could lead to radiological emergencies. 

Physical security requirements for SNM may inadvertently delay worker egress and fire

fighter access. Physical security procedures should allow offsite fire departments quick

and efficient access to fire emergencies. An adequate application documents the design

criteria used for worker egress and procedures for firefighter access. The staff recognizes

NFPA 801 as one standard that specifies acceptable worker egress design criteria;

however, the applicant may use other nationally recognized codes and standards if

appropriate. The staff’s fire safety and physical security specialists will review for

adequacy.

• Design and construction of new facilities should: (1) comply with the baseline design

criteria (BDC) specified in 10 CFR 70.64(a), (2) comply with the defense-in-depth

requirements of 10 CFR 70.64(b), and (3) be consistent with the guidance provided in

NFPA 801 or other appropriate nationally recognized fire protection codes and standards. 
Fire Safety 7-5 NUREG-1520

The staff’s fire safety specialist will review the adequacy of the fire safety design of new

facilities.

7.4.3.4 Process Fire Safety

Many hazardous chemicals and processes used by fuel cycle facilities contribute to the fire

hazards affecting radiological areas. In areas that have fire hazards that may threaten licensed

material, the application should identify the hazardous chemicals, processes, and design

standards used to ensure fire safety. The staff recognizes NFPA 801 as one standard that

provides acceptable design criteria for radiological process areas that may contain: hazardous

material, laboratories, high-temperature equipment, hot cells, and/or glove boxes. However,

the applicant may use other nationally recognized codes and standards if appropriate. The

staff’s fire safety and chemical safety specialists will review the adequacy of application. 

The following are a few of the more common hazardous materials used at fuel cycle facilities:

• Anhydrous ammonia is an explosive, flammable, and toxic gas used to make hydrogen.

• Fluorine reacts violently with organic material or metal powders and water vapor.

• Hydrogen is an explosive and flammable gas used in reduction processes.

• Hydrogen peroxide, off-gases hydrogen and oxygen, and is incompatible with some

extinguishers.

• Nitric acid, nitrates organic material, which lowers the ignition temperature of combustibles.

• Sulfuric acid absorbs water from organic material in an exothermic reaction, thereby causing

ignition.

• Zirconium is a combustible metal that burns at elevated temperatures.

7.4.3.5 Fire Protection and Emergency Response

The application should document the fire protection systems and fire emergency response

organizations provided for licensed facilities. The ISA Summary (see SRP Chapter 3) should

identify the fire protection IROFS. An adequate application describes the fire protection

provided for areas in which licensed material is present. The application should describe which

standards the fire protection systems and equipment meet. The staff recognizes the

NFPA’s national fire codes as acceptable standards for the design, installation, testing, and

maintenance of the fire protection systems and equipment. However, the applicant may use

other nationally recognized codes and standards if appropriate. 

Facilities with significant fire risks may need a fire emergency response team. One acceptable

standard is NFPA 600, “Industrial Fire Brigades.” However, the applicant may use other

nationally recognized codes and standards if appropriate. If offsite fire departments are needed

for facility fire safety, periodic training with the fire departments is necessary to enable them to

become familiar with facility access procedures, facility layout, and pre-fire plans. A

memorandum of understanding (MOU) between the applicant and the fire departments may be

necessary to define the required protection. The staff’s fire safety specialist will review the

adequacy of the applicant’s fire protection and emergency response commitments.

7.5 REVIEW PROCEDURES

7.5.1 Acceptance Review

During the acceptance review, the primary reviewer evaluates the application for completeness

as required by 10 CFR Part 70 and determines whether the application addresses the criteria
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discussed in SRP Section 7.3, “Areas of Review.” If significant deficiencies in the application

are identified, the primary reviewer should recommend the return of the application or

requested additional information before the start of the safety evaluation.

7.5.2 Safety Evaluation

During the safety evaluation, the primary and secondary reviewers evaluate whether the

application is adequate to comprehensively describe the fire safety of the licensed activity as

identified in Section 7.3, “Areas of Review,” and the commitments made in response to the

criteria specified in Section 7.4, “Acceptance Criteria.” The staff may request the applicant or

licensee to provide additional information or modify the submittal to meet the acceptance

criteria.

Reviewers should note that NFPA 801 uses “administrative control” in a different sense than 

10 CFR Part 70 and elsewhere in this SRP. In 10 CFR Part 70, an administrative control is an

IROFS if it is the human action necessary to meet safety performance requirements, and it is

supported by management measures (training, quality assurance, procedures, etc.) that ensure

that the action will be taken if needed. In NFPA 801, “administrative controls” are the training,

qualifications, procedures, etc., behind the human action; however these elements are referred

to as “Management Measures” in 10 CFR Part 70, and in this SRP.

7.6 EVALUATION FINDINGS

The staff’s review should verify that the applicant (1) provides sufficient information to satisfy

the intent of 10 CFR Part 70 requirements related to the overall safety program, and (2) is

consistent with the fire safety criteria in this SRP. On the basis of this information, the staff

should be able to evaluate whether the application meets the appropriate criteria. As an

example, the staff might document the fire safety review in an SER in the following manner:

The applicant has established a fire protection function meeting the acceptance criteria

in Chapter 7 of the “SRP for the Review of a License Application fro a Fuel Cycle

Facility.” The function includes a facility safety review committee responsible for

integrating modifications to the facility and a Fire Safety Manager responsible for day-today program implementation. Fire prevention, inspection, testing, and maintenance of

fire protection systems, and the qualification, drills, and training of facility personnel are

in accordance with applicable NFPA codes and standards. (Note that fire protection

training requirements are described in SER Section 11.3.) 

The applicant has conducted risk analyses in accordance with NFPA 801, “Standard for

Fire Protection for Facilities Handling Radioactive Material.” The FHAs identified

credible fire scenarios that bound the fire risk. The ISA used these scenarios and

identified fire protection IROFS, (in particular, wet pipe sprinkling in the process areas,

isolation of the high-temperature equipment within fire barriers, and a fire brigade

meeting NFPA 600, “Industrial Fire Brigades”). An MOU with the fire department

documents the required assistance and the annual exercises. Procedures are in place

to allow the fire department efficient access to process areas during fire emergencies. 

Worker egress is designed and maintained in accordance with NFPA 101, “Life Safety

Code.” 

The staff concludes that the applicant’s capabilities meet the criteria in Chapter 7 of the

SRP. The staff concludes that the applicant’s proposed equipment, facilities, and

procedures provide a reasonable level of assurance that adequate fire protection will be
Fire Safety 7-7 NUREG-1520

provided and maintained for those IROFS to meet the safety performance requirements

and BDC of 10 CFR Part 70.

7.7 REFERENCES

National Fire Protection Association, “National Fire Codes,” latest edition.

U.S. Nuclear Regulatory Commission, Information Notice No. 92-14, ”Uranium Oxide Fires at

Fuel Cycle Facilities,” February 21, 1992.

U.S. Nuclear Regulatory Commission, NUREG-1513, “Integrated Safety Analysis Guidance

Document,” 2001.

Accession #: ML013370359
Emergency Plan 8-1 NUREG-1520

8 EMERGENCY MANAGEMENT

8.1 PURPOSE OF REVIEW

The purpose of reviewing the applicant’s emergency management plan is to determine if the

applicant has established, before the start of operations, adequate emergency management

facilities and procedures to protect workers, the public, and the environment. In preparing its

emergency plan, the applicant may use either this SRP or Regulatory Guide 3.67, “Standard

Format and Content for Emergency Plans for Fuel Cycle and Materials Facilities,” dated

January 1992. Information requested for the emergency plan may be provided once and 

cross-referenced in other sections.

Licensed facilities requiring an emergency management plan include those that are authorized

to possess (1) enriched uranium or plutonium for which a criticality accident alarm system is

required, (2) uranium hexafluoride in excess of 50 kg (110 lb.) in a single container or 1,000 kg

(2200 lb.) total or (3) plutonium in excess of 2 Ci in unsealed form or on foils or plated sources. 

A licensed facility also requires an emergency management plan when an evaluation (or the

ISA Summary referenced in lieu of the evaluation) shows that the maximum dose to a member

of the public offsite from a release of radioactive materials would exceed 0.01 Sv (1 rem)

effective dose equivalent or an intake of 2 milligrams of soluble uranium.

Emergency capability is incorporated into the baseline design criteria (BDC) of 10 CFR Part 70,

as revised, and is intended to ensure control of licensed material, evacuation of personnel, and

availability of emergency facilities.

8.2 RESPONSIBILITY FOR REVIEW

Primary: Assigned Licensing Staff

Secondary: Licensing Project Manager

Supporting: Regional Emergency Preparedness Inspector

ISA Reviewer

Fuel Facility Inspection Staff

8.3 AREAS OF REVIEW

The NRC staff should review the applicant’s submittal for an acceptable level of evidence of

planning for emergency preparedness directed at situations involving real or potential

radiological hazards. In particular, the review should address those design features, facilities,

functions, and equipment that may affect some aspect of emergency planning or the capability

of an applicant to cope with facility emergencies. In addition, the review should address

coordination with offsite emergency response organizations. The staff should either review the

emergency plan in accordance with 10 CFR 70.22(i)(1)(ii) and the guidance contained in the

acceptance criteria below, or review the applicant’s evaluation (or the ISA Summary referenced

in lieu of the evaluation) that an emergency plan is not needed in accordance with 

10 CFR 70.22(i)(1)(i).

The NRC staff reviewer should address the areas of review, as described in Sections 8.3.1 and

8.3.2, below.
Emergency Plan 8-2 NUREG-1520

8.3.1 Emergency Plan

If the applicant submits an emergency plan, the staff should evaluate the emergency plan

against 10 CFR 70.22(i)(1)(ii) and Regulatory Guide 3.67, “Standard Format and Content for

Emergency Plans for Fuel Cycle and Materials Facilities,” which provides a standard format and

content for an emergency plan. Elements in the emergency plan to be reviewed should include

the following: 

(1) facility description (including both onsite and offsite emergency facilities)

(2) types of accidents

(3) classification of accidents

(4) detection of accidents

(5) mitigation of consequences (and safe shutdown)

(6) assessment of releases

(7) responsibilities of licensee

(8) notification and coordination

(9) information to be communicated and parties to be contacted

(10) training

(11) safe shutdown (recovery and facility restoration)

(12) exercises and drills

(13) hazardous chemicals inventories and locations

(14) responsibilities for developing and maintaining the emergency program and its procedures

8.3.2 Evaluation That No Emergency Plan is Required

If the applicant submits an evaluation or references the ISA Summary to demonstrate that an

emergency plan is not required, the staff should review the information against 

10 CFR 70.22(i)(1)(i), and NUREG-1140, “A Regulatory Analysis of Emergency Preparedness

for Fuel Cycle and Other Radioactive Material Licensees.” NUREG/CR-6410, “Nuclear Fuel

Cycle Facility Accident Analysis Handbook,” dated March 1998, also contains useful

information. Areas evaluated should include the following:

(1) a description of the facility

(2) types of materials used, including both radioactive material and hazardous chemicals

(3) types of accidents

(4) detection of accidents

(5) site specific information used to support the evaluation

(6) an evaluation of the consequences 

8.4 ACCEPTANCE CRITERIA

8.4.1 Regulatory Requirements

10 CFR 70.22(i)(1)(i) specifies when an applicant is not required to submit an emergency plan

to the NRC; if an applicant is required to submit an emergency plan, 10 CFR 70.22(i)(3)

contains the information that must be included in the emergency plan. In addition, 

10 CFR 70.64(a)(6) requires applicants to address the control of licensed material, evacuation

of personnel, and availability of emergency facilities for the design of new facilities.
1

 The map should include the location of sensitive facilities near the site, such as

hospitals, schools, nursing homes, nearest residents, fire department, prisons, environmental

sampling locations, and other structures and facilities that are important to emergency

management.

Emergency Plan 8-3 NUREG-1520

8.4.2 Regulatory Guidance

Regulatory guidance for preparing an emergency plan includes the following sources:

• U.S. Nuclear Regulatory Commission Regulatory Guide 3.67, "Standard Format and Content

for Emergency Plans for Fuel Cycle and Materials Facilities," January 1992.

• U.S. Nuclear Regulatory Commission NUREG-1140, “A Regulatory Analysis of Emergency

Preparedness for Fuel Cycle and Other Radioactive Materials,” January 1988.

• U.S. Nuclear Regulatory Commission NUREG/CR-6410, “Nuclear Fuel Cycle Facility Accident

Analysis Handbook,” 1998.

8.4.3 Regulatory Acceptance Criteria

8.4.3.1 Emergency Plan

The reviewer should evaluate the adequacy of the proposed emergency plan against the

requirements in 10 CFR 70.22(i)(3) and the specific acceptance criteria given in Sections

8.4.3.1.1 through 8.4.3.1.14 of this SRP. The applicant’s emergency plan should be acceptable

if the regulatory requirements and the following acceptance criteria are met.

8.4.3.1.1 Facility Description

The emergency plan should include a description of the facility and site, the area near the site,

and the licensed activities. These descriptions should include the following: 

(1) a detailed drawing of the site showing the following features:

(a) onsite and near offsite (within 1.61 km or 1 mile) structures, with building numbers and

labels

(b) roads and parking lots onsite and main roads near the site

(c) site boundaries, showing fences and gates

(d) major site features

(e) water bodies within approximately 1.61 km (1 mile)

(2) a general area map covering a radius of approximately 16.1 km (10 miles); a U.S.

Geological Survey topographical quadrangle (71⁄2 minute series; including the adjacent

quadrangle(s) if the site is located less than 1.61 km (1 mile) from the edge of the

quadrangle); and a map or aerial photograph indicating onsite and near-site structures

within a radius of approximately 1.61 km (1 mile)1

(3) stack heights, typical stack flow rates, and efficiencies of any emission-control devices
Emergency Plan 8-4 NUREG-1520

(4) a general description of licensed and other major activities conducted at the facility, and

the type, form, and quantities of radioactive and other hazardous materials that are

normally onsite, by location (use and storage) and building, and hazardous characteristics

(exposure rates, pH, temperature, and other characteristics) that are important to

emergency management

(5) certification by the Plant Manager (or the individual authorized by the applicant) that the

applicant has met all responsibilities under the Emergency Planning and Community Right

To Know Act of 1986, Title III, Public Law 99-499, in accordance with 

10 CFR 70.22(i)(3)(xiii).

8.4.3.1.2 Onsite and Offsite Emergency Facilities

The emergency plan should list and describe onsite and offsite facilities that could be relied on

in case of an emergency. The emergency plan should include the following:

(1) a list and description of both onsite and offsite emergency facilities, by location and

purpose

(2) a description of emergency monitoring equipment that is available for personnel and area

monitoring, as well as that for assessing the release to the environment of radioactive or

hazardous chemicals incident to the processing of licensed material

(3) a description of the onsite and offsite services that support emergency response

operations which should include the following:

(a) decontamination facilities

(b) medical treatment facilities

(c) first aid personnel

(d) fire fighters

(e) law enforcement assistance

(f) ambulance services 

(4) the applicant’s commitment to the following:

(a) facilities of adequate size and appropriate location that are designated, equipped, and

ready for emergency use

(b) adequate backup facilities required by the emergency plan and supporting documents

that are available and ready for use

(c) appropriate equipment and supplies necessary to support emergency response

activities that are accessible during accident conditions

(d) emergency equipment that is inventoried, tested, and serviced on a periodic basis to

ensure accountability and reliability

(e) sufficient reliable primary and backup communications channels that are available to

accommodate emergency needs
Emergency Plan 8-5 NUREG-1520

(f) offsite emergency resources and services that are identified and ready to ensure their

timely mobilization and use

(g) operational engineering information, such as current as-built drawings and

procedures, that are readily available in the emergency facilities

(h) sufficient equipment for personnel protection and monitoring

(i) systems in place to alert onsite and offsite personnel in case of an emergency

8.4.3.1.3 Types of Accidents

For each general type of accident identified in the ISA Summary, for which protective actions

may be needed, the emergency plan should describe the following:

(1) the process and physical location(s) where the accidents could occur

(2) complicating factors and possible onsite and offsite consequences, including releases of 

non-radioactive hazardous chemicals incident to the processing of licensed material that

could impact emergency response efforts 

(3) the accident sequence that has the potential for the greatest radiological and/or toxic 

chemical impact

(4) figure(s) projecting doses and toxic substance concentrations as a function of distance and

time for various meteorological stability classes, including a description of how such

doses/concentrations were projected (e.g., computer models, assumptions, etc.)

8.4.3.1.4 Classification of Accidents

The emergency plan should classify accidents as follows:

(1) The emergency plan classification system should include the following two classifications:

• Alert: Events that may occur, are in progress, or have occurred, which could lead to a

release of radioactive material or hazardous chemicals incident to the processing of

licensed material; however, the release is not expected to require a response by an

offsite response organization, to protect persons offsite.

• Site area emergency: Events that may occur, are in progress, or have occurred, which

could lead to a significant release of radioactive material or hazardous chemicals

incident to the processing of licensed material, that could require a response by offsite

emergency response organizations to protect persons offsite.

(2) The emergency plan should identify the classification (alert or site area emergency) that is

expected for each accident identified in the emergency plan.

(3) The emergency plan should specify emergency action levels (EALs) at which an alert or

site area emergency will be declared. EALs are specific conditions that require emergency

response measures to be performed. The applicant’s EALs should be consistent with
Emergency Plan 8-6 NUREG-1520

Appendix A to Regulatory Guide 3.67, and should be comparable to the Protective Action

Guides of the U.S. Environmental Protection Agency (EPA 400-R-92-001, May 1992). 

Transportation accidents more than 1.61Km (1 mile) from the facility should not be

classified.

(4) The emergency plan should designate the personnel positions and alternates with the

responsibility for accident classification during normal operations and back shifts.

8.4.3.1.5 Detection of Accidents

For each type of accident identified, the emergency plan should describe the following:

(1) the means of detecting the accident

(2) the means of detecting any release of radioactive material or hazardous chemicals incident

to the processing of licensed material

(3) the means of alerting the operating staff

(4) the anticipated response of the operating staff

8.4.3.1.6 Mitigation of Consequences

For each accident identified in the ISA Summary, the emergency plan should briefly describe

measures and equipment to be used for safe shutdown and mitigating the consequences to

workers onsite and offsite, as well as to the public offsite.

8.4.3.1.7 Assessment of Releases

(1) The emergency plan should describe the following aspects of the applicant’s procedures to

be used to promptly and effectively assess the release of radioactive material or hazardous

chemicals incident to the processing of licensed material:

(a) procedures for estimating or measuring the release rate or source term

(b) valid computer codes used to project doses or concentrations to the public or

environment and associated assumptions, along with adequate justifications to show

the validity of the assumptions

(c) types, methods, frequencies, implementation times, and other details of onsite and

offsite sampling and monitoring that will be performed to assess a release of

radioactive materials or hazardous chemicals incident to the processing of licensed

material

(d) method for assessing collateral damage to the facility (especially IROFS)

(2) The emergency plan should describe the applicant’s procedure for validating any code

used to assess releases of radioactive material or hazardous chemicals incident to the

processing of licensed material. 
Emergency Plan 8-7 NUREG-1520

8.4.3.1.8 Responsibilities

The emergency plan should describe the emergency response organization and administration

that ensure effective planning, implementation, and control of emergency preparedness

activities. In addition, the applicant should make the following commitments:

(1) The organizational structure and chain of command will be clearly defined in procedures.

(2) Staffing and resources will be sufficient to accomplish all assigned tasks.

(3) Responsibilities and authority for each management, supervisory, and professional position

will be clearly defined in procedures. Responsibility is assigned for the coordination of

onsite and offsite emergency response preparedness.

(4) Interfaces with supporting groups, both onsite and offsite, will be clearly defined in

procedures.

(5) Mutual cooperation agreements exist or will be entered into with local agencies, such as

fire, police, ambulance/rescue, and medical units.

(6) Plant management measures will be in place by procedures to audit and assess

emergency preparedness to ensure site readiness to handle emergencies and to identify

and correct problems.

(7) The onsite emergency response organization will provide effective command and control of

the site during the assessment, mitigation, and recovery phase of an accident.

(8) The emergency public information system will provide advance and ongoing information to

the media and public on subjects that would be discussed during an emergency, such as

radiation hazards, chemical hazards, site operation, and site emergency plans.

(9) The schedule of emergency preparedness procedure development will provide for

availability of procedures to support startup and operation of new processes/facilities

onsite.

8.4.3.1.9 Notification and Coordination

(1) The emergency plan should provide reasonable assurance that emergency notification

procedures will enable the emergency organization to correctly classify emergencies, notify

emergency response personnel, and initiate or recommend appropriate actions in a timely

manner, on the basis of the following:

(a) Emergency events are classified on the basis of the current emergency plan.

(b) Notification procedures minimize distraction of shift operating personnel and include

concise, preformatted messages. Appropriate followup messages to offsite authorities

are issued in a timely manner.

(c) Information on the nature and magnitude of the hazards is made available to

appropriate emergency response personnel.
Emergency Plan 8-8 NUREG-1520

(d) Radiological and chemical source term data are available to the command post,

technical support center, emergency operation center, and appropriate State

personnel, in cooperation with the NRC.

(e) When available, offsite field monitoring data are logged, compared with source term

data, and used in the protective action recommendation process.

(f) Protective Action Guides are available and used by appropriate personnel in a timely

manner.

(g) The emergency public information program ensures timely dissemination of accurate,

reliable, and understandable information.

(h) Systems are in place, if required, to alert, notify, and mobilize onsite and offsite

response personnel in case of an emergency.

(i) Procedures are in place to notify and coordinate with responsible parties when some

personnel, equipment, and facility components are not available.

(2) The emergency plan should describe how and by whom the following actions will promptly

and effectively be taken:

(a) decision to declare an alert or site area emergency

(b) activation of the onsite emergency response organization during all shifts 

(c) prompt notification of offsite response authorities that an alert or site area emergency

has been declared, including the licensee's initial recommendation for offsite

protective actions (normally within 15 minutes of classification)

(d) notification to the NRC Operations Center (as soon as possible and, in any case, no

later than 1 hour after a declared emergency)

(e) decision regarding what onsite protective actions to initiate

(f) decision regarding what offsite protective actions to recommend

(g) decision to request support from offsite organizations

(h) decision to terminate the emergency or enter recovery mode

8.4.3.1.10 Information To Be Communicated 

The emergency plan should describe the information to be communicated during an

emergency, including the following: 

(1) a standard reporting checklist to facilitate timely notification

(2) the types of information to be provided concerning facility status, radioactive releases or

hazardous chemicals incident to the processing of licensed material, and protective action

recommendations
Emergency Plan 8-9 NUREG-1520

(3) a description of preplanned protective action recommendations to be made to each

appropriate offsite organization

(4) the offsite officials to be notified, as a function of the classification of the event

(5) the recommended actions to be implemented by offsite organizations for each accident

treated in the emergency plan 

8.4.3.1.11 Training

The emergency plan should include a description of the frequency, performance objectives and

plans for the training that the licensee will provide workers on how to respond to an emergency. 

The following should be included:

(1) the topics and general content of training programs used for training the licensee’s onsite

and offsite emergency response personnel to satisfy the objectives described above 

(2) the administration of the training program, including responsibility for training, the positions

to be trained, the schedules for training, the frequency of retraining, use of team training,

and the estimated number of hours of initial training and retraining

(3) the training to be provided on the use of protective equipment, such as respirators,

protective clothing, monitoring devices, and other equipment used in emergency response

(4) the training program for onsite personnel who are not members of the emergency

response staff

(5) any special instructions and orientation tours that the licensee would offer to fire, police,

medical, and other non-licensee emergency personnel who may be required to respond to

an emergency to ensure knowledge of the emergency plan, assigned duties, and effective

response to an actual emergency

8.4.3.1.12 Safe Shutdown (Recovery and Facility Restoration)

The emergency plan should describe the following aspects of the applicant’s plans for

adequately restoring the facility to a safe status after an accident and recovery after an

emergency:

(1) the methods and responsibilities for assessing the damage to and status of the facility’s

capabilities to safely control radioactive material or hazardous chemicals associated with

the process

(2) the procedures for promptly determining the actions necessary to reduce any ongoing

releases of radioactive material or hazardous chemicals incident to the processing of

licensed material, and to prevent further incidents

(3) the provisions for promptly and effectively accomplishing required restoration actions

(4) key positions in the recovery organization 
Emergency Plan 8-10 NUREG-1520

8.4.3.1.13 Exercises and Drills

The emergency plan should state the applicant’s commitment to conduct exercises and drills in

a manner that demonstrates the capability of the organization to plan and perform an effective

response to an emergency. An adequate plan should demonstrate the following:

(1) Task-related knowledge is demonstrated through periodic participation by all qualified

individuals for each position in the emergency response organization.

(2) Drill performance is assessed against specific scenario objectives, using postulated

accidents, that adequately test personnel, equipment, and resources, including previously

identified weaknesses.

(3) Effective player, controller, evaluator, and observer pre-drill briefings are conducted.

(4) Scenario data and exercise messages provided by the controllers effectively maintain the

timeline and do not interfere with the emergency organization's response to exercise

scenario events, except where safety considerations are involved.

(5) Trained evaluators are used to identify and record participant performance, scenario

strengths and deficiencies, and equipment problems.

(6) Prestaging of equipment and personnel is minimized to realistically test the activation and

staffing of emergency facilities.

(7) Critiques are conducted in a timely manner and include a followup plan for correcting any

identified weaknesses and improving training effectiveness.

(8) Emergency drills demonstrate that resources are effectively used to control the site,

mitigate further damage, control radiological releases, perform required onsite activities

under simulated radiation/airborne and other emergency conditions, accurately assess the

facility’s status during an accident, and initiate recovery.

(9) Emergency drills demonstrate personnel protection measures, including controlling and 

minimizing hazards to individuals during fires, medical emergencies, mitigation activities,

search and rescue, and other similar events.

(10) The emergency drills demonstrate that onsite communications effectively support

emergency response activities.

(11) The emergency drills demonstrate that the emergency public information

organization disseminates accurate, reliable, timely, and understandable

information.

(12) Provisions are made for conducting quarterly communications checks with offsite

response organizations.

(13) Offsite organizations are invited to participate in the biennial onsite exercise, which

tests the major elements of the emergency plan and response organizations.
2

 Provisions for approving changes to the emergency plan and the procedures and

individuals authorized to make those changes should be clearly stated.

3

 Amendments to emergency plans that do not affect an organization and those allowed

by 10 CFR 70.32(i) need not be provided to offsite organizations before being submitted them

to the NRC.

Emergency Plan 8-11 NUREG-1520

8.4.3.1.14 Responsibilities for Developing and Maintaining the Emergency Program and

Its Procedures Current 

The emergency plan should describe the following aspects of the responsibilities for developing

and maintaining the emergency program and its procedures:

(1) the means for ensuring that revisions to the emergency plan and the procedures used to

implement the emergency plan are adequately prepared, kept up to date (normally within

30 days of any changes), and distributed to all affected parties including the NRC

(2) the provisions for approving the implementing emergency procedures, making and

distributing changes to the procedures, and ensuring that each person responsible for an

emergency response function has immediate access to a current copy of emergency

procedures 2

(3) procedures for allowing offsite response organizations 60 days to comment on any new

emergency plan or significantly updated emergency plans 3

8.4.3.2 Evaluation That No Emergency Plan Is Required

The adequacy of the evaluation (or the referenced ISA Summary) that no emergency plan is

required should be reviewed by the staff against the requirements in 10 CFR 70.22(i)(2) and the

specific criteria given in Sections 8.4.3.2.1 through 8.4.3.2.4 of this SRP. This evaluation

should be acceptable if the regulatory requirements and the following acceptance criteria are

met.

8.4.3.2.1 Facility Description

The evaluation should include a description of the facility and site, the area near the site, and

the licensed activities conducted at the facility. To be considered sufficient to support the

evaluation, these descriptions should include the following:

(1) a detailed drawing of the site showing (a) onsite and near offsite (within 1.61 Km, or 1 mile)

structures, with building numbers and labels; (b) roads and parking lots onsite and main

roads near the site; (c) site boundaries, showing fences and gates; (d) major site features;

(e) water bodies within approximately 1.61Km (1 mile); and (f) the location(s) of nearest

residents

(2) the stack heights, typical stack flow rates, and efficiencies of any emission-control devices

(3) a general description of licensed and other major activities conducted at the facility, and

the type, form, and quantities of radioactive material used
Emergency Plan 8-12 NUREG-1520

8.4.3.2.2 Types of Accidents

The evaluation should describe or refer to each type of accident identified by the ISA Summary

that has maximum offsite consequences that exceed the limit specified in 10 CFR 70.22(i)(1)(i). 

In addition, the following information should be available for review: 

(1) the process and physical location where it could occur

(2) complicating factors and offsite consequences, including non-radioactive hazardous

chemicals incident to the processing of licensed material that are released

(3) the accident sequence that has the potential for the greatest radiological and toxic

chemical impact

8.4.3.2.3 Detection of Accidents

For each type of accident identified, the evaluation should describe the following:

(1) the means of detecting the accident

(2) the means of detecting any release of radioactive or hazardous chemicals incident to the

processing of licensed material

(3) the means of alerting the operating staff

(4) the anticipated response of the operating staff

8.4.3.2.4 Evaluation of Maximum Public Exposure

To demonstrate that no emergency plan is required, an applicant may either (1) request that its

total possession limit for radioactive material be reduced below the emergency plan threshold in

10 CFR 70.22(i)(1); or (2) perform a site-specific evaluation (or refer to the ISA Summary, as

appropriate) to demonstrate that maximum public exposure is less than the limits specified in

10CFR 70.22(i)(1)(i).

The evaluation should make available the following information sufficient to allow for

independent verification:

(1) type of accident (e.g., fire, explosion, hazardous chemicals incident to the processing of

licensed material that are released, and nuclear criticality) 

(2) location of accident 

(3) maximum source term 

(4) solubility of material

(5) facility design or IROFS and the proposed release fraction 

(6) location and distance of nearest member of the public to the facility 

(7) dose model used and the process used to verify the reliability of the model and validity of

the assumptions 

(8) assumed worst-case weather condition

(9) maximum calculated exposure to a member of the public at the facility boundary 
Emergency Plan 8-13 NUREG-1520

The evaluation should list and describe the factors in 10 CFR 70.22(i)(2) that the applicant

considered in evaluating the maximum dose to members of the public. The applicant should

demonstrate why the factors used in the evaluation are appropriate when compared with the

factors in NUREG-1140. If the factors and evaluation show that the maximum dose to a

member of the public offsite from a release of radioactive materials could not exceed 0.01 Sv 

(1 rem) effective dose equivalent or the intake of soluble uranium of 2 milligrams, no

emergency plan is required in accordance with 10 CFR 70.22(i)(1)(i).

8.5 AMENDMENTS OR CHANGES TO THE EMERGENCY PLAN

The applicant may make changes to the approved emergency plan without NRC approval,

provided that the changes do not decrease the effectiveness of the plan and the applicant

provides copies of the changes to the NRC and appropriate organizations within 6 months of

making the changes in accordance with 10 CFR 70.32(i). Proposed changes that decrease the

effectiveness of the emergency plan may not be implemented without prior application to and

prior approval of the NRC.

8.6 REVIEW PROCEDURES

8.6.1 Acceptance Review

The primary reviewer should evaluate the application to determine whether it addresses the

“Areas of Review” discussed in Section 8.3, above. If significant deficiencies are identified, the

reviewer(s) should request the applicant to submit additional material before the safety

evaluation begins.

8.6.2 Safety Evaluation

After determining that the application is acceptable for review in accordance with Section 8.6.1,

above, the primary reviewer should perform a safety evaluation against the acceptance criteria

described in Section 8.4, above. If, during the course of the safety evaluation, the primary

reviewer identifies the need for additional information, the primary reviewer should coordinate

with the licensing project manager in preparing a request for additional information.

8.6.2.1 Emergency Plan

After an acceptable application has been received from the applicant, the primary reviewer

should conduct a complete review of the applicant’s emergency plan and assess its

acceptability in accordance with Section 8.4.3.1, above. The reviewer should verify that

emergency planning is consistent with the potential accident sequences described in the ISA

Summary. The ISA Summary reviewer and emergency plan reviewer should coordinate to

ensure the resolution of any issues concerning the emergency plan relative to ISA Summary

information.

Although the bulk of this information should be found in the “Emergency Management Program”

section of the licensee’s submittal, the primary and secondary reviewers should gain familiarity

with the site, including its demography, land use, facility design and layout, and major accidents

postulated by the applicant, which are presented in relevant sections of the application. The

primary and secondary reviewers should also gain familiarity with proposed radiation protection

activities and other operational matters that interface with emergency plans (particularly the
Emergency Plan 8-14 NUREG-1520

functions reviewed using Chapters 4 and 11 of this SRP). Draft and final environmental

statements for the proposed facility should be consulted. This information may be

supplemented by a personal visit to the site by the reviewer and meetings with the applicant. 

As the final step, the primary reviewer should prepare an SER in accordance with Section 8.7,

“Evaluation Findings,” of this SRP.

8.6.2.2 Evaluation That No Emergency Plan Is Required 

The primary reviewer should verify that the evaluation is consistent with the potential accident

sequences described in the ISA Summary. The ISA Summary reviewer and the primary

reviewer should coordinate to ensure the resolution of any issues concerning the evaluation

relative to ISA information. As the final step, the primary reviewer should prepare an SER, in

accordance with Section 8.7, that either agrees with the applicant’s conclusion that no

emergency plan is required or indicates that the staff does not accept the applicant’s evaluation

and recommends that an emergency plan be required.

8.7 EVALUATION FINDINGS

The primary reviewer writes an SER section addressing each topic reviewed under this SRP

chapter and explains why the NRC staff has reasonable assurance that the emergency

management part of the application is acceptable. License conditions may be proposed to

impose requirements where the application is deficient. The report includes a summary

statement of what was evaluated and why the reviewer finds the submittal acceptable. The

staff can document the evaluation as follows:

The staff has evaluated [Insert a summary statement of what was evaluated and

why the reviewer finds the submittal acceptable.] In accordance with 

10 CFR 70.22(i), the licensee commits to maintain and execute an emergency

plan for responding to the radiological hazards resulting from a release of

radioactive material or hazardous chemicals incident to the processing of

licensed material. The NRC staff reviewed the emergency plan with respect to

10 CFR 70.22(i) and the acceptance criteria in Section 8.4.3 of the SRP. The

NRC staff determined that the applicant’s emergency plan is adequate to

demonstrate compliance with 10 CFR 70.22(i), in that (1) the facility is properly

configured to limit releases of radioactive materials in the event of an accident;

(2) a capability exists for measuring and assessing the significance of accidental

releases of radioactive materials; (3) appropriate emergency equipment and

procedures are provided onsite to protect workers against radiation and other

chemical hazards that might be encountered after an accident; (4) a system has

been established to notify Federal, State, and local Government agencies and to

recommend appropriate protective actions to protect members of the public; and

(5) necessary recovery actions are established to return the facility to a safe

condition after an accident.

The requirements of the emergency plan are implemented through approved

written procedures. Changes that decrease the effectiveness of the emergency

plan may not be made without NRC approval. The NRC will be notified of other

changes that do not decrease the effectiveness of the emergency plan within 6

months of making the changes.
Emergency Plan 8-15 NUREG-1520

8.8 REFERENCES

U.S. Nuclear Regulatory Commission, “Part 30 Statements of Consideration and Emergency

Preparedness for Fuel Cycle and Other Radioactive Material Licensees,” 54 FR 14051, 1989.

U.S. Nuclear Regulatory Commission, “Nuclear Fuel Cycle Accident Analysis Handbook,”

NUREG/CR-6410, 1998. 

U.S. Nuclear Regulatory Commission, “Response Technical Manual (RTM) 96,” NUREG/BR0150, Vol. 1, Rev. 4, 1996.

U.S. Environmental Protection Agency, “Manual of Protective Action Guides and Protective

Actions for Nuclear Incidents,” EPA 400-R-92-001, May 1992.

Accession #: ML13370369
SRP- Environmental Protection 9-1 NUREG-1520

9 ENVIRONMENTAL PROTECTION

9.1 PURPOSE OF REVIEW

The purpose of this review is to determine whether the applicant’s proposed environmental

protection measures are adequate to protect the environment and the health and safety of the

public and comply with the regulatory requirements imposed by the Commission in 

10 CFR Parts 20, 51, and 70. In addition, the staff will determine whether the applicant has

submitted an environmental report that is adequate for staff use in preparing an Environmental

Assessment (EA) and Finding of No Significant Impact (FONSI), or an Environmental Impact

Statement (EIS), pursuant to Part 51. 

The staff should coordinate the preparation of an EA and FONSI, or EIS, with the

Environmental Review Team in the Division of Waste Management. Staff coordination is

described in Section 9.6.2 of this chapter.

9.2 RESPONSIBILITY FOR REVIEW

Primary: Environmental Engineer/Scientist

Secondary: Licensing Project Manager

Supporting: Fuel Cycle Facility Inspector

Radiation Safety Reviewer

ISA Primary Reviewer

9.3 AREAS OF REVIEW

The two distinct components of the application that require an environmental review are (1) the

environmental report and (2) the description of environmental protection measures. (The review

of environmental protection measures includes a review of the applicant’s ISA.) Sections 9.3.1

and 9.3.2, below, identify the areas of review for each of these components, and Section 9.4 of

this chapter, which specifies the review acceptance criteria, provides greater detail.

9.3.1 Environmental Report

The regulatory requirements for the environmental report are contained in 10 CFR Part 51. The

Commission promulgated regulations to implement the National Environmental Policy Act

(NEPA) of 1969, which requires an assessment of the environmental impacts associated with

all major Federal actions. For all licensing actions that may have a significant effect on the

environment, the NRC staff conducts an independent assessment on the basis of the

information provided in the applicant’s environmental report. This assessment is documented

in an EA or EIS. The Commission has determined that actions listed in 10 CFR 51.22(c) have

insignificant environmental impacts and are categorically excluded from the requirement for an

EA and an environmental report. However, the applicant may be required to submit information

to the NRC to justify the applicability of the categorical exclusion. In accordance with 

10 CFR 51.21, the Commission may, in special circumstances, prepare an EA on an action

covered by a categorical exclusion.

The areas of review for the environmental report correspond to the content specified in 

10 CFR 51.45:
SRP- Environmental Protection 9-2 NUREG-1520

• Date of Application

• Environmental Considerations

- Description of the proposed action

- Purpose of the proposed action

- Description of the affected environment

- Discussion of considerations (including environmental impacts and alternatives 

 to the proposed action)

• Analysis

• Status of Compliance

• Adverse Information

The environmental report may include, reference, or supplement information that the applicant

previously submitted to the NRC for prior licensing actions.

9.3.2 Environmental Protection Measures

The regulatory requirements for environmental protection are contained in 10 CFR Parts 20, 51

and 70. The NRC staff focuses its environmental review on that part of the plant-wide safety

program that the applicant established to control and assess the level of radioactive and

nonradioactive releases (gaseous, liquid, and solid) to the environment. Therefore, the staff

reviews the effluent control portion of the applicant’s radiation protection program, as well as

the applicant’s effluent and monitoring practices.

To receive authorization to possess a critical quantity of SNM, as defined in 10 CFR 70.4, an

applicant must also perform an ISA and prepare an ISA Summary in accordance with 

Subpart H of 10 CFR Part 70. Guidance on the ISA is presented in Chapter 3 of this SRP. The

environmental safety review of the ISA Summary will include a review of the identified potential

accident sequences that result in radiological and nonradiological releases to the environment,

the IROFS that the applicant specified to reduce the risk of those accidents, and the associated

management measures that provide reasonable assurance that the IROFS will perform their

designated safety functions. 

Thus, environmental protection encompasses three main components, as necessary, including 

(1) the radiation protection program; (2) effluent and environmental monitoring; and (3) the ISA

Summary and other ISA documentation as described in Sections 9.3.2.1 - 9.3.2.3, respectively.

9.3.2.1 Radiation Protection

The staff’s review of the environmental radiation protection program encompasses the following

areas:

• ALARA goals for effluent control

• effluent controls to maintain public doses ALARA

• ALARA reviews and reports to management

• waste-minimization practices and, for new operations, design plans for waste-minimization
SRP- Environmental Protection 9-3 NUREG-1520

9.3.2.2 Effluent and Environmental Monitoring

The staff’s review of the applicant’s effluent and environmental monitoring practices

encompasses the following areas:

• in-place filter-testing procedures for air-cleaning systems

• known or expected concentrations of radionuclides in effluents

• physical and chemical characteristics of radionuclides in discharges

• discharge locations

• environmental media to be monitored and the sample locations

• sampling collection and analysis procedures, including the minimum detectable

concentrations of radionuclides

• action levels and actions to be taken when the levels are exceeded

• permits, including air discharge and National Pollutant Discharge and Elimination system

permits

• leak-detection systems for ponds, lagoons, and tanks

• pathways analysis methods to estimate public doses

• recording and reporting procedures

• solid waste handling and disposal programs

9.3.2.3 ISA Summary

The staff’s review of the applicant’s ISA Summary related to environmental protection includes

the following areas:

• accident sequences (and associated facility processes) that, if unmitigated, would

result in releases to the environment

• likelihood and environmental consequences of these accident sequences

• controls relied on to reduce the unmitigated risk from “high” risk to an acceptable level

é availability and reliability of controls

9.4 ACCEPTANCE CRITERIA

Acceptance criteria for the environmental report and environmental protection measures are

described in Sections 9.4.3.1 and 9.4.3.2, respectively.

9.4.1 Regulatory Requirements

To be considered acceptable, the applicant must satisfy the follow’s regulatory requirements

regarding environmental protection:

(1) 10 CFR Part 20 specifies the effluent control and treatment measures necessary to

meet the dose limits and dose constraints for members of the public specified in 

Subparts B, D, and F; the survey requirements of Subpart F; the waste disposal

requirements of Subpart K; the records requirements of Subpart L; and the reporting

requirements of Subpart M.

(2) 10 CFR Part 51 specifies that the applicant must establish effluent and environmental

monitoring systems to provide the information required by 10 CFR 51.60(a).

(3) 10 CFR Part 51 also specifies that the applicant must submit an environmental report, as

required by 10 CFR 51.60(b), or support a categorical exclusion as described in 

10 CFR 51.22(c).
SRP- Environmental Protection 9-4 NUREG-1520

(4) 10 CFR Part 70 specifies that the applicant must demonstrate that proposed facilities and

equipment, including measuring and monitoring instruments and devices for the disposal

of radioactive effluents and wastes, are adequate to protect the environment and the

health and safety of the public, as specified in 10 CFR 70.22(a)(7).

(5) 10 CFR Part 70 also specifies that the applicant for a facility (as described in 

10 CFR 70.4) must submit a safety assessment of the design basis of the principal

structures, systems, and components of the plant, including provisions for protection

against natural phenomena, as specified in 10 CFR 70.22(f).

(6) 10 CFR Part 70 also specifies that an applicant for a facility must provide an ISA

Summary that includes a list of the IROFS established by the applicant and other

elements, as described in 10 CFR 70.65(b).

(7) 10 CFR 70.59 outlines the radiological effluent monitoring reporting requirements for a 

10 CFR Part 70 licensee.

9.4.2 Regulatory Guidance

The regulatory guidance for environmental protection is contained in the following NRC industry

documents: 

(1) NRC Regulatory Guide 4.5, “Measurements of Radionuclides in the Environment -

Sampling and Analysis of Plutonium in Soil.”

(2) NRC Regulatory Guide 4.15, “Quality Assurance for Radionuclide Monitoring Programs

(Normal Operations) - Effluent Streams and the Environment.”

(3) NRC Regulatory Guide 4.16, “Monitoring and Reporting Radioactivity in Releases of

Radioactive Materials in Liquid and Gaseous Effluents from Nuclear Fuel Processing

and Fabrication Plants and Uranium Hexafluoride Production Plants.”

(4) NRC Regulatory Guide 4.20, “Constraint on Releases of Airborne Radioactive Materials

to the Environment for Licensees Other than Power Reactors.”

(5) NRC Regulatory Guide 8.37, “ALARA Levels for Effluents from Materials Facilities.”

(6) NRC Information Notice 94-07, “Solubility Criteria for Liquid Effluent Releases to

Sanitary Sewerage Under the Revised 10 CFR Part 20,” January 28, 1994.

(7) NRC Information Notice 94-23, “Guidance to Hazardous, Radioactive, and Mixed Waste

Generators on the Elements of a Waste Minimization Program,” March 1994.

(8) ANSI N13.1-1982, “Guide to Sampling Airborne Radioactive Materials in Nuclear

Facilities.”

(9) ANSI N42.18-1980, “Specification and Performance of On-site Instrumentation for

Continuously Monitoring Radioactive Effluents.”
SRP- Environmental Protection 9-5 NUREG-1520

(10) NCRP Report No. 123, “Screening Models for Releases of Radionuclides to

Atmosphere, Surface Water, and Ground,” January 1996.

(11) Draft Environmental Standard Review Plan, 2000.

9.4.3 Regulatory Acceptance Criteria

9.4.3.1 Environmental Report or Categorical Exclusion

An environmental report is required for actions listed in 10 CFR 51.60(b). The acceptance

criteria for the environmental report are discussed in Section 9.4.3.1.1, below.

An environmental report is not required for licensing actions that meet the requirements for a

categorical exclusion, as defined in 10 CFR 51.22(c). However, if pursuant to 

10 CFR 51.23(c)(11), the action involves an amendment to licenses for fuel cycle plants,

radioactive waste disposal sites, and other materials licenses identified in 10 CFR 51.60(b)(1),

for changes in process operations or equipment, the applicant must justify that the action will

not result in significant effects on the environment. The acceptance criteria for this categorical

exclusion are given in Section 9.4.3.1.2, below. 

9.4.3.1.1 Environmental Report

(1) Date of Application

The date of an application for a license to possess and use SNM for processing and fuel

fabrication, scrap recovery, conversion of uranium hexafluoride, or any other activity that

the NRC has determined (pursuant to 10 CFR Part 51, Subpart A) will significantly affect

the quality of the environment, is acceptable if the application is submitted at least 9

months before construction begins, as required by 10 CFR 70.21(f). However, an EIS is

generally estimated to take 2 years to complete.

(2) Environmental Considerations

An adequate environmental report addresses the requirements of 10 CFR 51.45(b), 

as follows:

a. Description of the proposed action

The description of the proposed action includes a brief summary of the significant

characteristics of the proposed facility, including the major site features and plant

design and operating parameters. The description includes a complete discussion of

how SNM will be processed at the facility. If future construction is proposed, the

description includes a proposed project schedule showing the initiation dates for site

preparation, plant construction, and operation.

b. Purpose of the proposed action

The purpose statement demonstrates a need for the proposed action. At a minimum, 

this demonstration provides (i) the quantities of SNM used for domestic benefit , (ii) a

projection of domestic and foreign requirements for the services and (iii) alternative

sources of supply for the facility's proposed services. If delay of the proposed project
SRP- Environmental Protection 9-6 NUREG-1520

would affect the Nation's energy program or the applicant's business (such as loss of

contracts, jobs, or future business), these effects are discussed.

c. Description of the affected environment

The description of the affected environment includes the following details

i. site location (including longitude and latitude) and facility layout

ii. regional demography and land use

iii. socioeconomic information, including low-income and minority populations within

a 50 mile radius

iv. regional historic, archaeological, architectural, scenic, cultural, and

natural landmarks

v. local meteorology and air quality 

vi. local surface water and groundwater hydrology

vii. regional geology and seismology

viii. local terrestrial and aquatic ecology

To the extent possible, this information is current and affects observations and

measurements made over a period of years, especially for conditions that are

expected to vary seasonally (e.g., precipitation, wind speed and direction, and

groundwater levels).

d. Discussion of considerations

The discussion of considerations includes the following details: 

i. Impact of the proposed action on the environment

• effects of site preparation and construction on land and water use

• effects of facility operation on the human population (including consideration of

occupational and public radiation exposure) and important biota

• any irreversible commitments of resources because of site preparation and

facility construction and operation, such as destruction of wildlife habitat,

removal of land from agricultural use, and diversion of electrical power

• plans and policies regarding decommissioning and dismantling at the end of

the facility’s useful life

• environmental effects of the transportation of radioactive materials to and from

the site

• environmental effects of accidents

é impacts on air and water quality

é impacts on cultural and historic resources
SRP- Environmental Protection 9-7 NUREG-1520

 This section of the environmental report discusses the impacts on the

environment in proportion to their significance, and considers the cumulative

impacts of the proposed action. In addition, accident analyses provided in the

report must be consistent with the applicant’s ISA.

ii. Adverse environmental effects

The information submitted describes any adverse environmental effects that

cannot be avoided if the proposal is implemented. To the maximum extent

possible, this description is presented in quantitative terms. This discussion

makes clear which of these effects are unavoidable and subject to later

amelioration, and which are unavoidable and irreversible. The description

includes specific measures that the applicant could take to mitigate adverse

effects.

iii. Alternatives to the proposed action

The discussion of alternatives to the proposed action is sufficiently complete to

aid the NRC in developing and exploring, pursuant to Section 102(2)(E) of the

NEPA, "appropriate alternatives to recommended courses of action in any

proposal which involves unresolved conflicts concerning alternative uses of

available resources." To the extent practicable, the environmental impacts of the

proposal and the alternatives are presented in comparative form.

The discussion of alternatives includes siting and design alternatives. 

Comparable levels of information on each site need not be presented as long as

the applicant presents sufficient information to facilitate a fair and reasonable

comparison. The following factors are considered when comparing alternative

sites:

• physical characteristics of the area, including demographic, geological,

hydrological, meteorological, and seismological conditions of the site and

surrounding area

• location of power sources and transmission lines

• location of the major product market

• location of raw materials, components, and sources of supply

• availability of air, rail, roads, and water for transport of raw materials and

supplies, finished products, and solid wastes

• commitment of natural resources for site preparation and facility construction,

including but not limited to the destruction or diminution of wildlife habitats,

flora, woodlands, and marshlands

• commitment of capital for site preparation and facility construction

• cost of operation, including consideration of labor supply, prevailing wage

rates, and other recurring or nonrecurring costs
SRP- Environmental Protection 9-8 NUREG-1520

• availability of municipal services and facilities or, conversely, the cost of

providing services such as water and sewage treatment

• requirements for relocating homes and families

• existing and projected land use and economic status of the community 

(e.g., urban, rural, industrial, stable)

iv. Relationship between short-term uses and long term productivity

The applicant should discuss relationship between local short-term uses of man's

environment and the maintenance and enhancement of long-term productivity. 

Short-term uses are considered to be those that occur during the active life of the

facility. Long-term productivity represents the use of the environment beyond

decommissioning of the facility.

v. Irreversible and irretrievable commitments of resources

The applicant should discuss any irreversible environmental commitments and

irretrievable material resources that would be involved in the proposed action.

(3) Analysis of Environmental Effects of Proposed Action and Alternatives

An acceptable environmental report analyzes the environmental effects of the proposed

action and alternatives. In accordance with 10 CFR 51.45(c), the analysis considers and

balances the environmental effects of the proposed action and the alternatives available to

reduce or avoid adverse environmental effects, as well as the environmental, economic,

social, and other benefits of the proposed action.

To the fullest extent practicable, this analysis quantifies the various factors considered. If

the application involves renewal or amendment of a current license, this analysis quantities

environmental impacts using environmental monitoring data collected by the licensee. To

the extent that there are important qualitative considerations or factors that cannot be

quantified, the analysis discusses those considerations and factors in qualitative terms. 

The analysis contains sufficient data to aid the staff in developing an independent analysis.

(4) Status of Compliance

As required by 10 CFR 51.45(d), the applicant should list all Federal permits, licenses,

approvals, and other entitlements, that must be obtained in connection with the proposed

action. The list is acceptable if it is complete and current as of the 

application date.

In addition, 10 CFR 51.45(d) requires that the environmental report must include a

discussion of the status of compliance with applicable environmental quality standards and

requirements including, but not limited to, applicable zoning and land use regulations, and

thermal and other water pollution limitations or requirements that have been imposed by

Federal, State, regional, and local agencies having responsibility for environmental

protection. The discussion is acceptable if it addresses whether each alternative will
SRP- Environmental Protection 9-9 NUREG-1520

comply with such applicable environmental quality standards and requirements. The

discussion includes, but is not limited to, the following Federal laws: 

• National Historic Preservation Act of 1966

• Fish and Wildlife Coordination Act of 1966

• Wild and Scenic Rivers Act of 1968

• Endangered Species Act Amendments of 1978

• Coastal Zone Management and Improvement Act of 1990

(5) Adverse Information

In accordance with 10 CFR 51.45(e), the preceding discussions and analyses are

acceptable if they include all information that supports and/or impedes the proposed

actions.

9.4.3.1.2 Categorical Exclusion

An environmental report is not required (that is, categorically excluded) for actions, identified in

10 CFR 51.60(b)(1), that involve an amendment to licenses for fuel cycle plants, radioactive

waste disposal sites, and other radioactive material uses that are not expected to result in

significant environmental impacts. However, when amendments involve changes in process

operations or equipment as defined in 10 CFR 51.22(c)(11), the applicant needs to justify that

the changes will not result in significant environmental effects.

The information provided by the applicant to justify the categorical exclusion determination is

acceptable if it demonstrates the following, as specified in 10 CFR 51.22(c)(11):

• There is no significant change in the types or significant increase in the amounts of any

effluents that may be released offsite.

• There is no significant increase in individual or cumulative occupational radiation exposure.

• There is no significant construction impact.

• There is no significant increase in the potential for or consequences from radiological

accidents.

9.4.3.2 Environmental Protection Measures

An applicant’s proposed environmental protection measures are acceptable if they provide for

qualified and trained staff, effluent control, and effluent and environmental monitoring, in

accordance with the NRC’s requirements. Using the acceptance criteria defined in Chapter 11

of this SRP, the NRC staff will review qualifications and training that the applicant has

established for plant personnel who are associated with environmental protection. This will

include the qualification and training of managers, supervisors, technical staff, operators,

technicians, and maintenance personnel whose levels of knowledge are important to the

environment and protect the health and safety of the public. Managers and staff will be

expected to have levels of education and experience commensurate with the responsibilities of

their positions. 
SRP- Environmental Protection 9-10 NUREG-1520

9.4.3.2.1 Radiation Safety

In accordance with 10 CFR 20.1101, each licensee must implement a radiation protection

program, which is discussed in detail in Chapter 4 of this SRP. The environmental review of the

radiation protection program focuses on the applicant’s methods to maintain public doses

ALARA in accordance with 10 CFR 20.1101. NRC guidance on compliance with these

regulations can be found in Regulatory Guide 8.37, “ALARA Levels for Effluents from Materials

Facilities,” dated July 1993.

Specifically, 10 CFR 20.1101(d) requires the applicant to establish constraints on airborne

emissions of radioactive material to the environment, excluding radon-222 and its decay

products. Such constraints must ensure that the individual member of the public who is likely to

receive the highest dose will not be expected to receive a TEDE in excess of 0.1 mSv 

(10 mrem) per year from these emissions. To meet the reporting requirements of 10 CFR

20.2203, the applicant must have (and describe) procedures to report to the NRC when these

dose constraints are exceeded, and must take prompt appropriate corrective action to prevent

recurrence. NRC guidance on compliance with this regulation can be found in Regulatory

Guide 4.20, “Constraint on Releases of Airborne Radioactive Materials to the Environment for

Licensees Other than Power Reactors,” dated December 1996.

The environmental review of the radiation protection program also focuses on the applicant’s

waste minimization practices. Applicants for new licenses are required to comply with 10 CFR

20.1406, which states that the applicant must describe how facility design procedures for

operation will, to the extent practicable, minimize contamination of the facility and the

environment, facilitate eventual decommissioning, and minimize the generation of radioactive

waste. Applicants requesting amendment or renewal of existing licenses must minimize and

control waste generation during operations as part of the radiation protection program, in

accordance with 10 CFR 20.1101 [62 FR 39082].

Guidance for waste minimization programs can be found in NRC Information Notice 94-23,

“Guidance to Hazardous, Radioactive, and Mixed Waste Generators on the Elements of a

Waste Minimization Program,” dated March 25, 1994. More information on compliance with the

decommissioning aspects of the waste minimization regulations can be found in Chapter 10 of

this SRP.

The proposed radiation protection program is acceptable if it satisfies the following criteria:

(1) Radiological (ALARA) Goals for Effluent Control

ALARA goals are set at a modest fraction (10 to 20 percent) of the values in 

10 CFR Part 20, Appendix B, Table 2 Columns 1 and 2, and Table 3 and the external

exposure limit in 10 CFR 20.1302(b)(2)(ii), or the dose limit for members of the public, if

the applicant proposes to demonstrate compliance with 10 CFR 20.1301, through a

calculation of the TEDE to the individual likely to receive the highest dose.

An applicant’s constraint approach is acceptable if it is consistent with guidance found in

Regulatory Guide 4.20, and the applicant’s description of the constraint approach provides

sufficient detail to demonstrate specific application of the guidance to proposed routine and

non-routine operations, including anticipated events.
SRP- Environmental Protection 9-11 NUREG-1520

(2) Effluent Controls to Maintain Public Doses ALARA

The applicant describes and commits to use effluent controls (e.g., procedures,

engineering controls, and process controls) to maintain public doses ALARA. Common

control practices include filtration, encapsulation, adsorption, containment, recycling,

leakage reduction, and storage of materials for radioactive decay. Practices for large,

diffuse sources (such as contaminated soils or surfaces) include covers, wetting during

operations, and application of stabilizers. The applicant must demonstrate a commitment

to reduce unnecessary exposure to members of the public and releases to the

environment.

Engineering options that do not substantially reduce the collective dose and require

unreasonable costs are not required. Reasonableness can be founded on qualitative or

quantitative cost/benefit analyses. Quantitative analyses may use a value of $2000 per

person-cSv (person-rem), as discussed in NUREG-1530, “Reassessment of the NRC’s

Dollar per Person-Rem Conversion Factor Policy.”

(3) ALARA Reviews and Reports to Management

The applicant commits to annual review of the content and implementation of the radiation

protection program, which includes the ALARA effluent control program. This review

includes analysis of trends in release concentrations, environmental monitoring data, and

radionuclide usage; determines whether operational changes are needed to achieve the

ALARA effluent goals; and evaluates all designs for system installations or modifications. 

The applicant also commits to report the results to senior management, along with

recommendations for changes in facilities or procedures that are necessary to achieve

ALARA goals.

(4) Waste minimization

To comply with 10 CFR 20.1406, applications for new licenses must describe how the

facility’s design procedures for operation will minimize, to the extent practicable,

contamination of the facility and the environment, and generation of radioactive waste. 

Waste minimization programs proposed by applicants for both new and existing licenses

are acceptable if the programs include:

• top management support

• methods to characterize waste generation (including types and amounts) and waste

management costs (including costs of regulatory compliance, paperwork, transportation,

treatment, storage, disposal, etc.)

• periodic waste minimization assessments to identify waste minimization opportunities

and solicit employee or external recommendations

• provisions for technology transfer to seek and exchange technical information on waste

minimization

• methods to implement and evaluate waste minimization recommendations
SRP- Environmental Protection 9-12 NUREG-1520

9.4.3.2.2 Effluent and Environmental Controls and Monitoring

(1) Effluent Monitoring

The applicant's effluent monitoring is considered acceptable if it meets the following

criteria:

(a) The known or expected concentrations of radioactive materials in airborne and liquid 

effluents are ALARA and are below the limits specified in Table 2 of Appendix B to 

10 CFR Part 20, or site-specific limits established in accordance with 

10 CFR 20.1302(c).

If, in accordance with 10 CFR 20.1302(c), the applicant proposes to adjust the effluent

concentrations in Appendix B to 10 CFR Part 20, to account for the actual physical

and chemical characteristics of the effluents, the applicant must provide information

regarding aerosol-size distributions, solubility, density, radioactive-decay equilibrium,

and chemical form. This information must be complete and accurate to justify the

derivation and application of the alternative concentration limits for the radioactive

materials.

(b) If the applicant proposes to demonstrate compliance with 10 CFR 20.1301 using a

calculation of the TEDE to the individual who is likely to receive the highest dose in

accordance with 10 CFR 20.1302(b)(1) , the calculation of the TEDE by pathway

analyses must be supported by appropriate models, codes, and assumptions that

accurately represent the facility, site, and, the surrounding area. In addition, the

assumptions must be reasonable, input data must be accurate, all applicable

pathways must be considered, and the results must be interpreted correctly. 

NCRP Report No. 123, “Screening Models for Releases of Radionuclides to

Atmosphere, Surface Water, and Ground,” dated January 1996, provides acceptable

methods for calculating the dose from radioactive effluents. The use of computer

codes is acceptable for pathway analyses if the applicant is able to demonstrate that

any code used has undergone validation and verification to demonstrate the validity of

estimates developed using the codes for established input sets. Dose-conversion

factors are acceptable for use in the pathway analyses if they are founded on the

methodology described in ICRP 30, "Limits for Intakes of Radionuclides by Workers,"

as reflected in USEPA’s Federal Guidance Report 11. Such methods are acceptable

for determining the dose to the maximally exposed individual during normal facility

operations and anticipated events.

(c) All liquid and airborne effluent discharge locations are identified and monitored. 

Monitoring locations are identified and, for those effluent discharge points that have

input from two or more contributing sources within the facility, sampling each

contributing source is considered necessary for effective process and effluent control. 

(d) The applicant continuously samples airborne effluents from all routine and non-routine

operations, as well as anticipated events associated with the plant, including effluents

from areas that are not used for processing SNM such as laboratories, experimental

areas, storage areas, and fuel element assembly areas. 
SRP- Environmental Protection 9-13 NUREG-1520

Effluents are sampled unless the applicant has established (by periodic sampling or

other means) that radioactivity in the effluent is insignificant and will remain so. In

such cases, the effluent is sampled at least quarterly to confirm that its radioactivity is

not significant. For the purposes of this SRP, radioactivity in an effluent is significant if

the concentration averaged over a calendar quarter is equal to 10 percent or more of

the appropriate concentration listed in Table 2 of Appendix B to 10 CFR Part 20.

(e) The sample collection and analysis methods and frequencies are appropriate for the

effluent medium and the radionuclide(s) being sampled. Sampling methods ensure

that representative samples are obtained using appropriate sampling equipment and

sample collection and storage procedures. For liquid effluents, representative

samples are taken at each release point to determine the concentrations and

quantities of radionuclides that are released to an unrestricted area, including

discharges to sewage systems. For continuous releases, samples are continuously

collected at each release point. For batch releases, a representative sample of each

batch is collected. If periodic sampling is used in lieu of continual sampling, the

applicant shows that the samples are representative of actual releases. Monitoring

instruments are calibrated at least annually, or more frequently if suggested by the

manufacturer.

(f) Radionuclide-specific analyses are performed on selected composited samples unless

either of the following criteria exists:

(i) The gross alpha and beta activities are so low that individual radionuclides could

not be present in concentrations greater than 10 percent of the concentrations

specified in Tables 2 or 3 of Appendix B to 10 CFR Part 20.

(ii) The radionuclide composition of the sample is known through operational data,

such as the composition of the feed material. 

Monitoring reports in which the quantities of individual radionuclides are estimated on

the basis of methods other than direct measurement include an explanation and

justification of how the results were obtained.

Operational data may not be adequate for the determination of radionuclide

concentration in certain cases. Such cases include but are not limited to (i) plants that

process uranium in which extraction, ammonium diuranate precipitation, ion exchange,

or other separation process could result in concentration of thorium isotopes

(principally Th-234); (ii) plants that process uranium of varying enrichments; and 

(iii) plants that process plutonium in which significant variation in the Pu-238/Pu-239

ratio among batches and the continuous in-growth of Am-241 would preclude the use

of feed material data to determine the radionuclide composition of effluents.

Radionuclide analyses are performed more frequently than usual (i) at the beginning

of the monitoring program until a predictable and consistent radionuclide composition

in effluents is established; (ii) whenever there is a significant, unexplained increase in

gross radioactivity in effluents, and (iii) whenever a process change or other

circumstance might cause a significant variation in the radionuclide composition.

(g) The minimum detectable concentration (MDC) for sample analyses is not more than 
SRP- Environmental Protection 9-14 NUREG-1520

5 percent of the concentration limits listed in Table 2 of Appendix B to 10 CFR Part 20. 

If the actual concentrations of radionuclides in samples are known to be higher than 

5 percent of the 10 CFR Part 20 limits, the analysis methods need only be adequate to

measure the actual concentration. However, in such cases, the MDC must be low

enough to accommodate fluctuations in the concentrations of the effluent and the

uncertainty of the MDC.

(h) The laboratory quality control (QC) procedures are adequate to validate the analytical

results. These QC procedures include the use of established standards such as those

provided by the National Institute of Standards and Technology (NIST), as well as

standard analytical procedures such as those established by the National

Environmental Laboratory Accreditation Conference.

(i) The proposed action levels and actions to be taken if the action levels are exceeded

are appropriate. The action levels are incremental, such that each increasing action

level results in a more aggressive action to ensure effluent control. A slightly higher

than normal concentration of a radionuclide in an effluent triggers an investigation into

the cause of the increase. Specified action level will result in the shutdown of an

operation if the specified level is exceeded. These action levels are selected on the

basis of the likelihood that a measured increase in concentration could indicate

potential violation of the effluent limits.

(j) The applicant completely and accurately describes all applicable Federal and/or State

standards for discharges, as well as any permits issued by local, State, or Federal

governments for gaseous and liquid effluents.

(k) The systems for detecting of leakage from ponds, lagoons, and tanks are adequate to

detect and ensure against any unplanned releases to groundwater, surface water,

or soil.

(l) Releases to sewer systems are controlled and maintained to meet the requirements of

10 CFR 20.2003, including (i) the material is water soluble; (ii) known or expected

discharges meet the effluent limits specified in Table 3 of Appendix B to 10 CFR Part

20; and (iii) the known or expected total quantity of radioactive material released into

the sewer system in a year does not exceed 5 Ci (185 GBq) of 3H, 1 Ci (37 GBq) of

14C, and 1 Ci (37 GBq) of all other radioactive materials combined. Solubility is

determined in accordance with the procedure described in NRC Information 

Notice 94-07.

(m) Reporting procedures comply with the requirements of 10 CFR 70.59 and the

guidance specified in Regulatory Guide 4.16. Reports of the concentrations of

principal radionuclides released to unrestricted areas in liquid and gaseous effluents

are provided and include the MDC for the analysis and the error for each data point.

(n) The applicant’s procedures and facilities for solid and liquid waste handling, storage,

and monitoring result in safe storage and timely disposition of the material.

(2) Environmental Monitoring

The scope of the applicant’s environmental monitoring is acceptable if it is commensurate with

the scope of activities at the facility and the expected impacts from operations as identified in 
SRP- Environmental Protection 9-15 NUREG-1520

the environmental report and meets the following criteria:

(a) Background and baseline concentrations of radionuclides in environmental media

have been established through sampling and analysis.

(b) Monitoring includes sampling and analyses for monitoring air, surface water,

groundwater, soil, sediments, and vegetation, as appropriate.

(c) The description of monitoring identifies adequate and appropriate sampling locations

and frequencies for each environmental medium, the frequency of sampling, and the

analyses to be performed on each medium.

(d) Monitoring procedures employ acceptable analytical methods and instrumentation. 

The applicant commits to an instrument maintenance and calibration program that is

appropriate to the given instrumentation. If the applicant proposes to use its own

analytical laboratory for the analysis of environmental samples, the applicant commits

to provide third-party verification of the laboratory's methods (such as may be obtained

by participation in a round-robin measurement program).

(e) Appropriate action levels and actions to be taken if the levels are exceeded are

specified for each environmental medium and radionuclide.

Action levels are selected on the basis of a pathway analysis that demonstrates that

below those concentrations, doses to the public will be ALARA and below the limits

specified in 10 CFR Part 20, Subpart B. The action levels specify the concentrations

at which an investigation would be performed, as well as levels at which process

operations would be shut down.

(f) MDCs are specified for sample analyses, and are at least as low as those selected for

effluent monitoring in air and water. MDCs for sediment, soil, and vegetation are

selected on the basis of action levels, to ensure that sampling and analytical methods

are sensitive and reliable enough to support application of the action levels.

(g) Data analysis methods and criteria to be used in evaluating and reporting the

environmental sampling results are appropriate and will indicate when an action level

is being approached in time to take corrective actions.

(h) The description of the status of all licenses, permits, and other approvals of facility

operations required by Federal, State, and local authorities is complete and accurate.

(i) Environmental monitoring is adequate to assess impacts to the environment from

potential radioactive and nonradioactive releases, as identified in high and

intermediate consequence accident sequences in the ISA.

9.4.3.2.3 Integrated Safety Analysis

In accordance with 10 CFR 70.60, applicants requesting a license to possess and process

greater than a critical mass of SNM are required to perform an ISA and submit an ISA

Summary to the NRC for approval. The applicant’s treatment of environmental protection in the 
SRP- Environmental Protection 9-16 NUREG-1520

ISA is acceptable if it fulfills the following criteria:

• The ISA provides a complete list of accident sequences that result in radiological and

nonradiological releases to the environment.

• The ISA provides a reasonable estimate for the likelihood and consequences of each

accident sequence identified.

• Adequate engineering and/or administrative controls are identified for each accident

sequence of environmental significance. These controls will prevent or mitigate “high”

and “intermediate” consequence sequences to an acceptable level. (Definitions of

consequence categories are given in Chapter 3 of this SRP.) IROFS provide the

indicated level of protection.

• Adequate levels of assurance are afforded so that IROFS will satisfactorily perform

their safety functions. This may be accomplished through configuration management,

training, and maintenance activities.

• The ISA uses acceptable methods to estimate environmental effects that may result

from accident sequences.

9.5 REVIEW PROCEDURES

The staff will review the environmental report and environmental protection measures to verify

that each meets the acceptance criteria defined in Section 9.4 of this chapter. If the applicant

has not provided sufficient information to make these determinations, the reviewer(s) will

request additional information in coordination with the facility project manager. The format for a

request for additional information (RAI) is specified in Chapter 4 of the Fuel Cycle Licensing

Branch “Materials Licensing Procedures Manual.” Additional review procedures are provided in

Sections 9.5.1 through 9.5.3, below.

9.5.1 Environmental Report

Review of the environmental report or information presented to support a categorical exclusion

should be coordinated with the radiation safety reviewer to assess the adequacy of the

information provided by the applicant. 

9.5.2 Environmental Protection

For renewal and amendment applications, an environmental specialist will review the applicant’s

environmental protection measures in coordination with the fuel cycle facility inspector

responsible for environmental protection. Any comments or concerns that the inspector

identifies will be addressed and resolved, and the SER (described in Section 9.6.1of this SRP)

for the licensing action will contain a statement indicating whether the inspection staff has any

objections to approval of the proposed licensing action. In addition, the review will include

evaluation of inspection reports and semi-annual effluent reports submitted in accordance with

10 CFR 70.59, to ensure licensee performance in environmental protection.

As part of the environmental protection review, the environmental specialist will review the ISA

Summary, including all identified accident sequences that can have significant environmental
SRP- Environmental Protection 9-17 NUREG-1520

consequences, to determine whether the list completely and properly identifies an potential

accident. This review will be coordinated with the ISA reviewer. Detailed review will only be

conducted of the accident sequences that, when left unmitigated, are rated as “high”

consequence by the applicant, as well as approximately 10 percent of the “intermediate”

consequence sequences, and a smaller number of accident sequences whose consequences

are less than “intermediate.” However, additional “intermediate” and “low” consequence

sequences may be evaluated on the basis of the results of the initial review.

Evaluation of the ISA Summary requires coordination with other technical reviewers. The

environmental review of the IROFS will be coordinated with the reviewers for the specific

assurance functions, such as training and maintenance.

Finally, review of the ISA Summary may require examination of the ISA and of detailed

supporting ISA documents that are located at the facility. On the basis of these reviews, the

reviewer should decide what supporting documents need to be reviewed. The reviewer will

clearly cite in the SER what materials were examined, and what descriptions and commitments

were considered and relied on, or the basis for the staff’s safety decision.

9.6 EVALUATION FINDINGS

9.6.1 Introduction

The evaluation findings for the environmental protection review are documented in (1) an SER

that documents the technical review of the application, including the review of the environmental

protection program and the ISA Summary, and (2) an EA or EIS that documents the staff’s

independent assessment of the environmental impacts of the proposed action.

The staff reviewers will verify that the information submitted by the applicant is in accordance

with 10 CFR Parts 20, 51, and 70, and is consistent with the guidance in this SRP as it applies

to environmental protection. In the input to the SER, the primary reviewer should document the

bases for determining the adequacy of the application with respect to environmental protection,

and should recommend additional license conditions in areas where the application is not

adequate. The primary reviewer also describes the applicant’s approach to ensuring the quality

and reliability of the IROFS required for environmental protection. 

Environmental protection is often reviewed and evaluated in conjunction with the environmental

report, and the environmental protection function is summarized in the EA or EIS. However, the

EA or EIS does not become part of the license. Issues identified during the review should be

discussed briefly in the SER, and any recommended license conditions founded on the analysis

in the EA or EIS should be added to the license.

If an EA or EIS is prepared for the licensing action, the date the document was issued should

be reported in the environmental safety section of the SER. If the EA resulted in a FONSI, the

FONSI’s publication date in the Federal Register should be included in the SER. If an EIS is

prepared, the SER would include the Federal Register publication date for the Record of

Decision. When applicable, the SER will also document the determination that an action meets

the requirements for a categorical exclusion.
SRP- Environmental Protection 9-18 NUREG-1520

9.6.2 NEPA Documentation and Coordination

Before taking a licensing action, the appropriate NRC Branch Chief will determine whether the

proposed action qualifies for a categorical exclusion under 10 CFR 51.22, whether an EA

should be prepared, or whether conditions warrant the staff proceeding directly to an EIS. The

NRC Branch Chief should then initiate the appropriate coordination with the Division of Waste

Management.

(1) An EIS will be prepared if the action meets the criteria defined in 10 CFR 51.20. An EA is

not necessary if the NRC Branch Chief determines that an EIS will be prepared. The NRC

Branch Chief should initiate coordination with the Division of Waste Management to

prepare the EIS.

(2) An EA will be prepared if the action meets the criteria defined in 10 CFR 51.21. If the EA

results in a FONSI, no further environmental review is necessary unless significant new

information comes to light or the licensee changes its plan. Review of the EA by the

Environmental Review Team is required, and the EA should be sent to the team for review

before being finalized. If the EA does not result in a FONSI, an EIS must be prepared. 

(3) A categorical exclusion will suffice if the action meets the criteria defined in 

10 CFR 51.22(c). An action that qualifies for a categorical exclusion is usually identified at

the start of the licensing review, and an environmental report is not required. No

coordination with the Division of Waste Management is required.

Requirements for preparing an EIS, EA, or FONSI are described in detail in 10 CFR Part 51. 

Documents prepared in accordance with NEPA will follow pertinent NMSS procedures,

including consultation with States (Policy & Procedures Letter 1-48), evaluation of

environmental justice (Policy & Procedures Letter 1-50, Revision 2), and Chapter 6 of the NRC

Division of Fuel Cycle Safety and Safeguards, “Fuel Cycle Licensing Branch Manual.” Sections

9.6.2.1 and 9.6.2.2 , below, contain an overview of the regulatory requirements for the EA,

FONSI, EIS, and Record of Decision specified in 10 CFR Part 51. However, this discussion is

not intended to be all-inclusive.

9.6.2.1 EA

The staff will prepare an EA using information contained in the license application and the

environmental report, which is separately submitted by the licensee, in accordance with 

10 CFR 51.45. The EA that the staff prepares will identify the proposed action and include the

following, in accordance with 10 CFR 51.30:

(1) brief discussion of the following considerations:

a. the need for the proposed action

b. alternatives to the proposed action, as required by NEPA Section 102(2)(E)

c. the environmental impacts of the proposed action and alternatives, as appropriate

d. any disproportionately high and adverse human health or environmental effects on 

low-income and minority populations, as required for “special case EAs,” as defined by

NMSS Policy & Procedures Letter 1-50, Revision 2, dated September 7, 1999

(2) lists of agencies and persons consulted and sources used

In preparing an EA, the staff will consult with affected States, the Fish and Wildlife Service,

the State Historic Preservation Officer, and the Tribal Officer (as necessary), on
SRP- Environmental Protection 9-19 NUREG-1520

environmental issues and will document such contacts in the EA. This documentation

(item 2,above) will include the following information identified in NMSS Policy & Procedures

Letter 1-48, dated January 1995:

(a) the name of each State, agency (including contacted individual's name), or

person consulted

(b) date of consultation(s)

(c) purpose of the consultation(s)

(d) brief summary of the views or comments expressed by the consulted party and the

staff's resolution

(e) reference to publicly available documents containing additional information, 

if applicable

Much of the information used to prepare the EA is provided in the applicant’s environmental

report. However, the staff will perform independent analyses of the environmental impacts of

the proposed action, and will discuss the conclusions of these analyses in the EA. The EA

should focus on the impacts of the proposed action, and should be no more than 15 pages

unless additional pages are necessary to explain any complicated environmental issues

associated with the proposed action.

On completion, the EA should be forwarded to the Environmental Review Team for review.

The Environmental Review Team will review the EA to ensure consistency among all EAs

prepared by NMSS. When the Environmental Review Team completes its review, the

appropriate NRC Branch Chief will determine whether to prepare an EIS or a FONSI on the

proposed action. As discussed in Section 9.6.2.2, below, and provided in 10 CFR 51.33, the

NRC may make a determination to prepare a draft FONSI. As provided in 10 CFR 51.25, an

EA is not necessary if an EIS will be prepared.

9.6.2.2 FONSI

When the staff makes a final finding that there are no significant environmental impacts from

the proposed action, a final FONSI will be published in the Federal Register. The Commission

will not take the proposed action, including license issuance, renewal, or amendment, until after

the FONSI has been published. Requirements for preparing a FONSI for materials licensing

actions are contained in 10 CFR 51.32 - 51.35. A FONSI will include the following elements:

(1) identification of the proposed action

(2) statement that the Commission has determined not to prepare an EIS for the proposed

action

(3) brief presentation of the reasons why the proposed action will not have a significant impact

on the quality of the human environment

(4) the EA or a summary of the EA

(5) a note of any other related environmental documents
SRP- Environmental Protection 9-20 NUREG-1520

(6) a statement that the finding and any related environmental documents are available for

public inspection and where the documents may be inspected

The NRC Branch Chief may make a determination to prepare and issue a draft FONSI for

public review and comment before making a final determination on whether to prepare an EIS

or a final FONSI on the proposed action. A draft FONSI may be prepared if a FONSI appears

warranted, but the proposed action is similar to one that normally requires an EIS or is without

precedent.

The draft FONSI will be identified as a “draft” and will contain the information specified above

for a final FONSI. The draft FONSI will be accompanied by or will include a request for

comments on the proposed action and the draft findings within 30 days, or a longer period may

be specified in the notice of the draft findings. This draft FONSI will be published in the Federal

Register, distributed as provided in 10 CFR 51.74(a), and made available in accordance with 10

CFR 51.123.

When a draft FONSI is issued, a final determination to prepare an EIS or final FONSI will not be

made until the last day of the public comment period has expired.

9.6.2.3 EIS

When determining that an EIS will be prepared for the licensing action, the NRC Branch Chief

should initiate coordination with the Environmental Review Team, which will assume

responsibility for preparing the EIS. The project manager should coordinate with Environmental

Review Team to (1) ensure consistency between the environmental review and the preparation

of the EIS and (2) ensure that the results of the NEPA analysis are appropriately incorporated

into the SER and license.

9.6.3 Safety Evaluation

The following language would be appropriate for a licensing action that required an EIS in

accordance with 10 CFR 51.20:

The applicant has committed to adequate environmental protection measures,

including (1) environmental and effluent monitoring and (2) effluent controls to

maintain public doses ALARA as part of the radiation protection program. The NRC

staff concludes, with reasonable assurance, that the applicant’s conformance to the

application and license conditions is adequate to protect the environment and the

health and safety of the public and comply with the regulatory requirements imposed

by the Commission in 10 CFR Parts 20, 51, and 70. The bases for these conclusions

are as follows:

[State the bases for the conclusion, including any recommended license conditions.]

The NRC staff prepared an environmental impact statement (EIS) [publication date]

for this licensing action as required by 10 CFR 51.20. On the basis of the EIS, the

NRC stated in its Record of Decision [publication date in the Federal Register] that the

preferred option was [state preferred option here]. 
SRP- Environmental Protection 9-21 NUREG-1520

9.7 REFERENCES

American National Standards Institute, “Guide to Sampling Airborne Radioactive Materials in

Nuclear Facilities, ANSI N13.1-1982.”

American National Standards Institute, “Specification and Performance of On site

Instrumentation for Continuously Monitoring Radioactive Effluents,” ANSI N42.18-1980.

National Council on Radiation Protection and Measurements, “Screening Models for Releases

of Radionuclides to Atmosphere, Surface Water, and Ground,” NCRP Reports No. 123 I & II, 

January 1996.

U.S. Environmental Protection Agency, “Limiting Values of Radionuclide Intake And Air

Concentration and Dose Conversion Factors For Inhalation, Submersion And Ingestion,”

Federal Guidance Report No. 11, September 1988.

U.S. Nuclear Regulatory Commission, “Guidance to Hazardous, Radioactive and Mixed Waste

Generators on the Elements of a Waste Minimization Program,” Information Notice No. 94-23,

March 25, 1994.

U.S. Nuclear Regulatory Commission, “Solubility Criteria for Liquid Effluent Releases to

Sanitary Sewerage Under the Revised 10 CFR Part 20,” Information Notice 94-07, 

January 28, 1994.

U.S. Nuclear Regulatory Commission, NMSS/FCSS/Fuel Cycle Licensing Branch, “Materials

Licensing Procedures Manual,” Rev. 5, September 1996. 

U.S. Nuclear Regulatory Commission, “Quality Assurance for Radiological Monitoring Programs

(Normal Operations)!Effluent Streams and the Environment,” Regulatory Guide 4.15, Rev. 2,

February 1979.

U.S. Nuclear Regulatory Commission, ”Monitoring and Reporting Radioactivity in Releases of

Radioactive Materials in Liquid and Gaseous Effluents from Nuclear Fuel Processing and

Fabrication Plants and Uranium Hexafluoride Production Plants,” Regulatory Guide 4.16, 

Rev. 2, December 1985. 

U.S. Nuclear Regulatory Commission, “Constraint on Releases of Airborne Radioactive

Materials to the Environment for Licensees other Than Power Reactors,” Regulatory Guide

4.20, December 1996.

U.S. Nuclear Regulatory Commission, “ALARA Levels for Effluents from Materials Facilities,”

Regulatory Guide 8.37, July 1993.

Accession #: ML013370390
Decommissioning 10-1 NUREG-1520

10 DECOMMISSIONING

10.1 PURPOSE OF REVIEW

The purpose of the review of the applicant’s decommissioning plans is to determine with

reasonable assurance that the applicant will be able to decommission the facility safely and in

accordance with NRC requirements.

At the time of the initial license application and again at license renewal, the applicant/licensee

may be required to submit a decommissioning funding plan (DFP). The purpose of the NRC’s

evaluation of the DFP is to determine whether the applicant/licensee has considered

decommissioning activities that may be needed in the future, has performed a credible sitespecific cost estimate for those activities, and has presented the NRC with financial assurance

to cover the cost of those activities in the future. The DFP, therefore, should contain an

overview of the proposed decommissioning activities, the methods used to determine the cost

estimate, and the financial assurance mechanism. This overview must contain sufficient detail

to enable the reviewer to determine whether the decommissioning cost estimate is reasonably

accurate.

If required by 10 CFR 70.38(g), the licensee must also submit, for NRC approval a

decommissioning plan (DP) before beginning its decommissioning actions. The DP must detail

the specific decommissioning activities to be performed, and must describe the radiation

protection procedures that will be used to protect workers, the public, and the environment

during decommissioning. This information must be sufficient to the reviewer to assess the

appropriateness of the decommissioning activities and the adequacy of the procedures to

protect the health and safety of workers, the public, and the environment. It must also update

the cost estimate originally presented in the DFP to undertake the facility decommissioning. 

Approval of a DP is often obtained through application for a license amendment. The reviewer

must ascertain that the applicant understands the decommissioning requirements and

procedures, and commits to the protection of the health and safety of workers, the public and

the environment during decommissioning. 

10.2 RESPONSIBILITY FOR REVIEW

Primary: Licensing Project Manager

Secondary: Environmental Reviewer

Technical and Financial Specialists in the Division of Waste Management

Supporting: Fuel Facility Inspection Staff

10.3 AREAS OF REVIEW

The reviewer will evaluate the applicant’s DFP and/or DP in accordance with the “NMSS

Decommissioning Standard Review Plan” NUREG-1727.

Before beginning to review starting the DFP or DP, the reviewer should first evaluate the

applicant’s proposed “Environmental Protection Measures” (SRP Chapter 9) and, specifically,

the commitments to minimize waste associated with decommissioning, as well as the “Radiation

Protection Program” (SRP Chapter 4) as it applies to radiological decontamination and

management of radiological effluents.
Decommissioning 10-2 NUREG-1520

10.4 ACCEPTANCE CRITERIA

10.4.1 Regulatory Requirements

The following NRC regulations require planning, financial assurance and recordkeeping for

decommissioning, as well as procedures and activities to minimize waste and contamination:

! 10 CFR 70.22(a)(9) “Decommissioning Funding Plan”

! 10 CFR 70.25 “Financial Assurance and Recordkeeping for 

Decommissioning”

! 10 CFR 70.38 “Expiration and Termination of Licenses and 

Decommissioning of Sites and Separate Buildings 

or Outdoor Areas”

! 10 CFR 20.1401-1406 “Radiological Criteria for License Termination”

(Subpart E)

10.4.2 Regulatory Guidance

The “NMSS Decommissioning Standard Review Plan,” NUREG-1727, defines relevant

regulatory guidance and appropriate acceptance criteria for DFPs and DPs contained in license

applications and/or amendment requests.

10.5 REVIEW PROCEDURES

The primary reviewer will evaluate the application against the NRC requirements and

acceptance criteria identified in the “NMSS Decommissioning Standard Review Program.” This

review will be supplemented (as appropriate) by a detailed review of any contamination and

waste minimization plans submitted by the applicant in response to 10 CFR 20.1406. The

reviewer will also coordinate with the principal reviewers for environmental protection listed in

(Chapter 9), to confirm the review of a new applicant’s plans to minimize for waste, as well as

plans for existing licensees to minimize contamination and reduce exposures and effluents as

part of the radiation protection program established under 10 CFR Part 20. The purpose of this

coordination is to ensure that any issues that are relevant to the environmental review are

properly conveyed to the primary reviewers for consideration and resolution as part of the

review discussed in Chapter 9. Similarly, any decommissioning issues that arise in the

environmental review that are best suited for review using guidance in this chapter are

conveyed to the primary reviewer for consideration and resolution.

If the decommissioning review identifies the need for the applicant to submit information that

has not already been included in the application, the reviewer will document these additional

information needs in a request for additional information (RAI). The RAI transmitted to the

applicant will specify a reasonable amount of time (e.g., 30 to 60 days) for the applicant to

reply. Failure of the applicant to provide the requested information by the specified date, or on

an alternative schedule that is mutually agreeable, could be grounds for terminating or

suspending the application review.

In accordance with the Fuel Cycle and International Safeguards Branch licensing manual, the

primary reviewer will coordinate with the Division of Waste Management to obtain appropriate

technical assistance in reviewing proposed DPs and financial assurance measures. The
Decommissioning 10-3 NUREG-1520

primary reviewer will coordinate with reviewers assigned by the Division of Waste Management 

incorporate, as appropriate, RAIs and review findings in licensing correspondence and SERs

related to decommissioning.

10.5.2 Safety Review

The reviewer should perform a safety analysis against the acceptance criteria in the “NMSS

Decommissioning Standard Review Plan,” NUREG-1727, to ensure that the proposed

decommissioning methodology, principal remediation activities, and worker and environmental

radiation protection programs are acceptable. 

10.6 EVALUATION FINDINGS

If the applicant provides sufficient information to satisfy the acceptance criteria and

requirements identified in Section 10.4, the staff will conclude that the DFP or DP evaluation is

complete and satisfactory. The primary reviewer will prepare an SER for the Licensing Project

Manager, in support of the licensing action. This SER should address each topic area

reviewed, and including an explanation of the bases for the reviewers’ conclusions, why the

NRC has reasonable assurance that the DFP or DP should be considered acceptable. The

SER may also include license conditions where the application is deficient. The SER should

include a summary statement of what was evaluated and. The staff will document its evaluation

as follows:

The NRC staff has evaluated the applicant’s/licensee’s plans and financial

assurance for decommissioning in accordance with the” NMSS Decommissioning

Program Standard Review Plan,” NUREG-1727. On the basis of this evaluation, the

NRC staff has determined that the applicant’s/licensee’s plans and financial

assurance for decommissioning comply with the NRC’s regulations, and provide

reasonable assurance of protection for workers, the public, and the environment.

10.7 REFERENCES

Code of Federal Regulations, Title 10, Part 70, “Domestic Licensing of Special Nuclear

Material.” 

Orlando, D.A., et al., “NMSS Handbook for Decommissioning Fuel Cycle and Materials

Licensees,” NUREG/BR-0241, U.S. Nuclear Regulatory Commission, 1997.

U.S. Nuclear Regulatory Commission, “NMSS Decommissioning Standard Review Plan,”

NUREG-1727,September 2000.

Accession #: ML013370403
Management Measures 11-1 NUREG-1520

11 MANAGEMENT MEASURES

11.1 PURPOSE OF REVIEW

Management measures are functions, performed by a licensee, generally on a continuing basis,

that are applied to items relied on for safety (IROFS) to provide reasonable assurance that the

IROFS are available and able to perform their functions when needed. The phrase “available

and reliable,” as used in 10 CFR Part 70, means that, based on the analyzed, credible

conditions in the ISA, IROFS will perform their intended safety function when needed to prevent

accidents or mitigate the consequences of accidents to an acceptable level. Management

measures will be implemented to provide reasonable assurance of compliance with the

performance requirements, considering factors such as necessary maintenance, operating

limits, common-cause failures, and the likelihood and consequences of failure or degradation of

the IROFS and the measures. The following discussion addresses each of the management

measures included in the 10 CFR Part 70 definition of management measures, i.e.,

configuration management (CM), maintenance, training and qualifications, procedures, audits

and assessments, incident investigations, records management, and other quality assurance

(QA) elements. The degree to which measures are applied to the IROFS may be a function of

the item’s importance in terms of meeting the performance requirements as evaluated in 

the ISA.

The applicant’s descriptions of management measures should address in sufficient detail how

the measure is designed, organized, and conducted to enable the reviewer to understand the

capability of the measure to be implemented at the facility. If a “graded” application of a

particular management measure is to be used for IROFS of differing importance to risk

management, then the variations should be described. 

Examples of the kind of information that the reviewer(s) require to assess the adequacy of a

management measure is provided in Appendix B to this SRP chapter. The purpose of this

review is to enable the staff to conclude, with reasonable assurance, that the management

measures applied to IROFS, as documented in the ISA Summary, provide reasonable

assurance that the IROFS will be available and able to perform their functions, when needed,

consistent with the performance requirements of 10 CFR 70.61. If a graded approach is used,

the review should also determine whether the measures are applied to the IROFS

commensurate with the IROFS’ importance to safety. 

11.2 RESPONSIBILITY FOR REVIEW

Primary: Licensing Project Manager

Secondary:

Configuration Management: Primary ISA Summary Reviewer, QA and Records

Management Reviewers

Maintenance: Criticality, Chemical, Fire, Radiation Protection, and

Environmental Reviewers 

Training and Qualification: Training Specialist, QA Reviewer

Procedures: Radiation Protection, Criticality and Fire Protection

Engineers, Fuel Cycle Facility Inspector 

Audits and Assessments: QA Reviewer

Incident Investigations: Inspection Specialist

Records Management: QA Reviewer 
Management Measures 11-2 NUREG-1520

QA: Quality Assurance Engineer

Supporting: Technical Discipline Engineers, Fuel Cycle Facility

Inspectors, Resident Inspectors

11.3 AREAS OF REVIEW

11.3.1 Configuration Management (CM)

This review should provide reasonable assurance that the applicant has committed to develop

and implement a CM function that is consistent with the requirements of 10 CFR 70.72(a). The

review should determine, with reasonable assurance, that the applicant has described and

committed to a CM function that assures consistency among the facility design and operational

requirements, the physical configuration, and the facility documentation. The review should

also determine that the applicant’s CM function captures formal documentation governing the

design and continued modification of the site structures, processes, systems, equipment,

components, computer programs, personnel activities, and supporting management measures,

as identified and described in the ISA Summary. The review should assure that the CM

function is adequately coordinated and integrated with the other management measures.

The NRC staff should review the applicant's descriptions and commitments for CM, including

descriptions of the organizational structure responsible for CM activities; descriptions of the

process, procedures, and documentation required by the applicant for modifying the site; and

descriptions of the various levels of CM to be applied to IROFS designated in the ISA

Summary. The staff review should focus on the applicant’s CM measures that provide

reasonable assurance of the disciplined documentation of engineering, installation, and

operation of modifications; the training and qualification of affected staff; the revision and

distribution of operating, test, calibration, surveillance, and maintenance procedures and

drawings; post-modification testing; and readiness review.

The NRC staff should review the following:

1. CM Policy

The review should cover the applicant's description of overall CM functions, including at

least the following topics: (a) the scope of the IROFS to be included in the CM function,

(b) objectives of each CM activity, (c) a description of each CM activity, and (d) the

organizational structure and staffing interfaces. 

The review should examine the applicant’s establishment of a CM policy applicable to all

operations, in accordance with 10 CFR 70.72.

2. Design Requirements

The reviewer should examine the applicant’s descriptions of how design requirements and

associated design bases have been established and are maintained. The reviewer should

evaluate the applicant’s CM controls on the design requirements and for the ISA Summary.
Management Measures 11-3 NUREG-1520

3. Document Control

The reviewer should examine the applicant’s description of its methods for controlling

documents within the CM function. 

4. Change Control

The review should examine the applicant’s commitments to provide reasonable assurance

that the CM function maintains consistency among the design requirements, the physical

configuration, and the facility documentation, in accordance with 10 CFR 70.72, “Facility

changes and change process.” 

5. Assessments

The review should examine the applicant’s commitments to conduct initial and periodic

assessments of the CM function, to determine the function’s effectiveness, and to correct

deficiencies, consistent with the acceptance criteria for “Audits and Assessments.”

6. Design Reconstitution (Existing Facilities Only)

The review should examine the applicant’s discussion of design reconstitution of the current

design basis that has been done for the purpose of the application, and how that

reconstitution was translated into a fixed baseline design basis against which subsequent

changes will be measured. 

11.3.2 Maintenance

The NRC staff will evaluate the applicant’s description of its maintenance function. The

reviewer will examine the applicant’s commitments to inspect, calibrate, test, and maintain

IROFS to a level commensurate with the items’ importance to safety to provide reasonable

assurance of their ability to perform their safety functions when required. The applicant

identifies these IROFS in the ISA Summary. The staff will review the applicant’s description of

how each of the following functions is implemented within the site organization. Note that not

every aspect of each of the four maintenance functions is necessarily required. The applicant

should justify the assignment of differing degrees of maintenance to individual IROFS, based

on the item’s contribution to the reduction of risk.

1. Corrective maintenance

a. A commitment to promptly perform corrective actions to remediate unacceptable

performance deficiencies in IROFS.

b. A description of the approach and methods for planning and implementing repairs to

IROFS with the objective of eliminating or minimizing the recurrence of unacceptable

performance deficiencies.

2. Preventive maintenance (PM)

a. A commitment to conduct preplanned and scheduled periodic refurbishing and/or

overhauls of IROFS.
Management Measures 11-4 NUREG-1520

b. A description of PM activities, including, for example, instrumentation calibration and

testing and the methods used to establish the frequency of PM activities.

3. Surveillance/monitoring

a. A commitment to design and implement a program to survey and monitor the

performance of IROFS.

b. A description of the components of the surveillance and monitoring program including

methods used to establish the frequency of such inspections for IROFS having different

degrees of safety importance.

4. Functional testing

a. A commitment to perform the appropriate post-maintenance functional testing to provide

reasonable assurance that the maintenance activity did not adversely affect the

reliability of the IROFS.

b. A general description of functional testing and the test results documentation.

11.3.3 Training and Qualifications

10 CFR Part 70 requires that all personnel who perform activities relied on for safety be trained

and tested, as necessary, to provide reasonable assurance that they understand, recognize the

importance of, and are qualified to perform these activities in a manner that adequately protects

(1) the health and safety of the public and workers and (2) the environment. As appropriate for

their authority and responsibilities, these personnel should have the knowledge and skills

necessary to design, operate, and maintain the facility in a safe manner. Therefore, the

training, testing, and qualification of these personnel who perform activities relied on for safety

should be described in the application and should be reviewed by the staff. The review should

examine the applicant’s experience and capabilities to provide this required training for its

personnel who will perform activities relied on for safety. The review of the training and

qualification should address the following training areas:

1. Organization and management of the training function

2. Analysis and identification of functional areas requiring training 

3. Position training requirements

4. Development of the basis for training, including objectives 

5. Organization of instruction, using lesson plans and other training guides 

6. Evaluation of trainee learning

7. Conduct of on-the-job training

8. Evaluation of training effectiveness

9. Personnel qualification

10. Applicant’s provisions for continuing assurance, including the needs for retraining or

reevaluation of qualification

11.3.4 Procedures

The review should examine the applicant’s process for the preparation, use, and management

control of written procedures. This should include the basic elements of identification,

development, verification, review and comment resolution, approval, validation, issuance,
Management Measures 11-5 NUREG-1520

change control, and periodic review. The applicant should prepare two general types of

procedures for use at the facility: 

1. Procedures used to directly control process operations, commonly called "operating

procedures." These are procedures for workstation operators and should include directions

for normal operations as well as off-normal events caused by human error or failure of an

IROFS. Procedures of this type include required actions to provide reasonable assurance

of nuclear criticality safety, chemical safety, fire protection, emergency planning, and

environmental protection.

2. Procedures used for activities that support the process operations, which are commonly

referred to as "management control procedures." These are procedures used to manage

the conduct of activities such as CM, radiation safety, maintenance, QA, training and

qualification, audits and assessments, incident investigations, record-keeping, and

reporting.

The actual operating procedures are not part of the license and would not normally be reviewed

for technical adequacy for low-risk processes, since this aspect is addressed by the inspection

function. For new licenses or processes, especially those that involve high-risk operations,

such as some highly enriched uranium liquid processes or some mixed-oxide processes, the

licensing review may require a site visit to make an adequate safety determination, at which

time some procedures may be reviewed. 

The NRC staff should review the commitments in the application to provide reasonable

assurance that the applicant’s program adequately addresses the following:

1. The method for identifying procedures that are needed plant-wide. The ISA Summary

identifies IROFS where human actions are important. Procedures should be provided for all

necessary steps or operations that are performed at the facility. Procedures should be

provided for every element of management control that is discussed in the SRP sections.

2. Essential elements that are generic to all procedures including criticality, chemical process

and fire safety, warning notes, reminders or pertinent information regarding specific hazards

or concerns (including station limits), Materials Safety Data Sheet availability, special

precautions, radiation and explosive hazards, and special personal protective equipment.

3. The method for creating and controlling procedures within plant management control

systems. This includes how procedures are managed within the facility CM function. 

4. The method for verifying and validating procedures before use. During procedure

development, workers and operators review procedures to provide assurance that they are

usable and accurate.

5. The method and schedule for periodically reverifying and revalidating procedures.

6. The method for ensuring that current procedures are available to personnel and that

personnel are qualified to use the latest procedures.
Management Measures 11-6 NUREG-1520

11.3.5 Audits and Assessments

The applicant should describe a system of audits and assessments that consists of two distinct

levels of activities: an audit activity structured to monitor compliance with regulatory

requirements and license commitments, and an assessment activity oriented to determining the

effectiveness of the activities in achieving applicant-specified objectives that provide reasonable

assurance of the continued availability and reliability of IROFS. An applicant may describe a

“corrective action program” which includes the functions of incident investigations as well as

audits and assessment (see Section 11.3.6). This approach is acceptable and the reviewer

should, in that case, review the applicant’s description and commitments with regard to the

acceptance criteria in this SRP chapter for incident investigations as well as audits 

and assessment.

The reviewer should examine the applicant's presentation with respect to:

1. The commitments to audit and assessment activities

2. The use of qualified and independent audit and assessment personnel

3. The general structure of typical audits and assessments

4. The facility procedures to be used to direct and control the audit and assessment activities

5. The planned use of the results of the audit and assessment activities

6. The documentation to record and distribute the findings and recommendations of these 

audits and assessments

7. The planning and implementation of corrective actions based on the findings and

recommendations 

11.3.6 Incident Investigations

The NRC staff should review the applicant's policy, procedures, and management structure for

investigating abnormal events and completing appropriate corrective actions. The review

should include the provisions for establishing investigating teams, the methods for determining

root causes, and the procedures for tracking and completing corrective actions and for

documenting the process for the purpose of applying the "lessons learned" to other operations. 

An applicant may describe a “corrective action program” which includes the functions of audits

and assessment as well as incident investigations. This approach is acceptable and the

reviewer should, in that case, review the applicant’s description and commitments with regard to

the acceptance criteria in this SRP chapter for audits and assessments as well as 

incident investigations.

11.3.7 Records Management

The requirements for the management of records vary according to the nature of the facility and

the hazards and risks posed by it. The staff should review areas related to the handling and

storing of health and safety records and the records generated or needed in the design,
Management Measures 11-7 NUREG-1520

construction, operation, and decommissioning phases of the facility. The staff should review

the following:

1. The process whereby records - training records, dosimetry records, effluents records,

records of classified information, records concerning facility IROFS, and records of their

failure - are created, selected, verified, categorized, indexed, inventoried, protected, stored,

maintained, distributed, deleted, or preserved. The review should provide reasonable

assurance that the records management function is adequately coordinated and integrated

with other management measures. 

2. The handling and control of various kinds of records (including contaminated and classified

records) and the media in which the records are captured.

3. The physical characteristics of the records storage area(s) with respect to the preservation

and protection of the records for their designated lifetimes.

11.3.8 Other QA Elements

The application must address other QA elements that will be applied to IROFS and other

management measures. The review should determine that a complete description of the

applicant’s application of QA elements to IROFS is included in the application. The review

objective is to obtain reasonable assurance of the implementation of accepted QA principles in

the design, construction, operation, maintenance, and modification phases of a facility's life. 

Fundamental to this effort is the applicant's application of QA elements to the identified IROFS

resulting from the ISA and identified in the ISA Summary. QA elements would also be

applicable, as appropriate, to the hazards analysis process in the applicant’s ISA.

The application defines the QA elements and the levels to be applied to IROFS identified in the

ISA Summary (SRP Chapter 3). Further, the manner in which the QA function is coordinated

and integrated with other management measures should be described. If the applicant applied

a graded safety program to the IROFS, the application should explain how the QA elements

were also applied in a graded manner to the IROFS. 

The reviewer(s) should recognize that facility safety may not be the only area at a fuel cycle

facility requiring QA elements. The applicant's customers and the NRC, under 10 CFR Part 50,

may impose product-related QA criteria. The focus of the review of QA measures per this SRP

is limited to ensuring the safety of workers and the public and protecting the environment (i.e.,

in relation to the performance requirements of 10 CFR 70.61).

Since many QA elements may be described in other sections of the application, the reviewer

should determine the applicant's commitment to overall QA, the selection of quality criteria and

quality level, and the proposed method for implementation. The applicant may reference other

areas of the application that present information relevant to QA. The reviewer should focus on

the management measures applied to criticality, containment of licensed materials, personnel

protection, and environmental safety. Application of graded QA and quality levels

commensurate with the risk involved should parallel the same risk levels established for

maintenance and other management measures.
Management Measures 11-8 NUREG-1520

11.4 ACCEPTANCE CRITERIA

The reviewer should find the applicant’s information acceptable if it provides reasonable

assurance that the following acceptance criteria are satisfactorily addressed.

11.4.1 Regulatory Requirements

The requirements for fuel cycle facility management measures are specified in 10 CFR Part 70,

“Domestic Licensing of Special Nuclear Material," as revised.

10 CFR 70.4 states that management measures include CM, maintenance, training and

qualifications, procedures, audits and assessments, incident investigations, records

management, and other QA elements.

10 CFR 70.62(a)(3) states that records must be kept for all IROFS failures, describes required

data to be reported, and sets time requirements for updating the records.

10 CFR 70.62(d) requires an applicant to establish management measures, for application to

engineered and administrative controls and control systems that are identified as IROFS,

pursuant to 10 CFR 70.61(e), so they are available and reliable to perform their functions when

needed.

A regulation specifically applicable to personnel training and qualification is 10 CFR Part 19,

“Notices, Instructions and Reports to Workers: Inspection and Investigations,” Section 19.12,

"Instructions to Workers."

The regulatory requirement for procedures that protect health and minimize danger to life is

specified in 10 CFR 70.22(a)(8).

Facility change processes are required to conform with 10 CFR 70.72.

Incident investigation and reporting are required by 10 CFR 70.74(a) and (b).

11.4.2 Regulatory Guidance

• American Society of Mechanical Engineers standard, “Quality Assurance Requirements for

Nuclear Facility Applications,” ANSI/ASME NQA -1, 1994.

• American National Standards Institute Standards for Quality Management, ANSI/ISO/ASQ

9000 series.

• International Atomic Energy Agency Safety Guide, “Establishing and Implementing a Quality

Assurance Program,” Safety Guide 50-SG-Q1, 1995.

• U.S. Department of Energy, Draft, “Implementation Guide for Use with 10 CFR Part 830.120

and DOE Order 5700.6C,” September 1997. 

• U.S. Nuclear Regulatory Commission, “Guidance on Management Controls/Quality

Assurance, Requirements for Operation, Chemical Safety, and Fire Protection for Fuel Cycle

Facilities,” Federal Register 54 (No. 53), 11590!11598, March 21, 1989.
Management Measures 11-9 NUREG-1520

• U.S. Nuclear Regulatory Commission, “Training Review Criteria and Procedures,” 

NUREG-1220, Revision 1, January 1993.

11.4.3 Regulatory Acceptance Criteria

11.4.3.1 Configuration Management (CM)

1. CM Policy

The applicant's description of overall CM functions covers at least the following topics: 

(a) the scope of the IROFS and management measures to be included in the CM function

(coordinate with the reviewer of Chapter 3 of this SRP), (b) the objectives of each CM

activity, (c) a description of each CM activity, and (d) the organizational structure and

staffing interfaces. The functional interfaces with maintenance and training and qualification

are of particular importance and should be addressed individually. The IROFS under CM

should include all IROFS listed in the ISA Summary.

An important element of an applicant’s overall CM policy is the establishment of a baseline

CM policy applicable to all new facilities or new processes at existing facilities, in

accordance with 10 CFR 70.64. That baseline initially includes all the CM functions

described in this SRP chapter. After an ISA is completed and IROFS are identified that may

not be associated with high or intermediate consequence accident sequences, as defined

by the ISA Summary, the applicant may choose to reduce or eliminate certain features of

the CM function as applied to those lesser-risk design or operational features. In that case,

in describing its CM policy, the applicant defines the specific attributes of the reduced level

or levels of CM that would be applied to selected IROFS, and in the ISA identifies those

items that will be assigned the lesser level or levels of CM.

The design process leading to drawings and other statements of requirements proceeds

logically from the design basis. IROFS to be listed under CM are clearly defined in the ISA

Summary, along with the assignment of any grades or quality levels. The applicant should

have indicated in the ISA Summary what level of CM attributes is applied to a particular

IROFS. However, in the ISA Summary, this indication may only consist of an index or

category designation. The definitions of the multiple CM levels, if used, should be in the CM

description in the application. 

2. Design Requirements

The applicant describes how design requirements and associated design bases are

established and are maintained through control of the design process. Technical

management review and approval functions are described. 

3. Document Control

The applicant describes an acceptable method to create and control documents within the

CM function, including cataloging the document database, the information content of the

document database, maintaining and distributing documents, document retention policies,

and document retrieval policies. The applicant describes how CM will capture documents
Management Measures 11-10 NUREG-1520

that are relevant and relied on for safety. The description includes design requirements, the

ISA, as-built drawings, specifications, all procedures that are IROFS, procedures involving

training, QA, maintenance, audits and assessments, emergency operating procedures,

emergency response plans, system modification documents, assessment reports, and

others that the applicant may deem part of CM. The document database is used to control

documents and track document change status. 

4. Change Control

The applicant describes how the CM function will maintain strict consistency among the

design requirements, the physical configuration, and the facility documentation. The

applicant commits to an acceptable process for identifying and authorizing proposed

changes; for performing appropriate technical, management, and safety reviews of

proposed changes in IROFS; for tracking and implementing changes; and for documenting

changes (including placement of documentation in a document control center and

dissemination to affected functions such as training, engineering, operations, maintenance,

and QA). The applicant also describes an acceptable process, within the CM function, for

providing reasonable assurance that the ISA is systematically reviewed and modified to

reflect design or operational changes from an established safety basis, and that all

documents outside the ISA that are affected by safety basis changes are properly modified,

authoritatively approved, and made available to personnel. When a change is made in

accordance with 10 CFR 70.72, changes to the affected onsite documentation must be

made promptly to avoid inadvertent access by facility personnel to outdated design and

other specifications for IROFS.

5. Assessments

The applicant confirms that initial and periodic assessments of the CM function are

conducted to determine the program's effectiveness and to correct deficiencies. Both

document assessments and physical assessments (system walkdowns) will be conducted

periodically to check the adequacy of the CM function. All assessments and followups are

documented. These reports can provide a basis for future changes. The applicant

indicates that such assessments are systematically planned and conducted in accordance

with an overall facility audit and assessment function (see the sections in Chapter 11 for

details on audits and assessments).

 6. Design Reconstitution (Existing Facilities Only)

The applicant describes whatever design reconstitution has been done for the purpose of

the application. Because this information may duplicate the facility design bases

information described elsewhere to support the ISA, this information may be included by

reference to other parts of the application. The applicant has available current design

bases, including design requirements, supporting analyses, and documentation supporting

all IROFS. A verification process, including walkdowns, is complete and has verified that

the configuration is consistent with as-built facility documentation. 
Management Measures 11-11 NUREG-1520

11.4.3.2 Maintenance

The reviewers should find the applicant’s submittal acceptable if the application includes

the following:

1. Surveillance/monitoring

For IROFS identified in the ISA Summary, the applicant describes the surveillance function

and its commitment to the organization and conduct of surveillance at a specified frequency. 

The surveillance activity should support the determination of performance trends for IROFS,

thus providing data useful in determining PM frequencies. The applicant describes how the

results from incident investigations, the review of the failure records required by 

10 CFR 70.62(a)(3), and identified root causes are used to modify the affected maintenance

function and eliminate or minimize the root cause. Records showing the current

surveillance schedule, performance criteria, and test results for all IROFS are maintained by

the applicant. For surveillance tests that can only be done while IROFS are out of service,

proper compensatory measures are prescribed for the continued normal operation of 

a process.

2. Corrective maintenance

The applicant provides the documented approach used to perform corrective actions or

repairs on IROFS. The maintenance function provides a planned, systematic, integrated,

and controlled approach for the repair and replacement activities associated with identified

unacceptable performance deficiencies of IROFS. After conducting corrective maintenance

and before returning an IROFS to operational status, if necessary, a functional test is

conducted to provide reasonable assurance that the safety control performs as designed

and provides the safety action expected. 

3. PM

The applicant provides a description of the PM function that demonstrates a commitment to

conduct preplanned and scheduled periodic refurbishing, or partial or complete overhaul, for

the purpose of ensuring that unanticipated loss of IROFS do not occur. This activity

includes using the results of the surveillance component of maintenance and the failure

records required by 70.62(a)(3). Instrumentation calibration and testing are addressed by

the applicant as part of this component. The applicant describes how the function will be

designed to assure that the objective of preventing failures through maintenance is

appropriately balanced against the objective of minimizing unavailability of IROFS because

of monitoring or PM. After conducting PM and before returning a safety control to

operational status, if necessary, a functional test is conducted to ensure that an IROFS

performs as designed and provides the safety action expected. The methodology or basis

used to determine PM frequency is described. The applicant describes how results from

incident investigations and identified root causes are used to modify the affected

maintenance function and eliminate or minimize the root cause from recurring. Feedback

from PM, corrective maintenance, and incident investigations is used, as appropriate, to

modify the frequency or scope of the PM activity. A rationale for deviations from industry

standards or from vendor recommendations for PM is provided. Records showing the PM

schedule, and results, for all IROFS subject to this maintenance component are maintained

by the applicant.
Management Measures 11-12 NUREG-1520

4. Functional testing

The application includes a general description of the methods used and the commitment to

perform functional testing, as needed, of IROFS after PM or corrective maintenance. These

tests should be conducted using applicant-approved procedures and should include

compensatory measures while the test is being conducted. The applicant designs the

functional test to include all operational aspects of the IROFS that are important to safety. 

For illustrative purposes only, the following scenario is provided:

A level controller, identified as an IROFS, is used to actuate a three-way valve and

divert flow to an alternate tank. The level monitor sending unit and the valve, power

supplies, utility services, and any corresponding local or control room displays should be

tested at the same time during the functional test. The objective should be to simulate

actual upset conditions and demonstrate that the IROFS is available and reliable and

will function in the field as intended. 

As necessary, during startup of new process equipment, functional tests are conducted and

documented and the documents are maintained for NRC review. Records showing the

functional test schedule and results for all IROFS subject to this maintenance component are

maintained by the applicant.

Administrative controls are often identified as IROFS. The applicant should provide a general

discussion about how these IROFS are assured to be available and reliable to perform their

intended safety function over extended periods of operation. Specific management measures

and how they are applied should be described.

A general acceptance criterion applicable to all maintenance functions is an adequate

description of work control methods. Listed below are methods or practices that should be

applied to the corrective, preventive, and functional-test maintenance elements, and for which

the applicant should commit to prepare written procedures. These include, as applicable: 

(a) authorized work instructions with detailed steps and a reminder of the importance of the

IROFS identified in the ISA Summary, (b) parts lists, (c) as-built or redlined drawings, (d) a

notification step to the operations function before conducting repairs and removing an IROFS

from service, (e) radiation work permits, (f) replacement with like-kind parts and the control of

new or replacement parts to ensure compliance with 10 CFR Part 21, (g) compensatory

measures while performing work on IROFS, (h) procedural control of removal of components

from service for maintenance and for return to service, (i) ensuring safe operations during the

removal of IROFS from service, and (j) notification to operations personnel that repairs have

been completed. Written procedures for the performance of maintenance include steps a

through j. The details of maintenance procedure acceptance criteria are addressed in 

Section 11.4.3.2 of this SRP. All work requests and maintenance procedures include technical

and safety discipline reviews and approval.

As applicable, contractors that work on or near IROFS identified in the ISA Summary should be

required by the applicant to follow the same maintenance guidelines described for the

corrective, preventive, functional, or surveillance/monitoring activities listed above for the

maintenance function.
Management Measures 11-13 NUREG-1520

The four maintenance elements described above are covered by elements of the management

measures discussed in Chapter 11 of this SRP. The applicant should include a discussion of

how, or provide references to, the maintenance function uses, interfaces with, or is linked to the

various management measures. As an example, since maintenance workers are trained and

qualified to perform their duties, the link between maintenance and the training and qualification

function should be described.

11.4.3.3 Training and Qualification

The applicant’s submittal regarding personnel training and qualification should be acceptable if

it satisfies the following criteria. In addition to the regulatory acceptance criteria given below,

SRP Section 4.4.5.3 provides specific criteria for training and qualification for radiation safety

personnel. Similarly, some of the information specified below may be found in other sections of

the SRP and may be incorporated by reference.

1. Organization and Management of Training - The organization and management of training

are acceptable if the design, construction, operation, modification, maintenance and

decommissioning of the facility are organized, staffed, and managed to facilitate planning,

directing, evaluating, and controlling a training process that fulfills the objectives for the training

identified by the licensee, especially where human factors are relied on for safety. Formal

training should be provided for each position or activity that is relied on for safety. Training may

be either or both classroom or on-the-job training. The application should state what training

will be conducted and which personnel will be provided with this training. 

The following commitments should be in the application regarding organization and

management of training:

a. Line management is responsible for the content and effective conduct of the training. 

b. The job function, responsibility, authority, and accountability of personnel involved in

managing, supervising, and implementing training are clearly defined.

c. Performance-based training is used as the primary management tool for analyzing,

designing, developing, conducting, and evaluating training. 

d. Procedures are documented and implemented to provide reasonable assurance that all

phases of training are conducted reliably and consistently. 

e. Training documents are linked to the CM system to provide reasonable assurance that

design changes and modifications are accounted for in the training.

f. Exemptions from training are granted to trainees and incumbents only when justified,

documented, and approved by management.

g. Both programmatic and individual training records are maintained. These records

support management information needs and provide required data on each individual's

training, job performance, and qualification.

2. Analysis and Identification of Activities Requiring Training - The analysis and identification of

activities requiring training are acceptable if the activities required for competent and safe job

performance are identified, documented, and addressed by the training. 

Design, construction, operations, training, and other subject matter experts, as appropriate,

should conduct an analysis to identify activities requiring training. The activities treated in this

manner should include - as a minimum - those for managing, supervising, performing, and

verifying the activities relied on for safety specified in the ISA Summary as preventing or

mitigating accident sequences. Each activity selected for training (initial or continuing) from the
Management Measures 11-14 NUREG-1520

facility-specific activities should be matrixed to supporting procedures and training materials. 

The facility-specific activities selected for training and the comparison with training materials

should be reviewed on an established schedule and updated as necessitated by changes in

procedures, facility systems/equipment, or job scope.

3. Position Training Requirements - The position training requirements are acceptable if

minimum requirements for positions are specified for candidates whose activities are relied on

for safety or who perform actions that prevent or mitigate accident sequences described in the

ISA Summary. Trainees should meet entry-level criteria defined for the position, including

minimum educational, technical, experience, and physical fitness (if necessary) requirements. 

4. Development of the Basis for Training, Including Objectives - The development of the basis

for training, including the objectives, is acceptable if the basis identifies training content, defines

satisfactory trainee performance and identifies objectives from the analysis of activities and

performance requirements. The objectives should state the knowledge, skills, and abilities the

trainee should acquire; the conditions under which required actions will take place; and the

standards of performance the trainee should achieve on completion of the training activity.

5. Organization of Instruction Using Lesson Plans and Other Training Guides - Lesson plans

and other training guides should provide guidance to assure the consistent conduct of training

activities, and should be based on required learning objectives derived from specific job

performance requirements. Plans or guides should be used for in-class training and on-the-job

training and should include standards for evaluating acceptable trainee performance. Review

and approval requirements should be established for all plans or guides and other training

materials before their issue and use.

6. Evaluation of Trainee Accomplishment - The evaluation of trainee accomplishment is

acceptable if trainees are evaluated periodically during training to determine their progress

toward full capability to perform the job requirements and, at the completion of training, to

determine their capability to perform the job requirements. 

7. Conduct of On-the-Job Training - The conduct of on-the-job training is acceptable if on-thejob training used for activities relied on for safety and listed in the ISA Summary is fully

described. On-the-job training should be conducted using well-organized and current training

materials. On-the-job training should be conducted by designated personnel who are

competent in the program standards and methods of conducting the training. Completion of

on-the-job training should be by actual task performance. When the actual task cannot be

performed and is, therefore, “walkeddown,” the conditions of task performance, references,

tools, and equipment should reflect the actual task to the extent possible.

8. Evaluation of Training Effectiveness - An evaluation of training effectiveness and its relation

to job performance is acceptable if it provides reasonable assurance that the training conveys

all required skills and knowledge and is used to revise the training, where necessary, based on

the performance of trained personnel in the job setting. A comprehensive evaluation of

individual training should be conducted periodically by qualified individuals to identify strengths

and weaknesses. Feedback from trainee performance during training and from former trainees

and their supervisors should be used to evaluate and refine the training. Change actions (for

example procedure changes, equipment changes, facility modifications) should be monitored

and evaluated for their impact on the development or modification of initial and continuing

training and should be incorporated in a timely manner. This should be accomplished with
Management Measures 11-15 NUREG-1520

document control through the CM function. Improvements and changes to initial and continuing

training should be initiated, evaluated, tracked, and incorporated to correct training deficiencies

and performance problems.

9. Personnel Qualification - Commitments should be provided regarding minimum qualifications

for personnel required to meet NRC regulations. Minimum qualifications should be

commensurate with the assigned functional responsibility and authority of the respective

personnel. The application should contain such commitments regarding personnel qualification

for managers, supervisors, designers, technical staff, construction personnel, facility operators,

technicians, maintenance personnel, and other staff required to meet NRC regulations: 

a. Managers should have a minimum of a B.S. or B.A. or the equivalent. Each manager

should have either management experience or technical experience in facilities similar

to the facility identified in the application. 

b. Supervisors should have at least the qualifications required of personnel being

supervised, plus, either 1 additional year of experience supervising the technical area at

a similar facility or completion of a supervisor training course.

c. Technical professional staff identified in the ISA Summary whose actions or judgments

are critical to satisfy the performance requirements identified in 10 CFR Part 70 (i.e.,

related to an IROFS) should have a B.S. in the appropriate technical field and 3 years of

experience. Other technical professional staff should have a B.S. in the appropriate

technical field and 1 year of experience.

d. Construction personnel, facility operators, technicians, maintenance personnel, and

other staff whose actions are required to comply with NRC regulations should have

completed the applicant’s training process or have equivalent experience or training.

e. Candidates for process operators should be required to meet the minimum qualifications

described in the application. Candidates for job functions other than process operators

should also be required to meet minimum qualifications, but these minimum

qualifications need not be described in the application. 

10. Applicant’s Provisions for Continuing Assurance - The applicant’s provisions for continuing

assurance of personnel training and qualification are acceptable if the submittal addresses

periodic requalification of personnel, by training and/or testing as necessary, to provide

reasonable assurance that they continue to understand, recognize the importance of, and be

qualified to perform their activities that are relied on for safety.

11.4.3.4 Procedures Development and Implementation

The reviewer should determine that the applicant's process for developing and implementing

procedures is acceptable if it satisfies the following:

1. Procedures are written or planned for the operation of IROFS and for all management

measures supporting those IROFS. 
Management Measures 11-16 NUREG-1520

2. Operating procedures contain the following elements: (a) purpose of the activity, 

(b) regulations, polices, and guidelines governing the procedure, (c) type of procedure,

(d) steps for each operating process phase, (e) initial startup, (f) normal operations, 

(g) temporary operations, (h) emergency shutdown, (I) emergency operations, (j) normal

shutdown, (k) startup following an emergency or extended downtime, (l) hazards and safety

considerations, (m) operating limits, (n) precautions necessary to prevent exposure to

hazardous chemicals (resulting from operations with SNM) or to licensed SNM, 

(o) measures to be taken if contact or exposure occurs, (p) IROFS associated with the

process and their functions, and (q) the timeframe for which the procedure is valid. It is

particularly important that safety limits and IROFS (such as mass limits, moderator

exclusion, and independent sampling requirements) be clearly identified as such in the

procedure for the operators.

3. Procedures reflect the important elements of the functions described in the applicable

chapters of this SRP. Procedures exist to direct the following activities: (a) design, (b) CM,

(c) procurement, (d) construction, (e) radiation safety, (f) maintenance, (g) QA elements, 

(h) training and qualification, (i) audits and assessments, (j) incident investigations, 

(k) records management, (l) criticality safety, (m) fire safety, (n) chemical process safety,

and (o) reporting requirements.

4. The applicant describes the method for identifying, developing, approving, implementing,

and controlling operating procedures. Identifying needed procedures includes consideration

of ISA results. The method ensures, as a minimum, that (a) operating limits and IROFS are

specified in the procedure, (b) procedures include required actions for off-normal conditions

of operation, as well as normal operations, (c) if needed, safety checkpoints are identified

at appropriate steps in the procedure, (d) procedures are validated through field tests, 

(e) procedures are approved by management personnel responsible and accountable for

the operation, (f) a mechanism is specified for revising and reissuing procedures in a

controlled manner, (g) the QA elements and CM functions at the facility provide reasonable

assurance that current procedures are available and used at all work locations, and (h) the

facility training program trains the required persons in the use of the latest 

procedures available.

5. The applicant includes the following commitment regarding procedure adherence: 

“Activities involving licensed SNM and/or IROFS will be conducted in accordance with

approved procedures.”

6. The applicant describes the types of procedures used during facility operation. These will

typically include management control, operating, maintenance, and emergency procedures. 

The applicant provides information regarding the procedure categories used at the facility. 

The applicant develops procedures for site-wide safe work practices to control processes

and operations with licensed SNM and/or IROFS and/or hazardous chemicals incident to

the processing of licensed material. These safe work practices apply to workers, visitors,

contractors, and vendors. An acceptable identification discussion clearly states areas for

which a procedure is required. Procedures are required for operator actions that are

necessary to prevent or mitigate accidents identified in the ISA Summary. The applicant

provides a listing (in an appendix) of the types of activities that are covered by written

procedures. The listing includes the topics of administrative procedures; system

procedures that address startup, operation, and shutdown; abnormal operation/alarm

response; maintenance activities that address system repair, calibration, inspection, and
Management Measures 11-17 NUREG-1520

testing; and emergency procedures. Appendix A to this SRP chapter provides an

acceptable listing of the items to be included under each topic.

7. The applicant reviews procedures after unusual incidents, such as an accident, unexpected

transient, significant operator error, or equipment malfunction, or after any modification to a

system, and revises procedures as needed.

8. The applicant verifies the technical accuracy of procedures and that they can be performed

as written. The discussion identifies who is responsible for verification. The verification

process provides reasonable assurance that the technical information, including formulas,

set points, and acceptance criteria, is all there and is correct, and includes either a

walkdown of the procedure in the field, or a tabletop walkthrough. The review process

includes technical, cross-disciplinary reviews by affected organizations. This process

includes both new procedures and procedure changes. The review provides reasonable

assurance that the operating limits and IROFS identified in the ISA Summary are specified

in the procedures and that QA requirements are identified and included in operating

procedures. The applicant specifies who can approve procedures and the approval level for

each procedure type. At a minimum, responsible management and the safety disciplines

approve new procedures and changes to existing procedures.

9. Documents are distributed in accordance with applicable distribution lists. A process is

used to limit the use of outdated procedures. Copies are available to appropriate personnel. 

Issuance and distribution of procedures are documented and refer to the records

management function.

10. The applicant has formal requirements governing temporary changes. Temporary

changes do not involve a change to the ISA. The review and approval process is

documented. Temporary procedures may be issued only when permanent procedures do

not exist to (a) direct operations during testing, maintenance, and modifications, (b) provide

guidance in unusual situations not within the scope of permanent procedures; and, 

(c) provide assurance of orderly and uniform operations for short periods when the plant, a

system, or a component of a system is performing in a manner not covered by existing

permanent procedures, or has been modified or extended in such a manner that portions

of existing procedures do not apply. The discussion establishes a time frame for use of the

temporary procedure and sets the same level of review and approval as for permanent

procedures.

11. Maintenance procedures involving IROFS commit to the topics listed below for corrective

and preventive maintenance, functional testing after maintenance, and surveillance

maintenance activities:

a. Pre-maintenance activities require reviews of the work to be performed, including

procedure reviews for accuracy and completeness.

b. Steps that require notification of all affected parties (operators and supervisors) before

performing work and on completion of maintenance work. The discussion includes

potential degradation of IROFS during the planned maintenance.
Management Measures 11-18 NUREG-1520

c. Control of work by comprehensive procedures to be followed by maintenance

technicians. Maintenance procedures are reviewed by the various safety disciplines,

including criticality, fire, radiation, industrial, and chemical process safety. The

procedures describe, as a minimum, the following:

i. Qualifications of personnel authorized to perform the maintenance or surveillance

ii. Controls on and specification of any replacement components or materials to be used

(this should be controlled by the CM function, to ensure like-kind replacement and

adherence to 10 CFR Part 21)

iii. Post-maintenance testing to verify operability of the equipment

iv. Tracking and records management of maintenance activities

v. Safe work practices (e.g., lockout/tagout, confined space entry, moderation control or

exclusion area, radiation or hot work permits, and criticality, fire, chemical, and

environmental issues) 

12. The applicant conducts periodic reviews of procedures to assure their continued

accuracy and usefulness and establishes the time frame for reviews of the various types

of procedures. At a minimum all operating procedures are reviewed every 5 years and

emergency procedures are reviewed every year. The applicant describes the use and

control of procedures. Provisions allow for operations to stop and place the process in a

safe condition if a step of a procedure cannot be performed as written. Guidance

identifies the manner in which procedures are to be implemented. Routine procedural

actions that are frequently repeated may not require the procedure to be present. 

Procedures for complex jobs or dealing with numerous sequences where memory

cannot be trusted may require valve alignment check sheets, approved operator aids, or

in-hand procedures that are referenced directly when the job is conducted. 

11.4.3.5 Audits and Assessments

The NRC reviewers should find the applicant’s submittal acceptable if it provides reasonable

assurance that the following regulatory review criteria for audits and assessments are

adequately addressed and satisfied.

1. The applicant describes policy directives covering the audit and assessment function (i.e., at

a minimum, the activities to be audited, audit frequency, guidance in conducting the audit or

assessment, assigned responsibilities for each phase of the work, and procedures for

recording the results and recommending actions to be taken).

2. The applicant has committed to conduct internal audits and independent assessments of

activities significant to facility safety and environmental protection.

3. Audits will be conducted to verify that operations are being conducted in accordance with

regulatory requirements and commitments in the license application.
Management Measures 11-19 NUREG-1520

4. Independent assessments will be conducted by offsite groups or individuals not involved in

the licensed activity, to verify that the health, safety, and environmental compliance

functions are effectively achieving their designed purposes.

5. Audits and assessments will be conducted for the areas of radiation safety, nuclear

criticality safety, chemical safety, fire safety, environmental protection, emergency

management, QA, CM, maintenance, training and qualification, procedures, incident

investigation, and records management.

6. Qualified personnel without direct responsibility for the function and area being audited or

assessed will be used. The staff positions and committees responsible for audits and

assessments are specified. The levels of management to which results are reported, and

the systems to provide corrective actions are also described. 

11.4.3.6 Incident Investigations

The applicant’s description and commitments in the application will be acceptable if the

reviewer finds reasonable assurance of the following:

1. The applicant will establish a process to investigate abnormal events that may occur during

operation of the facility, to determine their specific or generic root cause(s) and generic

implications, to recommend corrective actions, and to report to the NRC as required by 10

CFR 70.50 and 70.74. The investigation process should include a prompt risk-based

evaluation and, depending on the complexity and severity of the event, one individual may

suffice to conduct the evaluation. The investigator(s) will be independent from the line

function(s) involved with the incident under investigation. Investigations will begin within 48

hours of the abnormal event, or sooner, depending on the safety significance of the event. 

The record of IROFS failures required by 10 CFR 70.62(a)(3)for IROFS should be reviewed

as part of the investigation.

2. The applicant will monitor and document corrective actions through completion.

3. The applicant will maintain documentation so that "lessons learned" may be applied to

future operations of the facility. Details of the event sequence will be compared with

accident sequences already considered in the ISA, and the ISA Summary will be modified to

include evaluation of the risk associated with accidents of the type actually experienced.

The applicant has a formal policy or procedure in place for conducting an incident investigation,

and the policy or procedures contain the following elements:

1. A documented plan for investigating an abnormal event. This plan is separate from any

required Emergency Plan. The investigation of an abnormal event should begin as soon as

possible, commensurate with the safety of the investigative team, after the event has been

brought under control.

2. A description of the functions, qualifications, and responsibilities of the manager who would

lead the investigative team and those of the other team members; the scope of the team's

authority and responsibilities; and assurance of cooperation of management.
Management Measures 11-20 NUREG-1520

3. Assurance of the team's authority to obtain all the information considered necessary and its

independence from responsibility for or to the functional area involved in the incident under

investigation.

4. Procedures requiring maintenance of all documentation relating to abnormal events for 2

years or for the life of the operation, whichever is longer.

5. Guidance for personnel conducting the investigation on how to apply a reasonable,

systematic, structured approach to determine the specific or generic root cause(s) and

generic implications of the problem. The level of investigation should be based on a graded

approach relative to the severity of the incident.

6. Requirements to make available original investigation reports to the NRC on request.

7. A system for monitoring the completion of appropriate corrective actions.

The assessment of the adequacy of the applicant's commitments to establish and use a plan

for the investigation of abnormal events will also be based on the following acceptance criteria:

1. The licensee has described the overall plan and method for investigating abnormal events.

2. The functions, responsibilities, and scope of authority of investigators and/or teams are

documented in the plan.

3. Qualified internal or external investigators are appointed to serve on investigating teams

when required. The teams will include at least one process expert and at least one team

member will be trained in root cause analysis.

4. The applicant commits to prompt investigation of any abnormal events and precursors to

abnormal events (such as undetected failure of IROFS).

5. The investigation process and investigating team are independent of the line management,

and participants are assured of no retaliation for participating in investigations.

6. A reasonable, systematic, structured approach is used to determine the specific or generic

root cause(s) and generic implications of abnormal events.

7. Auditable records and documentation related to abnormal events, investigations, and root

cause analysis are maintained. For each abnormal event, the incident report should include

a description, contributing factors, a root cause analysis, and findings and

recommendations. Relevant findings are reviewed with all affected personnel.

8. Documented corrective actions are taken within a reasonable period to resolve findings from

abnormal event investigations.

11.4.3.7 Records Management

The reviewer will find the applicant’s records management system for records acceptable if it

satisfies the following criteria:
Management Measures 11-21 NUREG-1520

1. Records are prepared, verified, characterized, and maintained 

2. Records are legible, identifiable, and retrievable for their designated lifetimes 

3. Records are protected against tampering, theft, loss, unauthorized access, damage, or

deterioration for the time they are in storage 

4. Procedures are established and documented specifying the requirements and

responsibilities for record selection, verification, protection, transmittal, distribution,

retention, maintenance, and disposition

5. The organization and procedures are in place to promptly detect and correct any

deficiencies in the records management system or its implementation 

The types of records that should be included in the system are listed in Appendix B to this SRP

chapter. Records are categorized by relative safety importance to identify record protection and

storage needs and to designate the retention period for individual kinds of records. The

procedures should (a) assign responsibilities for records management, (b) specify the authority

needed for records retention or disposal, (c) specify which records must have controlled access

and provide the controls needed, (d) provide for the protection of records from loss, damage,

tampering, theft or during an emergency, and (e) specify procedures for ensuring that the

records management system remains effective.

For computer codes and computerized data used for activities relied on for safety, as specified

in the ISA Summary, the applicant establishes procedure(s) for maintaining readability and

usability of older codes and data as computing technology changes. The procedures could

include transfer of the older forms of information (e.g., punched cards or paper tapes) and

codes for older computing equipment to contemporary computing media and equipment. 

Records of IROFS failures must be kept and updated in accordance with 10 CFR 70.62(a)(3). 

Record revisions necessitated by post-failure investigation conclusions should be made within 5

working days of the completion of the investigation (10 CFR 70.62(a)(3) states “promptly”).

11.4.3.8 Other QA Elements

To be acceptable, the applicant's QA elements should be structured to apply appropriate

measures to IROFS, which may include site design features. Both the number and safety

grading of QA elements may be applied in proportion to the importance of the item to safety (a

graded approach). Applicants’ and licensees’ QA elements are expected to differ based on the

purpose and complexity of the facility and processes to be controlled.

The ISA Summary should identify the IROFS, the degree of their importance to safety, and

related activities that are required for safety. An applicant may choose to apply all QA elements

and at the highest level to all IROFS or may grade the application in proportion to the

importance of the item to the achievement of safety. 

All IROFS should have all appropriate QA elements applied. If the applicant grades the

application of QA elements, the relative risk importance rankings of IROFS, as established

within the maintenance function, should parallel the rankings used in for QA elements.
Management Measures 11-22 NUREG-1520

A checklist for evaluating the application of QA elements is given below. If the application of

QA is graded, the attributes described for each element listed below are applied to accident

sequences based on the highest level of risk. The application of QA elements may be reduced

by modifying or eliminating either the number of elements or the attributes within each element,

based on evaluations performed and documented in the ISA. Attributes of QA elements are 

as follows:

1. The applicant describes the (a) organizational structure, (b) functional responsibilities, and

(c) charts of the lines, interrelationships, and areas of responsibility and authority for all

organizations performing activities relied on for safety, including the organization of the

applicant and, as applicable, its principal contractors (architect/engineer, constructor,

construction manager, and operator). Persons or organizations responsible for ensuring

that appropriate QA has been established and for verifying that activities affecting quality

have been correctly performed have sufficient authority, access to work areas, and

organizational independence to carry out their responsibilities.

2. The applicant may describe its application of QA elements in the form of a QA program, in

which the applicant commits to meet the applicable requirements of applicable industry

standards. The commitment may describe the applicant’s graded approach to QA, in which

measures are implemented consistent with an item’s importance to safety, or the

commitment may describe a QA program applied to all IROFS. The application of QA

elements should be well-documented, planned, implemented, and maintained to provide

reasonable assurance that, together with the other management measures, IROFS will be

available and reliable when needed. The QA program should be functional before

performing the ISA required by 10 CFR Part 70. See references in Section 11.7 (e.g.,

ANSI/ASME NQA-1).

3. A design control function is established that includes design inputs, process, analyses,

verification, interfaces, changes, and design documentation and records (see Sections

11.3.1, 11.4.3.1, 11.5.2.1, and 11.6.1 for details on CM).

4. Applicable design bases and other requirements necessary to provide reasonable

assurance of quality are included or referenced in documents for procurement of items or

services relied on for safety. To the extent necessary, suppliers are required to have QA

consistent with the quality level of the item or service to be procured.

5. Activities affecting quality are prescribed by and performed in accordance with documented

instructions, procedures, or drawings of a type appropriate for the circumstances (see

Sections 11.3.4, 11.4.3.4, 11.5.2.4, and 11.6.4 for details on procedures).

6. The preparation, issuance, and modification of documents that specify quality requirements

or prescribe activities affecting quality are controlled to provide reasonable assurance that

the appropriate documents are in use. Document changes are reviewed for adequacy and

approved for implementation by authorized personnel (see Sections 11.3.1, 11.4.3.1,

11.5.2.1, and 11.6.1 for details on CM and Sections 11.3.4, 11.4.3.4, 11.5.2.4, and 11.6.4

for details on procedures).

7. Purchased IROFS and services relied on for safety are controlled to provide reasonable

assurance of conformance with specified requirements.
Management Measures 11-23 NUREG-1520

8. Provisions are made to identify and control IROFS and to provide reasonable assurance

that incorrect or defective items are not use.

9. Measures are established to maintain the acceptability of special processes used in the

course of construction, maintenance, modifications, and testing activities (e.g., welding,

heat treating, nondestructive testing, and chemical cleaning) and to assure that they are

performed by qualified personnel using qualified procedures and equipment.

10. Inspections required to verify conformance of IROFS with requirements are planned and

executed. Inspection requirements are specified in written procedures with provisions

included for documenting and evaluating inspection results (see Sections 11.3.4,

11.4.3.4, 11.5.2.4, and 11.6.4 for details on procedures). Personnel qualification

programs are established for inspection test personnel (see Sections 11.3.3, 11.4.3.3,

11.5.2.3, and 11.6.3 for details on training and qualifications).

11. Tests are conducted to verify that IROFS conform to specified requirements and will

perform satisfactorily in service. Test requirements are specified in written procedures

with provisions included for documenting and evaluating test results (see Sections

11.3.4, 11.4.3.4, 11.5.2.4, and 11.6.4 for details on procedures). Personnel qualification

programs are established for test personnel (see Sections 11.3.3, 11.4.3.3, 11.5.3, and

11.6.3 for details on training and qualifications).

12. Provisions are made to provide reasonable assurance that tools, gauges, instruments,

and other measuring and testing devices are properly identified, controlled, calibrated,

and adjusted at specified intervals to maintain performance within required limits.

13. Provisions are made to control the handling, storage, shipping, cleaning, and

preservation of IROFS, in accordance with work and inspection instructions, to prevent

damage, loss, and deterioration caused by environmental conditions such as

temperature or humidity.

14. Provisions are made to control the inspection, test, and operating status of IROFS to

prevent inadvertent use of nonconforming items or bypassing of inspections and tests.

15. Provisions are made to control the identification, segregation, disposition, and

prevention of installation or use of nonconforming IROFS. 

16. Provisions are made to provide reasonable assurance that conditions adverse to safety

are promptly identified and corrected and that measures are taken to preclude

repetition. These actions should be documented and reported to appropriate levels of

management (see Sections 11.3.6, 11.4.3.6, 11.5.2.6, and 11.6.6 for details on incident

investigations, and Sections 11.3.5, 11.4.3.5, 11.5.2.5, and 11.6.5 for details on audits

and assessments).

17. Provisions are made for the identification, retention, retrieval, and maintenance of

records that furnish evidence of the control of quality for IROFS (see Sections 11.3.7,

11.4.3.7, 11.5.2.7, and 11.6.7 for details on records management). 

18. Provisions are made for planning and scheduling assessments and audits to verify

compliance with, and to determine the effectiveness of, QA; responsibilities and
Management Measures 11-24 NUREG-1520

procedures are identified for assessing, auditing, documenting, and reviewing results

and for designating management levels to review assessment and audit results; and

provisions are made for incorporating the status of findings and recommendations in

management reports (see Sections 11.3.5, 11.4.3.5, 11.5.2.5, and 11.6.5 for details on

audits and assessments).

19. The applicant’s provisions for continuing QA address reviews and updates of QA

documents based on reorganizations, revised activities, lessons learned, changes to

applicable regulations, and other QA program changes.

11.5 REVIEW PROCEDURES

11.5.1 Acceptance Review

The primary reviewer should evaluate the application to determine whether it addresses the

“Areas of Review” discussed in Section 11.3, above. If significant deficiencies are identified,

the applicant should be requested to submit additional material before the start of the safety

evaluation review.

11.5.2 Safety Evaluation

After the primary reviewer determines that the application is acceptable for review in

accordance with Section 11.5.1, above, the primary and secondary reviewers should perform a

safety evaluation review against the acceptance criteria described in Section 11.4. Review

procedures for each criterion are discussed in the sections below. If deficiencies are identified,

the applicant should be requested to submit additional information or modify the submittal to

meet the acceptance criteria in Section 11.4 of this SRP. The reviews for all management

measures should be coordinated with the primary reviewer of the ISA Summary.

11.5.2.1 CM

1. CM Policy Management

The primary reviewer should consider whether the CM plan acceptably states management

commitments, gives the policy directive, and defines key responsibilities, terminology, and

equipment scope. The secondary reviewers should examine the ISA Summary and the

ISA, as needed, to assure that identified IROFS will be subject to the CM function. 

Appropriate interfaces both within the CM function and with external organizations and

functions should be examined. In particular, the functional interfaces with QA,

maintenance, and training (including qualification) should be examined. The reviewers

should look for the applicant's identification of required databases and the rules for their

maintenance. The reviewers should examine implementing procedures for the CM function. 

2. Design Requirements

The primary reviewer should confirm that the design process leading to drawings and other

statements of requirements proceeds logically from the design basis. The design basis is a

set of facts about the systems covered by CM that has been reviewed and approved by

appropriate authority within the organization. The reviewers should verify that specific
Management Measures 11-25 NUREG-1520

personnel are assigned the responsibility for maintaining the design bases and

requirements. The reviewers should verify that the IROFS to be listed under CM will be

clearly defined in the requirements documents, along with the assignment of any grades or

quality levels. This part of the review should be coordinated with the ISA primary reviewer. 

The ISA Summary should specify all IROFS, and the applicant should have indicated in the

ISA Summary what level of CM attributes is applied to a particular item. However, in the

ISA Summary this indication may consist of only an index or category designation. The

definitions of the multiple CM levels, if used, should be in the CM section of the application. 

The primary reviewer for the CM section is responsible for determining if the reduced levels

the applicant would apply to IROFS for accident sequences with lesser consequences are

adequate. 

3. Document Control

The primary reviewer should evaluate the applicant's material showing that the CM system

will capture documents that are relevant and important to safety. The documents should

include design requirements, the ISA, the ISA Summary, as-built drawings, specifications,

all safety-important operating procedures, procedures involving training, maintenance and

audits and assessments, emergency operating procedures, emergency response plans,

system modification documents, assessment reports, and other documents that the

applicant deems to pertain to the CM function. The primary reviewer should determine

whether a controlled document database is used to control documents and track document

change status. Rules of storage for originals or master copies of documents within the

scope of the CM function follow the guidance of “Records Management.”

4. Change Control

The primary reviewer should be able to find that the description of change control within the

CM function commits to have acceptable methods for (a) the identification of changes in

configurations that are IROFS, (b) technical and management review of changes, and (c)

tracking and implementing changes, including placement of documentation in a document

control center and dissemination to affected functions such as training, engineering,

operations, maintenance, and other QA elements. 

5. Assessments

The primary reviewer should be able to find that both document assessments and physical

assessments (system walkdowns) will be conducted periodically to check the adequacy of

the CM function. The primary reviewer should be able to find that all assessments and

followups will be documented. These reports can provide a supporting basis for future

changes. 

6. Design Reconstitution (Existing Facilities Only)

Design reconstitution may be necessary for existing facilities if current design information is

not adequate. The primary reviewer examines the applicant's description of work to

establish, organize, and document design requirements and design bases for items for

which design information was not available before the application was submitted. Of

particular importance are the methods used to evaluate, verify, and validate reconstituted

design data for IROFS. For existing facilities, the design requirements and physical
Management Measures 11-26 NUREG-1520

configuration may have greatly changed according to the demands of a changed mission. If

documentation has not kept pace, it will be necessary for the applicant to walk down

systems, update drawings and specifications, perform new calculations and analyses, and

otherwise rebuild the design bases. The reviewer will seek evidence that the need for

design bases reconstitution was investigated, that reconstitution was accomplished as

necessary, and that new or revised documentation was properly incorporated into the CM

function. On the basis of its review, the staff may request that the applicant provide

additional information or modify the submittal to meet the acceptance criteria in Section 11.4

of this SRP.

When the safety evaluation is complete, the secondary staff reviewer, with assistance from

the other reviewers, should prepare the CM input for the Safety Evaluation Report (SER) as

described in SRP Section 11.6, using the regulatory acceptance criteria from SRP 

Section 11.4.3.1. 

11.5.2.2 Maintenance

The reviewer will evaluate the applicant’s description of how the maintenance function will

coordinate with and use the other management measures listed in this chapter. The primary

reviewer should consult with the supporting reviewers to identify any common weaknesses in

the applicant’s approach and consider these during the review.

An acceptable maintenance function includes descriptions and applicant’s commitments

regarding corrective maintenance, preventive maintenance, surveillance/monitoring, and

functional testing. 

When the safety evaluation is complete, the secondary staff reviewer, with assistance from the

other reviewers, should prepare the maintenance input for the SER as described in SRP

Section 11.6 using the regulatory acceptance criteria from SRP Section 11.4.3.2. 

11.5.2.3 Training and Qualification

The primary reviewer performs a safety evaluation against the acceptance criteria described in

Section 11.4, recognizing that the training objectives and methods and the required personnel

qualification may be graded to correspond to the hazard potential of the facility, and the IROFS

and to the complexity of the needed training. The review should evaluate the adequacy of

training and qualification on the basis of how well it fulfills the applicant’s training objectives,

especially when human factors are relied on for safety. The review should determine whether

the applicant has adequately planned for the training and personnel qualification to be

accomplished and whether necessary policies, procedures, and instructions will be in place and

appropriate training and qualification will be accomplished before personnel begin activities

relied on for safety. The reviewers should focus on the training and qualification of personnel

who will perform activities relied on for safety. 

The secondary reviewer should confirm that the applicant’s personnel training and qualification

commitments are consistent with other sections of the submittal. 

The supporting reviewer should become familiar with the applicant’s personnel training and

qualification commitments and determine whether ongoing activities are in agreement

with them.
Management Measures 11-27 NUREG-1520

The review should result in a determination that there is reasonable assurance that the

applicant's personnel training and qualification will result in only properly trained and qualified

personnel performing activities relied on for safety. 

When the safety evaluation is complete, the secondary staff reviewer, with assistance from the

other reviewers, should prepare the personnel training and qualification input for the SER as

described in Section 11.6, using the acceptance criteria from Section 11.4.3.3.

11.5.2.4 Procedures

On acceptance of the application for review, the secondary reviewer will evaluate whether the

applicant has adequately addressed the acceptance criteria listed in Section 11.4. The

secondary reviewer will document in an SER that the applicant has committed to the following:

1. IROFS identified in the ISA Summary are highlighted in safety procedures (including

procedures that constitute administrative controls for safety). There may be several

levels of requirements within procedures for diagnosing and correcting process upsets

and dealing with abnormal situations or other matters. There is a clear hierarchy of

requirements within procedures. Cautions and notes appearing in procedures precede

the steps to which they apply. Rules for entering and leaving a procedure are clear.

2. Procedures important to safety are independently verified and validated before use, and

this is documented in a policy on procedures.

3. Policy and administrative procedures, noncrucial operating procedures, and other

nonoperational procedures that do not impact IROFS or other environmental, safety,

and health concerns need not be controlled with the stringency applied to operating

procedures or management control procedures associated with IROFS specified by the

ISA Summary. The applicability of less stringent procedure control should be specified

to avoid misunderstandings in implementation.

4. Changes to operating management measure, or maintenance procedures are reviewed

and approved by an independent multidisciplinary safety review team and controlled by

the CM function. 

5. The applicant includes a statement to follow approved procedures while processing

licensed SNM.

6. Procedures exist for the notification of operations personnel before and after

maintenance is performed on IROFS, and activities are controlled by procedures.

When the safety evaluation is complete, the secondary staff reviewer, with assistance from

the other reviewers, should prepare the procedures input for the SER, as described in

Section 11.6, using the acceptance criteria from Section 11.4.3.4.

11.5.2.5 Audits and Assessments

After determining that the application is acceptable for review in accordance with Section

11.5.1, above, the secondary reviewer will perform a safety evaluation against the acceptance

criteria described in Section 11.4. The review should determine whether the applicant has
Management Measures 11-28 NUREG-1520

adequately planned for audits and assessments to be accomplished and whether necessary

policies, personnel, procedures, and instructions will be in place to begin audits and

assessments early, that is, during the design of IROFS. 

If the applicant references other sections of the application when describing its audits and

assessments, the primary reviewer should review these other sections of the application to

determine the applicant's overall commitment to audits and assessments and the proposed

method for implementation. The reviewers should focus on audits and assessments of IROFS. 

The secondary reviewer should confirm that the applicant’s audit and assessment commitments

are consistent with other sections of the submittal. The secondary reviewer is also responsible

for integrating the audit and assessment input into the SER. 

The supporting reviewer should become familiar with the applicant’s audit and assessment

commitments and determine whether ongoing audits and assessments of the applicant and the

applicant’s principal contractors are in agreement with them.

The review should result in a determination that there is reasonable assurance that the audits

and assessments of the applicant and the applicant’s principal contractors will provide

additional assurance that IROFS will perform satisfactorily in service and that activities relied on

for safety will be performed satisfactorily.

When the safety evaluation is complete, the secondary staff reviewer, with assistance from the

other reviewers, should prepare the audits and assessments input for the SER, as described in

Section 11.6, using the acceptance criteria from Section 11.4.3.5.

11.5.2.6 Incident Investigations

The primary reviewer will verify that the applicant has described a comprehensive incident

investigation function based on the areas of review in Section 11.3 and the acceptance criteria

presented in Section 11.4 of this SRP. 

For existing facilities, the primary reviewer will consult with the NRC inspection staff and review

any historical information regarding the adequacy of the applicant’s incident investigation

process. On the basis of its review, the staff may request that the applicant provide additional

information or modify the submittal to meet the acceptance criteria in Section 3.7.4 of this SRP.

When the safety evaluation is complete, the secondary staff reviewer, with assistance from the

other reviewers, should prepare the incident investigation input for the SER as described in

Section 11.6, using the acceptance criteria from Section 11.4.3.6.

11.5.2.7 Records Management

The reviewer will review the applicant's records management system to determine the

adequacy of the policies, procedures, and practices. The reviewer should coordinate this

review with the person reviewing the CM function. 

For fuel cycle facilities that are parts of larger organizations, certain documents may be retained

or stored at a site other than the facility site. For example, master drawings for structures might
Management Measures 11-29 NUREG-1520

be kept in the engineering department of the headquarters of the parent company. The

reviewer may choose to review the physical characteristics of these offsite record storage areas

as well, particularly for records for IROFS for high-consequence accident sequences. 

When the safety evaluation is complete, the secondary staff reviewer, with assistance from the

other reviewers, should prepare the records management input for the SER, as described in

Section 11.6, using the acceptance criteria from Section 11.4.3.7.

11.5.2.8 Other QA Elements

After the primary reviewer has determined that the application is acceptable for review in

accordance with Section 11.5.1, above, the primary reviewer should confirm that the applicant’s

(and the applicant’s principal contractors’) QA element commitments are consistent with other

sections of the submittal. The secondary reviewer should review the QA elements information

with respect to the acceptance criteria in Section 11.4. The secondary staff reviewer should

determine whether the applicant has adequately planned the work to be accomplished and

whether necessary policies, procedures, and instructions either are in place or will be in place

before work starts. The review is based on an assessment of the material presented. It should

provide reasonable assurance that the applicant's QA elements, maintenance, and CM are

coordinated and that the QA elements are an integral part of everyday work activities. The

review should provide reasonable assurance that the applicant will be able to monitor the

effectiveness of the implementation of QA elements and will make needed adjustments on a

timely basis. The staff is to look for and measure the effectiveness of the QA elements design,

not just the existence of appropriate QA elements.

The secondary reviewer should also determine that the applicant has specified the QA

elements criteria, the basis for choosing the criteria and the proposed method for

implementation. If the applicant references other sections of the application when describing its

QA elements, the reviewer should review these other sections of the application to determine

the applicant's commitment to the QA elements and the proposed method for implementation. 

The supporting reviewers should become familiar with the applicant’s (and principal

contractors’) QA element commitments and determine whether ongoing activities are in

agreement with them. 

Staff reviewers of SRP Chapters 3 through 11 should determine whether IROFS within their

areas of review are specified to be within the appropriate QA elements and level.

The review should result in a determination that there is reasonable assurance that the

applicant's (and the applicant’s principal contractors’) QA elements will provide reasonable

assurance that IROFS will perform their safety functions in a satisfactory manner. 

When the safety evaluation is complete, the secondary staff reviewer, with assistance from the

other reviewers, should prepare the QA input for the SER, as described in SRP Section 11.6,

using the acceptance criteria from SRP Section 11.4.3.8.
Management Measures 11-30 NUREG-1520

11.6 EVALUATION FINDINGS

The staff's evaluation should verify that the license application provides sufficient information to

satisfy the regulatory requirements of Section 11.4.1 and that the regulatory acceptance criteria

in Section 11.4.3 have been appropriately considered in satisfying the requirements. On the

basis of this information, the staff should conclude that this evaluation is complete. The

reviewers should write material suitable for inclusion in the SER prepared for the entire

application. The SER should include a summary statement of what was evaluated and the

basis for the reviewers' conclusions.

In cases where the SER is drafted in advance of resolving all open issues, the reviewer should

document the review as described below and include a list of open issues that require

resolution before the staff can reach a reasonable-assurance-of-safety conclusion. For partial

reviews, revisions, and process changes, the reviewer should use applicable sections of the

acceptance criteria and the SER should be written to reflect what portions were not reviewed

and the safety significance, if any. 

The staff can document the evaluation as follows:

11.6.1 CM

The staff has reviewed the CM function for (name of facility) according to Section 11 of the

SRP. [Insert a summary statement of what was evaluated and why the reviewer finds the

submittal acceptable.]

The applicant has suitably and acceptably described its commitment to a proposed CM

system, including the method for managing changes in procedures, facilities, activities, and

equipment for IROFS. Management-level policies and procedures, including an analysis

and independent safety review of any proposed activity involving IROFS, are described that

will provide reasonable assurance that consistency among design requirements, physical

configuration, and facility documentation is maintained as part of a new activity or change in

an existing activity involving licensed material. The management measures will include (or

do include) the following elements of CM:

1. CM Management

The organizational structure, procedures, and responsibilities necessary to implement

CM are in place or committed to.

2. Design Requirements

The design requirements and bases are documented and supported by analyses, and

the documentation is maintained current.

3. Document Control

Documents, including drawings, are appropriately stored and accessible. Drawings and

related documents captured by the system are those necessary and sufficient to

adequately describe IROFS.
Management Measures 11-31 NUREG-1520

4. Change Control

Responsibilities and procedures adequately describe how the applicant will achieve and

maintain strict consistency among the design requirements, the physical configuration,

and the facility documentation. Methods are in place for suitable analysis, review,

approval, and implementation of identified changes to IROFS. This includes appropriate

CM controls to assure configuration verification, functional tests, and accurate

documentation for equipment or procedures that have been modified.

5. Assessments

The applicant has committed to an adequate function that includes both initial and

periodic assessments as described in the acceptance criteria in this SRP. The

assessments are expected to verify and assure the adequacy of the CM function.

6. Design Reconstitution (Existing Facilities Only)

The applicant has adequately described that design reconstitution, if required, has been

done. Current design bases are available and verified for all IROFS, such that the

configuration is consistent with the as-built facility documentation.

11.6.2 Maintenance

The applicant has committed to maintenance of IROFS. The applicant’s maintenance

commitments contain the basic elements to maintain availability and reliability: corrective

maintenance, preventive maintenance, functional testing, equipment calibration, and work

control for maintenance of IROFS. The applicant’s maintenance function is proactive, using

maintenance records, PM records, and surveillance tests to analyze equipment

performance and to seek the root causes of repetitive failures.

The surveillance/monitoring, PM, and functional testing activities described in the license

application provide reasonable assurance that the IROFS identified in the ISA Summary will

be available and reliable to prevent or mitigate accident consequences.

The maintenance function (1) is based on approved procedures, (2) employs work control

methods that properly consider personnel safety, awareness of facility operating groups,

QA, and the rules of CM, (3) uses the ISA Summary to identify IROFS that require

maintenance and at what level, (4) justifies the PM intervals in the terms of equipment

reliability goals, (5) provides for training that emphasizes the importance of IROFS identified

in the ISA Summary identified IROFS, regulations, codes, and personal safety, and

(6) creates documentation that includes records of all surveillance, inspections, equipment

failures, repairs, and replacements of IROFS.

The staff concludes that the applicant’s maintenance functions meet the requirements of 10

CFR Part 70, and provide reasonable assurance that the health and safety of the worker

and the public are provided for.
Management Measures 11-32 NUREG-1520

11.6.3 Training and Qualification

Based on its review of the license application [Insert a summary statement of what was

evaluated and why the reviewer finds the submittal acceptable], the NRC staff has

concluded that the applicant has adequately described and assessed its personnel training

and qualification in a manner that (1) satisfies regulatory requirements, (2) is consistent with

the guidance in this SRP, and (3) is acceptable. 

There is reasonable assurance that implementation of the described training and

qualification will result in personnel who are qualified and competent to design, construct,

startup, operate, maintain, modify, and decommission the facility safely. The staff

concludes that the applicant's plan for personnel training and qualification meets the

requirements of 10 CFR Part 70.

11.6.4 Procedures

The application has described a suitably detailed process for the development, approval,

and implementation of procedures. IROFS have been addressed, as well as items

important to the health of facility workers and the public and to the protection of

the environment. The staff concludes that the applicant’s plan for procedures meets the

requirements of 

10 CFR Part 70.

11.6.5 Audits and Assessments

Based on its review of the license application [Insert a summary statement of what was

evaluated and why the reviewer finds the submittal acceptable], the NRC staff has

concluded that the applicant has adequately described its audits and assessments. The

staff has reviewed the applicant's plan for audits and assessments and finds it acceptable. 

The staff concludes that the applicant's plan for audits and assessments meets the

requirements of 10 CFR Part 70 and provides reasonable assurance of protection of the

health and safety of the public and workers and the environment. 

11.6.6 Incident Investigations

The applicant has committed to and established an organization responsible for 

(1) performing incident investigations of abnormal events that may occur during operation of

the facility, (2) determining the root cause(s) and generic implications of the event, and 

(3) recommending corrective actions for ensuring a safe facility and safe facility

operations, in accordance with the acceptance criteria of Subsection 11.4 of the SRP. 

The applicant has committed to monitoring and documenting corrective actions through to

completion.

The applicant has committed to the maintenance of documentation so that "lessons

learned" may be applied to future operations of the facility.
Management Measures 11-33 NUREG-1520

Accordingly, the staff concludes that the applicant's description of the incident

investigation process complies with applicable NRC regulations and is adequate. 

11.6.7 Records Management

The staff has reviewed the applicant’s records management system against the SRP’s

acceptance criteria and concluded that the system (1) will be effective in collecting,

verifying, protecting, and storing information about the facility and its design, operations,

and maintenance and will be able to retrieve the information in readable form for the

designated lifetimes of the records, (2) will provide a records storage area(s) with the

capability to protect and preserve health and safety records that are stored there during the

mandated periods, including protection of the stored records against loss, theft, tampering,

or damage during and after emergencies, and (3) will provide reasonable assurance that

any deficiencies in the records management system or its implementation will be detected

and corrected in a timely manner.

11.6.8 Other QA Elements

Based on its review of the license application [Insert a summary statement of what was

evaluated and why the reviewer finds the submittal acceptable]. The review record should

demonstrate the adequacy of the applicant's application of other QA elements, as applied to

IROFS, for design, construction, and operations] the NRC staff has concluded that the

applicant has adequately described the application of other QA elements (and the

applicable QA elements of its principal contractors). The staff concludes further that:

1. The applicant has established and documented a commitment to an organization

responsible for developing, implementing, and assessing the management measures for

providing reasonable assurance of safe facility operations in accordance with the criteria

in Section 11.4 of this SRP.

2. The applicant has established and documented a commitment to QA elements, and the

administrative measures for staffing, performance, assessing findings, and

implementing corrective action are in place.

3. The applicant has developed a process for preparation and control of written

administrative plant procedures, including procedures for evaluating changes to

procedures, IROFS, and tests. A process for review, approval, and documentation of

procedures will be implemented and maintained.

4. The applicant has established and documented surveillances, tests, and inspections to

provide reasonable assurance of satisfactory inservice performance of IROFS. 

Specified standards or criteria and testing steps have been provided.

5. Periodic independent audits are conducted to determine the effectiveness of the

management measures. Management measures will provide for documentation of audit

findings and implementation of corrective actions.

6. Training requirements have been established and documented to provide employees

with the skills to perform their jobs safely. Management measures have been provided
Management Measures 11-34 NUREG-1520

for evaluation of the effectiveness of training against predetermined objectives and

criteria.

7. The organizations and persons performing QA element functions have the required

independence and authority to effectively carry out their QA element functions without

undue influence from those directly responsible for process operations.

8. QA elements cover the IROFS, as identified in the ISA Summary, and measures are

established to prevent hazards from becoming pathways to higher risks and accidents.

Accordingly, the staff concludes that the applicant's application of other QA elements (and

the applicable QA elements of its principal contractors) meets the requirements of 

10 CFR Part 70 and provides reasonable assurance of protection of public health and safety

and of the environment.

11.7 REFERENCES

American National Standards Institute/American Society of Mechanical Engineers Standard,

“Quality Assurance Requirements for Nuclear Facility Applications,” ANSI/ASME NQA-1-1994.

International Atomic Energy Agency, “Establishing and implementing a Quality Assurance

Program,” Safety Guide 50-SG-Q1, 1995.

International Organization for Standardization (ISO) 9000 series of quality management

standards.

U.S. Code of Federal Regulations, Title 10, Part 70, “Domestic Licensing of Special Nuclear

Material,” U.S. Government Printing Office, Washington, DC.

U.S. Code of Federal Regulations, Title 10, Part 21, “Reporting of Defects and Noncompliance,”

U.S. Government Printing Office, Washington D.C., as revised.

U.S. Code of Federal Regulations, Title 29, Chapter XVII, Section 1910.119, “Process Safety

Management of Highly Hazardous Chemicals,” U.S. Government Printing Office, Washington

D.C., as revised.

U.S. Code of Federal Regulations, Title 40, Part 68, “Risk Management Program for Chemical

Accidental Release Prevention,” U.S. Government Printing Office, Washington D.C., as revised.

U.S. Department of Energy, “DOE Standard: Guide for Operational CM Function,” Parts I and

II, DOE-STD-1073-93.

U.S. Department of Energy, Draft, “Implementation Guide for Use with 10 CFR Part 830.120

and DOE Order 5700.6C,” September 1997.

U.S. Nuclear Regulatory Commission, “A Systematic Approach to Repetitive Failures,”

NUREG/CR-5665, February 1991.
Management Measures 11-35 NUREG-1520

U.S. Nuclear Regulatory Commission, “Guidance on Management Controls/Quality Assurance,

Requirements for Operation, Chemical Safety, and Fire Protection for Fuel Cycle Facilities,”

Federal Register 54 (No. 53), 11590!11598, March 21, 1989.

U.S. Nuclear Regulatory Commission, “Guide to NRC Reporting and Recordkeeping

Requirements,” NUREG-1460, Rev. 1, July 1994.

U.S. Nuclear Regulatory Commission, “Maintenance and Inspection,” Inspection Procedure

88062, January 16, 1996.

U.S. Nuclear Regulatory Commission, “Maintenance and Surveillance Testing,” Inspection

Procedure 88025, May 23, 1984.

U.S. Nuclear Regulatory Commission, “Proposed Method for Regulating Major Materials

Licensees,” Section 3.2.6, “Configuration Management,” NUREG-1324, 1992.

U.S. Nuclear Regulatory Commission, “Proposed Revision to Code of Federal Regulations, Title

10, Part 70, “Domestic Licensing of Special Nuclear Material,” as revised.

U.S. Nuclear Regulatory Commission, “Root Causes of Component Failures Program: Methods

and Applications,” NUREG/CR-4616, December 1986.

U.S. Nuclear Regulatory Commission, “Suggested Guidance Relating to Development and

Implementation of Corrective Action,” Information Notice 96-28, May 1966.

U.S. Nuclear Regulatory Commission, “Training Review Criteria and Procedures,” NUREG

1220, Rev. 1, January 1993.

Accession #: ML013370424
Appendix A to Chapter 11 11-A-1 NUREG-1520

APPENDIX A

CHECKLIST FOR PROCEDURES

All activities listed below are to be covered by written procedures. The list is not intended to be

all-inclusive or to imply that procedures be developed with the same titles as those on the list. 

This listing is divided into four categories and provides guidance on topics to be covered.

1. Management Control Procedures

Training

Audits and assessments

Incident investigation

Records management 

Configuration management

Quality assurance

Equipment control (lockout/tagout)

Shift turnover

Work control

Procedure management

Nuclear criticality safety

Fire protection

Radiation protection

Radioactive waste management

Maintenance

Environmental protection

Chemical process safety

Operations

Calibration control

Preventive maintenance

2. Operating Procedures 

a. System Procedures That Address Startup, Operation, Shutdown, Control of

Process Operations, and Recovery After a Process Upset

Ventilation

Criticality alarms

Shift routines, shift turnover, and operating practices

Decontamination operations

Uranium recovery

Facility utilities (air, other gases, cooling water, fire water, steam)

Temporary changes in operating procedures

b. Abnormal Operation/Alarm Response

Loss of cooling water

Loss of instrument air

Loss of electrical power

Loss of criticality alarm system
Appendix A to Chapter 11 11-A-2 NUREG-1520

Fires

Chemical process releases

3. Maintenance Activities That Address System Repair, Calibration, Surveillance,

and Functional Testing

Repairs and preventive repairs of items relied on for safety (IROFS)

Testing of criticality alarm units

Calibration of IROFS

High-Efficiency particulate air (HEPA) filter maintenance

Functional testing of IROFS

Relief valve replacement/testing

Surveillance/monitoring

Pressure vessel testing

Nonfired pressure vessel testing

Piping integrity testing

Containment device testing

4. Emergency Procedures

Response to a criticality

Hazardous process chemical releases (including uranium hexafluoride)
Appendix B to Chapter 11 11-B-1 NUREG-1520

APPENDIX B

RECORDS

The requirements for records management vary according to the nature of the facility and the

hazards and risks posed by it. Examples of the records required by 10 CFR Parts 19, 20, 21,

25, and 70 are listed below. The records are listed under the chapter headings of the Standard

Review Plan (SRP). The list is not intended to be exhaustive or prescriptive. Different or

additional records might be required in certain circumstances. Further, the applicant may

choose to organize the records in other ways.

Examples of Records

SRP Chapter

1 General Information

Construction records

Facility and equipment descriptions and drawings

Design criteria, requirements, and bases for items relied on for safety (IROFS)

as specified by the facility CM function

Records of facility changes and associated integrated safety analyses, as

specified by the facility CM function

Safety analyses, reports, and assessments

Records of site characterization measurements and data

Records pertaining to onsite disposal of radioactive or mixed wastes in surface

landfills

Procurement records, including specifications for IROFS

2 Organization and Administration

Administrative procedures with safety implications 

Change control records for material control and accounting program

Organization charts, position descriptions, and qualification records

Safety and health compliance records, medical records, personnel exposure

records, etc.

QA records

Safety inspections, audits, assessments, and investigations
Appendix B to Chapter 11 11-B-2 NUREG-1520

Safety statistics and trends

3 Integrated Safety Analysis

4 Radiation Safety

Bioassay data

Exposure records

Radiation protection (and contamination control) records

Radiation training records

Radiation work permits

5 Nuclear Criticality Safety

Nuclear criticality control written procedures and statistics

Nuclear criticality safety analyses

Records pertaining to nuclear criticality inspections, audits, investigations,

and assessments

Records pertaining to nuclear criticality incidents, unusual occurrences, or

accidents

Records pertaining to nuclear criticality safety analyses

6 Chemical Safety

Chemical process safety procedures and plans

Records pertaining to chemical process inspections, audits, investigations,

and assessments

Diagrams, charts, and drawings

Records pertaining to chemical process incidents, unusual occurrences, or

accidents

Chemical process safety reports and analyses

Chemical process safety training

7 Fire Safety

Fire Hazard Analysis
Appendix B to Chapter 11 11-B-3 NUREG-1520

Fire prevention measures, including hot-work permits and fire watch records

Records pertaining to inspection, maintenance, and testing of fire protection

equipment

Records pertaining to fire protection training and retraining of response teams

Pre-fire emergency plans

8 Emergency Management

Emergency plan(s) and procedures

Comments on emergency plan from outside emergency response organizations

Emergency drill records

Memoranda of understanding with outside emergency response organizations

Records of actual events

Records pertaining to the training and retraining of personnel involved in

emergency preparedness functions

Records pertaining to the inspection and maintenance of emergency response 

equipment and supplies

9 Environmental Protection

Environmental release and monitoring records

Environmental report and supplements to the environmental report, as applicable

10 Decommissioning

Decommissioning records

Financial assurance documents

Decommissioning cost estimates

Site characterization data

Final survey data

Decommissioning procedures
Appendix B to Chapter 11 11-B-4 NUREG-1520

11 Management Measures

11.1 Configuration Management

• Safety analyses, reports, and assessments that support the physical

configuration of process designs, and changes to those designs

• Validation records for computer software used for safety analysis or

material control and accounting

• Integrated Safety Analysis (ISA) documents, including process

descriptions, plant drawings and specifications, purchase specifications

for IROFS

• Approved, current operating procedures and emergency operating

procedures

11.2 Maintenance

• Record of IROFS failures (required by 10 CFR 70.62)

• PM records, including trending and root cause analysis

• Calibration and testing data for IROFS

• Corrective maintenance records

11.3 Training and Qualification

• Personnel training and qualification records

• Procedures

11.4 Procedures

• Standard operating procedures

• Functional test procedures

11.5 Audits and Assessments

• Audits and assessments of safety and environmental activities

11.6 Incident Investigations

• Investigation reports

• Changes recommended by investigation reports, how and when

implemented

• Summary of reportable events for the term of the license

• Incident investigation policy

11.7 Records Management

• Policy 

• Material storage records

• Records of receipt, transfer, and disposal of radioactive material
Appendix B to Chapter 11 11-B-5 NUREG-1520

11. 8 Other Quality Assurance Elements

• Inspection records

• Test records

• Corrective action records
Appendix A to SRP A-1 NUREG-1520

APPENDIX A

FILING STANDARDS FOR SUBMITTALS

The requirements of 10 CFR Part 70, “Domestic Licensing of Special Nuclear Material,” specify

that applications for a license to possess and use special nuclear material should be filed in

accordance with the general information in Sections 70.5, “Communications” and 70.21,

“Filing.” This appendix has been prepared to provide more specific guidance for staff reviewers

regarding the acceptable and preferred format for license and amendment applications. 

Use of a standard submittal format for new and renewal applications facilitates a uniform and

clear presentation by fuel cycle facility applicants. Such clarity and uniformity will facilitate a

timely and uniform review by the NRC staff, and a clear and cogent presentation for

understanding by parties other than the NRC and applicant who may have a legitimate interest

in the application. Information contained in previous submittals, statements, or reports filed with

the NRC with respect to an existing license may be incorporated by reference provided such

references are clear and specific. The information called for in this SRP that is incorporated by

reference to a previous application should be summarized.

Proprietary Information

Proprietary information should be submitted separately. When submitted, it should be clearly

identified and accompanied by the applicant's justifications for requesting its being withheld

from public disclosure, as specified by § 2.790, “Public Inspections, Exemptions, Requests for

Withholding,” of 10 CFR Part 2, “Rules of Practice for Domestic Licensing Proceedings.” The

NRC staff's review of the safety analysis should depend as much as possible on nonproprietary

information to ensure that the public is properly apprized of the reasons for and effects of

licensing actions.

Classified Information

Classified information should be submitted separately. When submitted, it should be clearly

identified and accompanied by the applicant’s justifications for requesting its being withheld

from public disclosure, as specified by applicable classification regulations. The NRC staff’s

review of the safety analysis should depend as much as possible on unclassified information to

ensure that the public is properly apprized of the reasons for and effects of licensing actions.

Style and Composition

Applications should clearly and concisely present the information to demonstrate compliance

with NRC requirements.

Where numerical values are stated, the number of significant figures given should reflect the

accuracy or precision to which the number is known. Where appropriate, estimated limits of

errors or uncertainty should be given.

Abbreviations should be consistent throughout the application and should be consistent with

generally accepted usage. Any abbreviations, symbols, or special terms not in general usage

or unique to the facility should be defined when they first appear in the application or should be

presented in a separate glossary of terms and definitions.
Appendix A to SRP A-2 NUREG-1520

Text pages should be single spaced, with a type face and style suitable for microfilming or

reproduction by image-copying equipment.

Pages should be numbered with the digits corresponding to the chapter followed by a hyphen

and a sequential number, e.g., the third page of Chapter 4 should be number 4-3. Sequential

numbers for the entire report should not be used.

A table of contents should be included in each volume of the renewal application.

Graphical Presentations

Graphical presentations such as drawings, maps, diagrams, sketches, and tables should be

employed where the information may be presented more adequately or conveniently by such

means. All information so presented should be legible, symbols should be clearly defined, and

scales large enough that visual aids are not necessary to interpret pertinent items of

information. These graphical presentations should be located in the section where they are

primarily referenced.

Physical Specifications

Paper size:

Text pages should be 8 1⁄2 × 11 inches; drawings and graphics should also be 8 1⁄2 × 11 inches;

however, a larger size is acceptable provided the finished copy when folded does not exceed

8 1⁄2 × 11 inches.

Paper stock and ink:

The paper stock should be of suitable quality and color, and the ink of suitable density for

handling and reproduction by microfilming or image-copying equipment.

Page margins:

A margin of no less than 1 inch should be maintained on the top, bottom, and binding side of all

pages submitted.

Printing:

The material may be mechanically or photographically reproduced. All pages of text should be

printed on both sides and the image printed head to head. Pages should be punched and

mounted in three-hole ring binders.

Submittal to NRC

In addition to paper copies of the applications submitted in accordance with 10 CFR 70.5 and

70.21, submittals to NRC should include a reproducible copy in electronic media in Corel

WordPerfect version 8.0 or other word processing format that converts accurately to the Corel

WordPerfect format.
Appendix A to SRP A-3 NUREG-1520

Procedures for Updating or Revising Pages

Data and text should be updated or revised by replacing pages. The changed or revised

portion on each page should be highlighted by a "change indicator" mark consisting of a bold

vertical line drawn in the margin opposite the binding margin. The line should be of the same

length as the portion actually changed. All pages submitted to update, revise, or add pages to

the report should show the date of change and a revision or amendment number. A guide page

listing the pages to be inserted and the pages to be removed should accompany the revised

pages. Where major changes or additions are made, a revised table of contents should be

provided.

Accession #: ML013370448