Consolidated Guidance 

about Materials Licenses

Program-Specific Guidance 

about Fixed Gauge Licenses

Draft Report for Comment

Office of Federal and State Materials and 

Environmental Management Programs

NUREG-1556,

Volume 4, Rev. 1
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Consolidated Guidance 

about Materials Licenses

Program-Specific Guidance 

about Fixed Gauge Licenses

Draft Report for Comment

Manuscript Completed: January 2013

Date Published: March 2013

Prepared by: 

J. Cook

B. Juran

A. McCraw

M. Reichard

G. Rivera-Capella

Office of Federal and State Materials and

 Environmental Management Programs

NUREG-1556

Volume 4, Rev. 1
COMMENTS ON DRAFT REPORT

Any interested party may submit comments on this report for consideration by the NRC staff.

Comments may be accompanied by additional relevant information or supporting data. Please

specify the report number, NUREG-1556, Volume 4, Revision 1, draft, in your comments, and 

send them by the end of the comment period specified in the Federal Register notice 

announcing the availability of this draft to the following address:

Cindy Bladey, Chief

Rules, Announcements, and Directives Branch

Division of Administrative Services

Office of Administration

Mail Stop: TWB-05-B01M

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

For any questions about the material in this report, please contact:

Tomas Herrera

Mail Stop: T-8E24

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Phone: (301) 415-7138

E-mail: tomas.herrera@nrc.gov

Please be aware that any comments that you submit to the NRC will be considered a public

record and entered into the Agencywide Documents Access and Management System

(ADAMS). Do not provide information you would not want to be publically available.
iii

ABSTRACT

This technical report contains information intended to provide program-specific guidance and 

assist applicants and licensees in preparing applications for materials licenses for fixed gauges.

In particular, it describes the types of information needed to complete U.S. Nuclear Regulatory 

Commission (NRC) Form 313, “Application for Materials License.” This document describes 

both the methods acceptable to the NRC license reviewers in implementing the regulations and 

the techniques used by the reviewers in evaluating the application to determine if the proposed 

activities are acceptable for licensing purposes.

This NUREG contains information collection requirements that are subject to the Paperwork 

Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These information collection requirements 

were approved by the Office of Management and Budget, approval numbers 3150-0044, 

3150-0014, 3150-0017, 3150-0008, and 3150-0120.

The NRC may not conduct or sponsor, and a person is not required to respond to, a request for 

information or an information collection requirement unless the requesting document displays a 

currently valid OMB control number.

Paperwork Reduction Act Statement

Public Protection Notification
v

FOREWORD

The U.S. Nuclear Regulatory Commission’s (NRC’s) NUREG-1556 technical report series 

provides a comprehensive source of reference information about various aspects of materials 

licensing and materials program implementation. These reports, where applicable, describe a 

risk-informed, performance-based approach to licensing consistent with the current regulations. 

The reports are intended for use by applicants, licensees, license reviewers, and other NRC 

personnel. The NUREG-1556 series currently includes the following volumes:

Volume No. Volume Title

1 Program-Specific Guidance about Portable Gauge Licenses

2 Program-Specific Guidance about Industrial Radiography Licenses

3 Applications for Sealed Source and Device Evaluation and Registration

4 Program-Specific Guidance about Fixed Gauge Licenses

5 Program-Specific Guidance about Self-Shielded Irradiator Licenses

6 Program-Specific Guidance about 10 CFR Part 36 Irradiator Licenses 

7 Program-Specific Guidance about Academic, Research and Development, and 

Other Licenses of Limited Scope Including Gas Chromatographs and X-Ray 

Fluorescence Analyzers

8 Program-Specific Guidance about Exempt Distribution Licenses

9 Program-Specific Guidance about Medical Use Licenses

10 Program-Specific Guidance about Master Materials Licenses

11 Program-Specific Guidance about Licenses of Broad Scope

12 Program-Specific Guidance about Possession Licenses for Manufacturing and 

Distribution

13 Program-Specific Guidance about Commercial Radiopharmacy Licenses

14 Program-Specific Guidance about Well Logging, Tracer, and Field Flood Study 

Licenses

15 Guidance about Changes of Control and about Bankruptcy Involving Byproduct, 

Source, or Special Nuclear Materials Licenses

16 Program-Specific Guidance about Licenses Authorizing Distribution to General 

Licensees
vi

Volume No. Volume Title

17 Program-Specific Guidance about Special Nuclear Material of Less Than 

Critical Mass Licenses

18 Program-Specific Guidance about Service Provider Licenses

19 Guidance for Agreement State Licensees about NRC Form 241 “Report of 

Proposed Activities in Non-Agreement States, Areas of Exclusive Federal 

Jurisdiction, or Offshore Waters” and Guidance for NRC Licensees Proposing to 

Work in Agreement State Jurisdiction (Reciprocity)

20 Program-Specific Guidance about Administrative Licensing Procedures

21 Program-Specific Guidance about Possession Licenses for Production of 

Radioactive Materials Using an Accelerator

22 Reserved

The current document, NUREG-1556, Volume 4, Revision 1, “Consolidated Guidance about 

Materials Licenses: Program-Specific Guidance about Fixed Gauge Licenses,” is intended for 

use by applicants, licensees, and NRC staff. This revision provides a general update to the 

previous information contained in NUREG-1556, Volume 4, issued October 1998.

This report takes a risk-informed, performance-based approach to licensing fixed gauges. 

A team composed of staff from NRC Headquarters, NRC regional offices, and Agreement 

States prepared this document, drawing on their collective experience in radiation safety in 

general and as specifically applied to fixed gauges.

NUREG-1556, Volume 4, Revision 1, is not a substitute for NRC regulations. The approaches 

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

different from those described in this report may be acceptable if they include a basis for the 

staff to make the determinations needed to issue or continue a license.

________________________________________

Brian J. McDermott, Director

Division of Materials Safety and State Agreements

Office of Federal and State Materials and Environmental Management Programs
vii

CONTENTS

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

FOREWORD ..............................................................................................................................v

ACKNOWLEDGMENTS............................................................................................................xi

ABBREVIATIONS ................................................................................................................... xiii

1. PURPOSE OF REPORT......................................................................................................1

2. AGREEMENT STATES .......................................................................................................3

3. MANAGEMENT RESPONSIBILITY.....................................................................................7

3.1 Commitments and Responsibilities..............................................................................7

3.2 Safety Culture..............................................................................................................8

4. APPLICABLE REGULATIONS..........................................................................................11

5. HOW TO FILE....................................................................................................................13

5.1 Paper Application ......................................................................................................13

5.2 Where to File .............................................................................................................13

5.3 Transfer to Electronic Format ....................................................................................14

6. IDENTIFYING AND PROTECTING SENSITIVE INFORMATION ......................................15

7. APPLICATION AND LICENSE FEES................................................................................17

8. CONTENTS OF AN APPLICATION ..................................................................................19

8.1 Item 1: License Action Type......................................................................................20

8.2 Item 2: Name and Mailing Address of Applicant........................................................20

8.2.1 Notification of Bankruptcy Proceedings ........................................................21

8.3 Item 3: Address(es) Where Licensed Material Will be Used or Possessed ...............21

8.4 Item 4: Person to Be Contacted about This Application ............................................22

8.5 Item 5: Radioactive Material .....................................................................................22

8.5.1 Sealed Sources and Devices ......................................................................22

8.5.2 Financial Assurance and Recordkeeping for Decommissioning ..................23

8.6 Item 6: Purpose(s) for Which Licensed Material Will Be Used ..................................25

8.7 Item 7: Individual(s) Responsible for Radiation Safety Program and Their 

 Training and Experience ...............................................................................26

8.7.1 Radiation Safety Officer ..............................................................................26

8.7.2 Authorized Users.........................................................................................28

8.8 Item 8: Training for Individuals who in the Course of Employment Are Likely To

 Receive Occupational Doses of Radiation in Excess of 1 Millisievert 

 (100 Millirem) in a Year (Occupationally Exposed Workers) and Ancillary

 Personnel .....................................................................................................29

8.9 Item 9: Facilities and Equipment..............................................................................29

8.10 Item 10: Radiation Safety Program...........................................................................30

8.10.1 Audit Program .............................................................................................30
viii

8.10.2 Instruments .................................................................................................31

8.10.3 Material Receipt and Accountability.............................................................33

8.10.4 Occupational Dose......................................................................................34

8.10.5 Public Dose.................................................................................................36

8.10.6 Operating and Emergency Procedures .......................................................39

8.10.7 Leak Tests ..................................................................................................44

8.10.8 Maintenance ...............................................................................................45

8.10.9 Transportation.............................................................................................49

8.10.10 Fixed Gauges Used at Temporary Job Sites ...............................................50

8.10.11 Security Program .........................................................................................52

8.11 Item 11: Waste Management—Gauge Disposal and Transfer ..................................53

8.12 Item 12: License Fees ..............................................................................................54

8.13 Item 13: Certification.................................................................................................54

9. AMENDMENTS AND RENEWALS TO A LICENSE..........................................................55

9.1 Timely Notification of Transfer of Control...................................................................55

10. APPLICATIONS FOR EXEMPTIONS................................................................................57

11. TERMINATION OF ACTIVITIES........................................................................................59

APPENDICES

 APPENDIX A U.S. NUCLEAR REGULATORY COMMISSION FORM 313...........A-1

APPENDIX B SUGGESTED FORMAT FOR PROVIDING 

INFORMATION REQUESTED IN ITEMS 5 THROUGH 11 OF 

U.S. NUCLEAR REGULATORY COMMISSION FORM 313...........B-1

APPENDIX C TYPICAL DUTIES AND RESPONSIBILITIES OF 

RADIATION SAFETY OFFICER.....................................................C-1

APPENDIX D CRITERIA FOR ACCEPTABLE TRAINING FOR AUTHORIZED 

USERS AND RADIATION SAFETY OFFICERS.............................D-1

APPENDIX E SUGGESTED FIXED GAUGE AUDIT CHECKLIST........................E-1

APPENDIX F MODEL SURVEY INSTRUMENT CALIBRATION PROGRAM ....... F-1

APPENDIX G DOSIMETRY-RELATED GUIDANCE.............................................G-1

APPENDIX H OPERATING AND EMERGENCY PROCEDURES ........................H-1

APPENDIX I MODEL LEAK TEST PROGRAM .................................................... I-1

APPENDIX J INFORMATION NEEDED TO SUPPORT APPLICAN’TS 

REQUEST TO PERFORM NONROUTINE OPERATIONS..............J-1

APPENDIX K APPLICABLE U.S. DEPARTMENT OF TRANSPORTATION 

REGULATIONS..............................................................................K-1

APPENDIX L NRC INCIDENT NOTIFICATIONS...................................................L-1

APPENDIX M INTERIM STAFF GUIDANCE ON CONSTRUCTION .....................M-1

APPENDIX N SAFETY CULTURE STATEMENT OF POLICY..............................N-1
ix

FIGURES

Figure 1.1 Examples of several different types of fixed gauges...............................................1

Figure 2.1 U.S. Map: Locations of NRC offices and Agreement States..................................3

Figure 8.1 Typical duties and responsibilities of RSOs .........................................................27

Figure 8.2 Annual dose limits for radiation workers...............................................................35

Figure 8.3 Limiting public dose .............................................................................................38

Figure 8.4 Proper handling of incident ..................................................................................42

Figure 8.5 Lock-out procedures ............................................................................................43

Figure 8.6 Maintenance ........................................................................................................46

Figure 8.7 DOT transportation requirements.........................................................................49

Figure 8.8 Examples of uses for fixed gauges at temporary job sites. ....................................... 50

Figure 8.9 Security................................................................................................................51

Figure 8.10 Installation of fixed gauges at temporary job sites................................................52

Figure G.1 Diagram of bottling line and fixed gauges ..........................................................G-4

TABLES

Table 2.1 Who Regulates the Activity? ..................................................................................4

Table 3.1 Traits of a Positive Safety Culture..........................................................................9

Table 8.1 Examples of Minimum Inventory Quantities Requiring Financial Assurance ........24

Table 8.2 Record Maintenance............................................................................................34

Table G.1 Dosimetry Evaluation.........................................................................................G-2

Table G.2 Information Known about Each Gauge...............................................................G-5

Table G.3 Calculational Method, Part 1 – Hourly and Annual Dose Received from 

Gauge 1.............................................................................................................G-5

Table G.4 Calculational Method, Part 1 – Hourly and Annual Dose Received from 

Gauge 2.............................................................................................................G-6

Table G.5 Calculational Method, Part 1 – Hourly and Annual Dose Received from 

Gauge 3.............................................................................................................G-6

Table G.6 Calculational Method, Part 1 – Total Hourly and Annual Dose Received 

From Gauges 1, 2, and 3 ...................................................................................G-6

Table G.7 Calculational Method, Part 2 – Annual Dose Received from Gauges 1, 2, 

and 3 .................................................................................................................G-7

Table G.8 Calculational Method, Part 3 – Summary of Information ....................................G-8

Table G.9 Calculational Method, Part 3 – Annual Dose Received from Gauges 1, 2, 

and 3 .................................................................................................................G-8

Table G.10 Combination Measurement – Calculational Method .........................................G-10

Table L.1 Typical NRC Incident Notifications Required for Fixed Gauge Licensees............L-1
xi

ACKNOWLEDGMENTS

The working group thanks the individuals listed below for assisting in the review and update of 

the report. All participants provided valuable insights, observations, and recommendations.

The working group would like to thank the staff in the regional offices of the NRC and all of the 

States who provided comments and technical information that assisted in the development of 

this report.

The working group also thanks Lisa Dimmick, John O’Donnell, Monica Orendi, Tara Weidner, 

and Duane White for developing the formatting and language used in many parts of the report.

The Participants for this Revision:

Cook, Jackie

Juran, Brandon

McCraw, Aaron

Reichard, Michael

Rivera-Capella, Gretchen 
xiii

ABBREVIATIONS

ALARA as low as is reasonably achievable

ANSI American National Standards Institute

AU authorized user

CFR Code of Federal Regulations

DOT United States Department of Transportation

FSME Office of Federal and State Materials and Environmental Management Programs

G-M Geiger-Mueller

hazmat hazardous materials

IAEA International Atomic Energy Agency

IN Information Notice

MDA minimum detectable activity

mrem millirem

mSv millisievert

NIST National Institute of Standards and Technology

NRC United States Nuclear Regulatory Commission

NSTS National Source Tracking System

NVLAP National Voluntary Laboratory Accreditation Program

OMB Office of Management and Budget

R Roentgen

Rev. revision

RIS regulatory issue summary

RQ reportable quantity

RSO radiation safety officer

SSD sealed source and device

TI transportation index

U.S.C. United States Code
1

1. PURPOSE OF REPORT

This report provides guidance to an applicant applying for a fixed gauge license and provides

the U.S. Nuclear Regulatory Commission (NRC) criteria for evaluating a fixed gauge license 

application. It is not intended to address the research and development of fixed gauges or the 

commercial aspects of manufacturing, distributing, and non-routine servicing of such devices. 

Within this document, the phrases or terms “fixed gauge,” “gauging devices,” or “gauges” are 

used interchangeably.

This report addresses a variety of radiation safety issues associated with fixed gauges of many 

designs. Typically, gauges are used for process control (e.g., to measure the thickness of 

paper, the density of coal, the level of material in vessels and tanks, and volumetric flow rate). 

Figure 1.1 illustrates various examples of different types of fixed gauges. Because of 

differences in design, manufacturers provide appropriate instructions and recommendations for 

proper operation and maintenance. In addition, with gauges of varying designs, the sealed 

sources may be oriented in different locations within the devices, resulting in different radiation 

safety problems.

Figure 1.1 Examples of several different types of fixed gauges

Chapter 8, “Contents of an Application,” of this report identifies the information needed to 

complete NRC Form 313, “Application for Material License” (see Appendix A), for the use of 

sealed sources in fixed gauges. The Office of Management and Budget (OMB) has approved 

the information collection requirements in Title 10 of the Code of Federal Regulations

(10 CFR) Part 30, “Rules of General Applicability to Domestic Licensing of Byproduct Material,”

and NRC Form 313 under OMB Clearance Nos. 3150-0017, and 3150-0120, respectively.

The format within this document for each item of technical information is as follows:

• Regulations—references the regulations applicable to the item

• Criteria—outlines the criteria used to evaluate the applicant’s response

• Discussion—provides additional information about the topic

• Response from Applicant—provides suggested response or responses, offers the option 

of an alternative reply, or indicates that no response is needed on that topic during the 

licensing process
2

Notes and references are self-explanatory and may not be found for each item on NRC 

Form 313.

NRC Form 313 does not have sufficient space for applicants to provide full responses to Items 5 

through 11, as indicated on the form. Applicants should address those items on separate 

sheets of paper and submit them along with the completed NRC Form 313. For the 

convenience and streamlined handling of fixed gauge applications, Appendix B, “Suggested 

Format for Providing Information Requested in Items 5 through 11 of U.S. Nuclear Regulatory 

Commission Form 313,” may be used to provide supporting information.
3

2. AGREEMENT STATES

Certain States, called Agreement States (see Figure 2.1), have entered into agreements with 

the NRC that give them the authority to license and inspect byproduct, source, and special 

nuclear materials, in quantities not sufficient to form a critical mass, which are used or 

possessed within their borders. Any applicant, other than a Federal entity, who wishes to 

possess or use licensed material in one of these Agreement States should contact the 

responsible officials in that State for guidance on preparing an application. These applications 

should be filed with State officials, not with the NRC. In areas under exclusive federal 

jurisdiction within an Agreement State, NRC continues to be the regulatory authority.

Figure 2.1 U.S. map: Locations of NRC offices and Agreement States

In the special situation of work at Federally controlled sites in Agreement States, it is necessary 

to ascertain the jurisdictional status of the land to determine whether the NRC or the Agreement 

State has regulatory authority. These areas can also include tribal lands of federally recognized 

Indian Tribes.

1

 1

 For the purposes of this guidance, an “Indian tribe” is defined as an Indian or Alaska Native tribe, band, nation,

 The NRC has regulatory authority over land determined to be “exclusive Federal 

jurisdiction,” while the Agreement State has jurisdiction over nonexclusive Federal jurisdiction 

land. Applicants are responsible for determining in advance the jurisdictional status of the 

 pueblo, village, or community that the Secretary of the Interior acknowledges to exist as an Indian tribe pursuant to

 the Federally Recognized Indian Tribe List Act of 1994. A list of federally recognized tribes is available at

 www.bia.gov.
4

specific areas where they plan to conduct licensed operations. The NRC recommends that 

applicants contact their local office of the Federal agency controlling the site (e.g., contract 

officer, base environmental health officer, district office staff) for assistance in determining the 

jurisdictional status of the land and to provide the information in writing to ensure compliance

with NRC or Agreement State regulatory requirements, as appropriate. Additional guidance on 

determining jurisdictional status is found in the All Agreement States Letter (SP-96-022), dated 

February 16, 1996, which is available at http://nrc-stp.ornl.gov/. Once on the Web site, use the 

link for “FSME Letters” in the left hand column under “Resources & Tools.” The link will take 

you to another Web page where you can search for letters to the Agreement States.

Table 2.1 provides a quick way to check on whether the NRC or an Agreement State has 

regulatory authority.

Table 2.1 Who Regulates the Activity?

Applicant and Proposed Location of Work Regulatory 

Agency

Federal agency regardless of location (except the U.S. Department of 

Energy and, under most circumstances, its prime contractors are exempt 

from licensing, in accordance with 10 CFR 30.12, “Persons using byproduct 

material under certain Department of Energy and Nuclear Regulatory 

Commission contracts”)

NRC

Non-Federal entity in non-Agreement State, District of Columbia, 

U.S. territory or possession, or in offshore Federal waters NRC

Federally recognized Indian Tribe or tribal member on Indian Tribal land NRC

Non-federal entity on Federally recognized Indian Tribal land NRC2

Federally recognized Indian Tribe or tribal member outside of Indian Tribal 

land in Agreement State. Agreement State

 2

 The NRC can exercise jurisdiction as the regulatory authority on tribal land of a Federally recognized Indian Tribe. 

 Section 274b. Agreements do not give States the authority to regulate nuclear material in these areas. However,

 there are few States that exercise regulatory authority over these areas based on treaties or agreements with

 specific tribes. Companies owned or operated by Federally recognized Indian Tribe members or non-Indians that

 wish to possess or use licensed material on tribal lands should contact the appropriate NRC regional office to

 determine the jurisdictional status of the tribal lands and identify the appropriate regulatory agency for licensing and

 reciprocity.
5

Applicant and Proposed Location of Work Regulatory 

Agency

Non-Federal entity in Agreement State Agreement 

State3

Non-Federal entity in Agreement State at Federally controlled site not 

subject to exclusive Federal jurisdiction

Agreement 

State3

Non-Federal entity in Agreement State at Federally controlled site subject to 

exclusive Federal jurisdiction NRC

Non-Federal entity in Agreement State using radioactive materials (except 

industrial radiography) directly connected with Part 50 or 52 reactor 

operations or needed during the construction and preoperational phases of a 

reactor. 

NRC

Non-Federal entity in Agreement State using radioactive materials not 

directly connected with Part 50 or 52 reactor operations or needed during the 

construction and preoperational phases of a reactor.

Agreement 

State3

Reference: A current list of Agreement States (including names, addresses, and telephone 

numbers of responsible officials) is available at the Office of Federal and State Materials and 

Environmental Management Programs’ public Web site, http://nrc-stp.ornl.gov. As an 

alternative, a request for the list can be made to an NRC regional office.

 3

 Section 274m. of the AEA gives the NRC regulatory authority over radioactive materials covered under the Section

 274b. Agreement when the activity can affect the Commission’s authority to protect the common defense and

 security, to protect restricted data, or guard against the loss or diversion of special nuclear material at a site. (This

 is an uncommon situation which NRC usually evaluates on a case-by-case basis.) Companies that wish to possess

 or use licensed material at these sites should contact the licensee to determine the jurisdictional status for specific

 AEA radioactive materials they intend to possess or use at the site.

7

3. MANAGEMENT RESPONSIBILITY

The NRC recognizes that effective radiation safety program management is vital to achieving 

safe, secure, and compliant operations. Consistent compliance with NRC regulations provides 

reasonable assurance that licensed activities will be conducted safely and that effective 

management will result in increased safety, security, and compliance.

“Management” as used in this volume refers to the processes for conduct and control of a 

radiation safety program and to the individuals who are responsible for those processes and 

who have authority to provide necessary resources to achieve regulatory compliance.

3.1 Commitments and Responsibilities

Pursuant to 10 CFR 30.32(c), each application shall be signed by the applicant or licensee or a 

person duly authorized to act for and on the behalf of the applicant or licensee. If it is not clear 

whether the application was signed by someone duly authorized to act for and on the behalf of 

the applicant or licensee, NRC license reviewers may ask for additional assurances that the 

individual that signed the application is duly authorized to act for and on the behalf of the 

applicant or licensee. The signature on an application acknowledges the licensee’s 

commitments and responsibilities for the following:

• Radiation safety, security, and control of radioactive materials and compliance with 

regulations;

• Completeness and accuracy of the radiation safety records and all information provided 

to the NRC (10 CFR 30.9, “Completeness and accuracy of information”);

• Knowledge about the contents of the license and application;

• Compliance with current NRC and U.S. Department of Transportation (DOT) regulations 

and the licensee’s operating, emergency, and security procedures;

• Commitment to provide adequate resources (including space, equipment, personnel, 

time, and, if needed, contractors) to the radiation protection program to ensure that the 

public and workers are protected from radiation hazards and compliance with regulations 

is maintained;

• Selection and assignment of a qualified individual to serve as the radiation safety officer 

(RSO) for licensed activities and confirmation that the RSO has independent authority to 

stop unsafe operations and will be given sufficient time to fulfill radiation safety duties 

and responsibilities;

• Commitment to ensure that radiation workers have adequate training;

• Prevention of discrimination of employees engaged in protected activities (10 CFR 30.7, 

“Employee protection”);
8

• Commitment to provide information to employees about the employee protection and 

deliberate misconduct provisions in 10 CFR 30.7 and 10 CFR 30.10, “Deliberate 

misconduct,” respectively;

• Commitment to obtain the NRC’s prior written consent before transferring control of the 

license (see Section 9.1, “Timely Notification of Transfer of Control,” of this report); and

• Notification of the appropriate NRC regional administrator in writing, immediately 

following the filing of a petition for voluntary or involuntary bankruptcy (10 CFR 30.34(h)), 

as discussed further in Section 8.2.1, “Notification of Bankruptcy Proceedings,” of this 

report.

For information on NRC inspection, investigation, enforcement, and other compliance programs, 

see the current version of the NRC’s Enforcement Policy and Inspection Procedures available in 

the NRC’s online library at http://www.nrc.gov/reading-rm.html. 

3.2 Safety Culture

Individuals and organizations performing regulated activities are expected to establish and 

maintain a positive safety culture commensurate with the safety and security significance of 

their activities and the nature and complexity of their organizations and functions. This applies 

to all licensees, certificate holders, permit holders, authorization holders, holders of quality 

assurance program approvals, vendors and suppliers of safety-related components, and 

applicants for a license, certificate, permit, authorization, or quality assurance program approval, 

subject to NRC authority.

“Nuclear safety culture” is defined in the Safety Culture Policy Statement as the core values and 

behaviors resulting from a collective commitment by leaders and individuals to emphasize 

safety over competing goals to ensure protection of people and the environment. Individuals 

and organizations performing regulated activities bear the primary responsibility for safely 

handling and securing these materials. Experience has shown that certain personal and 

organizational traits are present in a positive safety culture. A trait, in this case, is a pattern of 

thinking, feeling, and behaving that emphasizes safety, particularly in goal conflict situations 

(e.g., production versus safety, schedule versus safety, and cost of the effort versus safety). 

Refer to Table 3.1 for the traits of a positive safety culture from NRC’s safety culture policy 

statement.

The NRC, as the regulatory agency with an independent oversight role, reviews the 

performance of individuals and organizations to determine compliance with requirements and 

commitments through its existing inspection and assessment processes. However, NRC’s 

safety culture policy statement and traits are not incorporated into the regulations. Many of the 

safety culture traits may be inherent to an organization’s existing radiation safety practices and 

programs. For instance, many fixed gauges include shutters that allow the source to be 

shielded. If the fixed gauge shutter fails and cannot be closed, licensees are required to notify 

the NRC Headquarters Operations Center within 24 hours of discovery of an event involving 

licensed material in which equipment is disabled or fails to function when it is required to be 

available and operable to prevent exposures exceeding regulatory limits. The need to 

recognize an equipment failure and make a 24-hour report may correspond with the “Problem 

Identification and Resolution” (issues potentially impacting safety are promptly identified, fully 
9

evaluated, and promptly addressed and corrected commensurate with their significance) safety 

culture trait.

Refer to Appendix N for the NRC’s safety culture policy statement. More information on NRC 

activities relating to safety culture can be found at: http://www.nrc.gov/aboutnrc/regulatory/enforcement/safety-culture.html.

Table 3.1 Traits of a Positive Safety Culture

Leadership Safety Values 

and Actions 

Problem Identification and 

Resolution Personal Accountability 

Leaders demonstrate a 

commitment to safety in their 

decisions and behaviors 

Issues potentially impacting 

safety are promptly 

identified, fully evaluated, 

and promptly addressed and 

corrected commensurate 

with their significance 

All individuals take personal 

responsibility for safety 

Work Processes Continuous Learning 

Environment for Raising 

Concerns 

The process of planning and 

controlling work activities is 

implemented so that safety 

is maintained 

Opportunities to learn about 

ways to ensure safety are 

sought out and implemented 

A safety conscious work 

environment is maintained 

where personnel feel free to 

raise safety concerns without 

fear of retaliation, 

intimidation, harassment or 

discrimination 

Effective Safety 

Communications 

Respectful Work 

Environment Questioning Attitude 

Communications maintain a 

focus on safety 

Trust and respect permeate 

the organization 

Individuals avoid 

complacency and continually 

challenge existing conditions 

and activities in order to 

identify discrepancies that 

might result in error or 

inappropriate action 

11

4. APPLICABLE REGULATIONS

It is the applicant’s or licensee’s responsibility to obtain and have available up-to-date copies of 

applicable regulations, to read and understand the requirements of each of these regulations, 

and to comply with each applicable regulation. The following parts of Title 10 of the Code of 

Federal Regulations (10 CFR) contain regulations applicable to fixed gauges. These parts will 

apply to many, if not all, licensees.

The current versions of these parts can be found under the “Basic References” link at the NRC’s 

online library at http://www.nrc.gov/reading-rm.html; if viewing in a browser, the following list 

includes direct links to the rules:

• 10 CFR Part 2, “Rules of Practice for Domestic Licensing Proceedings and Issuance of 

Orders”

• 10 CFR Part 19, “Notices, Instructions and Reports to Workers: Inspection and 

Investigations”

• 10 CFR Part 20, “Standards for Protection Against Radiation”

• 10 CFR Part 21, “Reporting of Defects and Noncompliance”

• 10 CFR Part 30, “Rules of General Applicability to Domestic Licensing of Byproduct 

Material”

• 10 CFR Part 31, “General Domestic Licenses for Byproduct Material”

• 10 CFR Part 32, “Specific Domestic Licenses to Manufacture or Transfer Certain Items 

Containing Byproduct Material”

• 10 CFR Part 71, “Packaging and Transportation of Radioactive Material”

In 10 CFR Part 71, the NRC requires that licensees or applicants who transport licensed 

material, or who may offer such material to a carrier for transport, must comply with the 

applicable DOT requirements that are found in 49 CFR Parts 170 through 189.

• 10 CFR Part 150, “Exemptions and Continued Regulatory Authority in Agreement States 

and in Offshore Waters under Section 274”

• 10 CFR Part 170, “Fees for Facilities, Materials, Import and Export Licenses, and Other 

Regulatory Services Under the Atomic Energy Act of 1954, as Amended”

• 10 CFR Part 171, “Annual Fees for Reactor Licenses, and Fuel Cycle Licenses and 

Materials Licenses, Including Holders of Certificates of Compliance, Registrations, and 

Quality Assurance Program Approvals and Government Agencies Licensed by the NRC”

Copies of the above documents may be obtained by calling the Government Printing Office 

order desk toll-free at (866) 512-8600, in Washington, DC, at (202) 512-1800 or online at 

http://bookstore.gpo.gov.
12

A single copy of the above documents may be requested from the NRC’s regional offices (see 

Figure 2.1 for addresses and telephone numbers). In addition, 10 CFR Parts 1 through 199 can 

be found on the NRC’s Web site at http://www.nrc.gov/reading-rm/doc-collections/ under 

“Regulations (10 CFR).” 

NRC regulations and amendments can also be accessed from the “NRC Library” link on the 

NRC’s public Web site at http://www.nrc.gov. The NRC and all other Federal agencies publish 

amendments to their regulations in the Federal Register.
13

5. HOW TO FILE

5.1 Paper Application

Applicants for a materials license should do the following:

• Use the most recent guidance in preparing an application.

• Complete NRC Form 313 (Appendix A) Items 1 through 4, 12, and 13 on the form itself.

• Complete NRC Form 313 Items 5 through 11 on supplementary pages or use 

Appendix B.

• Provide sufficient detail for the NRC to determine that equipment, facilities, training, 

experience, and the radiation safety program are adequate to protect health and safety 

and minimize danger to life and property.

• For each separate sheet other than NRC Form 313 and Appendix B submitted with the 

application, identify and cross-reference submitted information to the item number on the 

application or the topic to which it refers.

• Submit all documents, typed, on 8-1/2 x 11-inch paper.

• Avoid submitting proprietary information and personally identifiable information.

• If submitted, proprietary information and other sensitive information (e.g., personal 

privacy and security related) should be clearly identified per 10 CFR 2.390, “Public 

inspections, exemptions, requests for withholding” (see Chapter 6, “Identifying and 

Protecting Sensitive Information”).

• Submit an original, signed application.

• Retain one copy of the license application for future reference.

Applications must be signed by the applicant, licensee, or a person duly authorized as 

required by 10 CFR 30.32(c) (see Section 8.13, “Certification”).

5.2 Where to File

Applicants wishing to possess or use licensed material in any State, U.S. territory, or U.S. 

possession subject to NRC jurisdiction must file an application with the NRC regional office for 

the locale in which the material will be possessed or used. Figure 2.1 identifies the NRC’s four 

regional offices and their respective areas for licensing purposes and the Agreement States. 

Note that all materials applications are submitted to Regions I, III, or IV. All applicants for 

materials licenses located in the Region II geographical area should send their applications to 

Region I.

In general, applicants wishing to possess or use licensed material in Agreement States must file 

an application with the Agreement State and not with the NRC. However, if work will be 

conducted at federally controlled sites, or federally recognized Indian Tribal lands, in Agreement 
14

States, applicants must first determine the jurisdictional status of the land in order to determine 

whether the NRC or the Agreement State has regulatory authority. See Chapter 2, “Agreement 

States,” for additional information.

5.3 Transfer to Electronic Format

Paper applications received by the NRC are scanned through an optical character reader and 

converted to an electronic format. To ensure a smooth transfer to an electronic format, 

applicants should do the following:

• Submit printed or typewritten—not handwritten—text on smooth, crisp paper that will 

feed easily into the scanner.

• Choose typeface designs that are sans serif, such as Arial, Helvetica, Future, or Univers 

(the text of this document is in the Arial font).

• Use 12-point or larger font.

• Avoid stylized characters, such as script or italics.

• Ensure that the print is clear and sharp.

• Ensure that there is high contrast between the ink and paper (black ink on white paper is 

best).

It is anticipated that the NRC may provide mechanisms for filing applications through the 

Internet. If this occurs, the NRC will provide additional filing instructions as the agency 

implements these new mechanisms.
15

6. IDENTIFYING AND PROTECTING SENSITIVE INFORMATION

All licensing applications, except for portions containing sensitive information, will be made 

available for review in the NRC’s Public Document Room and electronically at the NRC Library. 

For more information on the NRC Library, visit www.nrc.gov.

The licensee should identify, mark, and protect sensitive information against unauthorized 

disclosure to the public. Licensing applications that contain sensitive information should be 

marked as indicated below in accordance with 10 CFR 2.390 before the information is submitted 

to the NRC. Key examples are as follows:

• Proprietary Information/Trade Secrets: If it is necessary to submit proprietary 

information or trade secrets, follow the procedure in 10 CFR 2.390(b). Failure to follow 

this procedure could result in disclosure of the proprietary information to the public or 

substantial delays in processing the application.

• Personally Identifiable Information: Personally identifiable information (PII) about 

employees or other individuals should not be submitted unless specifically requested by 

the NRC. Examples of PII are social security number, home address, home telephone 

number, date of birth, and radiation dose information. If PII is submitted, a cover letter 

should clearly state that the attached documents contain PII and the top of every page of 

a document that contains PII should be clearly marked as follows: “Privacy Act 

Information—Withhold Under 10 CFR 2.390.” For further information, see Regulatory 

Issue Summary (RIS) 2007-04, “Personally Identifiable Information Submitted to the 

U.S. Nuclear Regulatory Commission,” dated March 9, 2007, which can be found on the 

NRC’s Generic Communications webpage under “Regulatory Issue Summaries”: 

http://www.nrc.gov/reading-rm/doc-collections/gen-comm/.

• Security-Related Information: Following the events of September 11, 2001, the NRC 

changed its procedures to avoid release of information that terrorists could use to plan or 

execute an attack against facilities or citizens in the United States. As a result, certain 

types of information are no longer routinely released and are treated as sensitive 

unclassified information. For example, certain information about the quantities and 

locations of radioactive material at licensed facilities, and associated security measures, 

are no longer released to the public. Therefore, a cover letter should clearly state that 

the attached documents contain sensitive security-related information and the top of 

every page of a document that contains such information should be clearly marked: 

“Security Related—Withhold under 10 CFR 2.390.” For the pages having securityrelated sensitive information, an additional marking should be included (e.g. an editorial 

note box) adjacent to that material. For further information, see RIS 2005-31, “Control of 

Security-Related Sensitive Unclassified Non-Safeguards Information Handled by 

Individuals, Firms, and Entities Subject to NRC Regulation of the Use of Source, 

Byproduct, and Special Nuclear Material,” dated December 22, 2005, which can be 

found on the NRC’s Generic Communications webpage under “Regulatory Issue 

Summaries”: http://www.nrc.gov/reading-rm/doc-collections/gen-comm/. Additional 

information on procedures and any updates is available at http://www.nrc.gov/readingrm/sensitive-info.html.

17

7. APPLICATION AND LICENSE FEES

Each application for which a fee is specified must be accompanied by the appropriate fee. 

Refer to 10 CFR 170.31, “Schedule of fees for materials licenses and other regulatory services, 

including Inspections, and import and export licenses,” to determine the amount of the fee. The 

NRC will not issue a license until the fee is received. Consult 10 CFR 170.11, “Exemptions,” for 

information on exemptions from these fees. Once the technical review has begun, no fees will 

be refunded; application fees will be charged regardless of the NRC’s disposition of an 

application or the withdrawal of an application.

Most NRC licensees are also subject to annual fees; refer to 10 CFR 171.16, “Annual fees: 

Materials licensees, holders of certificates of compliance, holders of sealed source and device 

registrations, holders of quality assurance program approvals, and government agencies 

licensed by the NRC.” Consult 10 CFR 171.11 for information on exemptions from annual fees 

and 10 CFR 171.16(c) on reduced annual fees for licensees that qualify as “small entities.”

Direct all questions about the NRC’s fees or completion of Item 12 of NRC Form 313 to the 

Office of the Chief Financial Officer at NRC Headquarters in Rockville, MD, (301) 415-7554. 

Information about fees may also be obtained by calling NRC’s toll-free number, (800) 368-5642, 

extension 415-7554. The e-mail address is Fees.Resource@nrc.gov.

19

8. CONTENTS OF AN APPLICATION

The following comments apply to the indicated items on NRC Form 313 (Appendix A).

All items in the application should be completed in enough detail for the NRC to determine that 

the proposed equipment, facilities, training and experience, and radiation safety program satisfy 

regulatory requirements and are adequate to protect public health and safety and minimize 

danger to life and property. Consideration shall be given, when developing the application, to 

the concepts of keeping exposure as low as is reasonably achievable (ALARA), minimizing 

contamination, and maintaining control of radioactive materials.

Refer to Appendix M for guidance regarding the definition of construction and the consideration 

of activities that can be performed by materials license applicants and potential applicants, and 

licensees before the NRC has concluded its environmental review of the proposed licensing 

action. The vast majority of materials licensing actions will meet the criteria in 10 CFR 51.22(c) 

for a categorical exclusion. This means that the licensing action will not require an 

environmental assessment or environmental impact statement in accordance with 10 CFR 

51.22(b), since the NRC has already determined that the licensing action does not have a 

significant impact on the environment. It is the applicant’s responsibility to review the guidance 

in Appendix M to determine whether the categorical exclusion applies to the licensing action.

10 CFR 20.1101(b) states: “The licensee shall use, to the extent practical, procedures and 

engineering controls based upon sound radiation protection principles to achieve occupational 

doses and doses to members of the public that are as low as is reasonably achievable 

(ALARA).” Regulatory Guide 8.10, “Operating Philosophy for Maintaining Occupational 

Radiation Exposures as Low as Is Reasonably Achievable,” discusses the ALARA concepts and 

philosophy. The application should document ALARA considerations, including establishing 

administrative action levels and monitoring programs.

10 CFR 20.1406, “Minimization of contamination,” requires applicants for licenses to describe 

how facility design and procedures for operation will minimize, to the extent practicable, 

contamination of the facility and the environment; facilitate eventual decommissioning; and 

minimize, to the extent practicable, the generation of radioactive waste. As with ALARA 

considerations, applicants should address concerns for all aspects of their programs.

10 CFR 20.1801, “Security of stored materials,” states that licensees shall secure from 

unauthorized removal or access licensed materials that are stored in controlled or unrestricted 

areas.

10 CFR 20.1802, “Control of material not in storage,” states that licensees shall control and 

maintain constant surveillance of licensed material that is in a controlled or unrestricted area 

and that is not in storage.

All information submitted to the NRC during the licensing process may be incorporated as part 

of the license and will be subject to review during inspection.
20

8.1 Item 1: License Action Type

Item 1 of NRC Form 313 states the following:

This is an application for (check appropriate item):

Type of Action License No.

[ ] A. New License Not Applicable

[ ] B. Amendment XX-XXXXX-XX

[ ] C. Renewal XX-XXXXX-XX

Check Box A for a new license request. Note that a pre-licensing visit may be required prior to 

issuance of the license. Also note that an initial security inspection may be conducted in 

accordance with NRC Inspection Manual Chapter 2800, “Materials Inspection Program,” before

issuance of the license.

Check Box B for an amendment to an existing license and provide the license number.

Check Box C for a renewal of an existing license and provide the license number.

Licensees are required to request and obtain an amendment to the license before making 

changes to their radiation safety programs. The following are examples of changes that require 

amendment:

• a change of RSO

• changes in the mailing address

• changes in the address(es) of storage locations

• changes in licensed material, including increases in the possession limit of byproduct 

material and adding new types of gauges

See “Amendments and Renewals to a License” in Chapter 9 of this report.

8.2 Item 2: Name and Mailing Address of Applicant

List the legal name of the applicant’s corporation or other legal entity with direct control over use 

of the radioactive material; a division or department within a legal entity may not be a licensee. 

An individual may be designated as the applicant only if the individual is acting in a private 

capacity and the use of the radioactive material is not connected with employment in a 

corporation or other legal entity. Provide the mailing address where correspondence should be 

sent. A post office box number is an acceptable mailing address.

Notify the NRC of changes in mailing address; these changes do not require a fee.
21

Note: The NRC must be notified before control of the license is transferred (see Section 9.1, 

“Timely Notification of Transfer of Control”) or when bankruptcy proceedings have been initiated

(see Section 8.2.1, “Notification of Bankruptcy Proceedings”). 

8.2.1 Notification of Bankruptcy Proceedings

Regulation: 10 CFR 30.34(h)

Criteria: Immediately following the filing of a voluntary or involuntary petition for bankruptcy for 

or against a licensee, the licensee must notify the appropriate NRC regional administrator in 

writing, identifying the bankruptcy court in which the petition was filed and the date of filing.

Discussion: Even though a licensee may have filed for bankruptcy, the licensee remains 

responsible for all regulatory requirements. The NRC must be notified when licensees are in 

bankruptcy proceedings in order to determine whether all licensed material is accounted for and 

adequately controlled and whether there are any public health and safety concerns (e.g.,

contaminated facility). The NRC shares the results of its determinations with other involved 

entities (e.g., trustee) so that health and safety issues can be resolved before bankruptcy 

actions are completed.

Response from Applicant: None is required at the time of application for a new license. 

Licensees must immediately notify the NRC in writing follow the filing of a voluntary or 

involuntary petition for bankruptcy by or against the licensee.

Reference: See NUREG-1556, Volume 15, “Consolidated Guidance about Materials Licenses:

Guidance about Changes of Control and about Bankruptcy Involving Byproduct, Source, or 

Special Nuclear Materials Licenses.”

8.3 Item 3: Address(es) Where Licensed Material Will Be Used or Possessed

Specify the street address, city, and State or other descriptive address (e.g., Highway 10, 5 

miles east of the intersection of Highway 10 and State Route 234, Anytown, State) for each 

facility. The descriptive address should be sufficient to allow an NRC inspector to find the 

facility location. A post office box address is not acceptable. In addition, applicants are 

encouraged to provide global positioning system coordinates, as appropriate. 

An NRC-approved license amendment is required before receiving, using, and storing licensed 

material at an address or location not included with the application or already listed on the 

license. This applies whether the gauge is an additional device or a relocation of an existing 

device.

To conduct operations at temporary job sites (i.e., locations where work is conducted for limited 

periods of time), refer to the section in Section 8.10.10 of this report called “Fixed Gauges Used 

at Temporary Job Sites.” That section offers examples of operations where fixed gauges might 

be used at temporary job sites and identifies information that should be provided to the NRC to 

support a request for these operations.

An NRC license does not relieve a licensee from complying with other applicable Federal, 

State, or local regulations (e.g., local zoning requirements).
22

Documents that give the exact location of use and storage for materials greater than or equal to 

International Atomic Energy Agency (IAEA) Category 2 quantities should be marked 

“Security-Related Information—Withhold under 10 CFR 2.390.” 

The applicant need not submit sketches or identify the specific location of the fixed gauge 

within the facility with the application. The NRC will review the acceptability of the gauge’s 

location during the inspection process.

Note: As discussed later in Section 8.5.2, “Financial Assurance and Recordkeeping for 

Decommissioning,” licensees must maintain permanent records describing where licensed 

material was used or stored while the license was in effect. This is important for making future 

determinations about the release of these locations for unrestricted use (e.g., before the license 

is terminated). For fixed gauge licensees, acceptable records are leak test records, sketches,

and written descriptions of specific locations where each gauge was used or stored and any 

information relevant to damaged devices or leaking radioactive sources.

8.4 Item 4: Person To Be Contacted about This Application

Identify the individual who can answer questions about the content of the application, and 

include a telephone number where the individual may be contacted. Also include business cell 

phone numbers and e-mail addresses. This individual, usually the RSO, will serve as the point 

of contact during the review of the application. If this individual is not a full-time employee of the 

licensed entity, his or her position and relationship to the licensee should be specified. The 

NRC should be notified if the person assigned to this function changes or if his or her telephone 

number, cell phone number, or e-mail address changes. Notification of a contact change is only 

in order to provide information and would not be considered an application for license 

amendment, unless the notification involves a change in the contact person who is also the 

RSO.

As indicated on NRC Form 313 (see Appendix A), Items 5 through 11 should be submitted on 

separate sheets of paper. Applicants may use Appendix B for this purpose and should note that 

using the suggested wording of responses and committing to use the model procedures in this 

report will facilitate the NRC’s review.

8.5 Item 5: Radioactive Material

8.5.1 Sealed Sources and Devices

Regulations: 10 CFR 30.32(g), 10 CFR 30.33(a)(2), and 10 CFR 32.210

Criteria: 

Applicants must provide the manufacturer’s name, model number, radionuclide, quantity, and 

nominal activity for each requested sealed source and the manufacturer’s name and model 

number for each device.

Licensees will be authorized to possess and use only those sealed sources and devices 

specifically approved and registered by the NRC or an Agreement State.
23

Discussion: The NRC and Agreement States perform safety evaluations of gauges before 

authorizing manufacturers to distribute fixed gauges to licensees. The safety evaluation is 

documented in a Sealed Source and Device (SSD) registration certificate.

Licensees may not make any changes to the sealed source, device, or source/device 

combination that would alter the description or specifications from those indicated in the 

respective registration certificates, without obtaining NRC’s prior permission in a license 

amendment. Such changes may necessitate a custom registration review, increasing the time 

needed to process a licensing action.

SSD registration certificates contain sections on “Conditions of Normal Use” and “Limitation and 

Other Considerations of Use.” These sections may include limitations derived from conditions 

imposed by the manufacturer or distributor, by particular conditions of use that would reduce the 

radiation safety of the device, or by circumstances unique to the sealed source or device. For 

example, the working life of the device or appropriate temperature and other environmental 

conditions may be specified. Except as specifically approved by the NRC, licensees are 

required to use gauges according to their respective SSD registration certificates. Accordingly, 

applicants should obtain a copy of the certificate and review it with the manufacturer or 

distributor, or with the NRC or the issuing Agreement State.

Fixed gauges could potentially be at or above Category 2 quantities described in the IAEA’s 

“Code of Conduct on the Safety and Security of Radioactive Sources.” Applicant/licensee 

information on manufacturers, model numbers, and possession limits is sensitive and should be 

marked accordingly (see Chapter 6, “Identifying and Protecting Sensitive Information”). 

Category 1 and Category 2 sources regulated by the NRC and Agreement States must be 

tracked in the National Source Tracking System (NSTS).

Response from Applicant:

• Identify each radionuclide and nominal activity in each gauge.

• Identify the manufacturer or distributor and model number of each type of gauge.

• State the total quantity of each type of gauge.

For existing licensees with Category 1 or 2 quantities of materials, mark the section related to 

the possession of sealed sources and devices that includes manufacturers, model numbers, 

and possession limits as follows: “Security-Related Information—Withhold Under 

10 CFR 2.390.”

Reference: For more information about the SSD registration process, see the current version 

of NUREG-1556, Volume 3, “Consolidated Guidance about Materials Licenses: Applications for 

Sealed Source and Device Evaluation and Registration.”

8.5.2 Financial Assurance and Recordkeeping for Decommissioning

Regulations: 10 CFR 30.34(b), 10 CFR 30.35

Criteria: Fixed gauge licensees authorized to possess sealed sources containing radioactive 

material in excess of the limits specified in 10 CFR 30.35, “Financial assurance and 
24

recordkeeping for decommissioning,” must provide evidence of financial assurance for 

decommissioning.

Licensees are required to maintain, in an identified location, decommissioning records related to 

leaking sources and to structures and equipment where gauges are used or stored. Licensees 

must transfer these records important to decommissioning either to the new licensee before 

licensed activities are transferred or assigned in accordance with 10 CFR 30.34(b) or to the 

appropriate NRC regional office before the license is terminated.

Discussion: The requirements for financial assurance are specific to the types and quantities 

of byproduct material authorized on a license. Most fixed gauge applicants and licensees do 

not need to take any action to comply with the financial assurance requirements because their 

total inventory of licensed material does not exceed the thresholds in 10 CFR 30.35(d). The 

thresholds for typical radionuclides used for fixed gauge sealed sources are shown in Table 8.1.

Table 8.1 Examples of Minimum Inventory Quantities Requiring Financial Assurance

Radionuclide

(Sealed Sources)

Activity in Gigabecquerels Activity in Curies

Cobalt-60 3.7 x 105 10,000

Krypton-85 3.7 x 107 1,000,000

Strontium-90 3.7 x 104 1,000

Cesium-137 3.7 x 106 100,000

Americium-241 3.7 x 103 100

Californium-252 3.7 x 103 100

A licensee would need to possess tens to hundreds of gauges before the financial assurance 

requirements would apply. Applicants and licensees desiring to possess gauges exceeding the 

threshold amounts must submit evidence of financial assurance. Licensees should follow the 

guidance provided in NUREG-1757, Volume 3, “Consolidated NMSS Decommissioning 

Guidance: Financial Assurance, Recordkeeping, and Timeliness.” 

Applicants requesting more than one radionuclide may determine whether financial assurance 

for decommissioning is required by calculating, for each radionuclide possessed, the ratio 

between the activity possessed, in curies, and the radionuclide’s threshold activity requiring 

financial assurance, in curies. If the sum of such ratios for all of the radionuclides possessed 

exceeds “1” (i.e., “unity”), then applicants must submit evidence of financial assurance for

decommissioning.

The same regulation also requires that licensees maintain records important to 

decommissioning in an identified location. All fixed gauge licensees need to maintain records of 

structures and equipment where each gauge was used or stored. As-built drawings with 

modifications of structures and equipment shown as appropriate fulfill this requirement. If 

drawings are not available, licensees shall substitute appropriate records (e.g., a sketch of the 
25

room or building or a narrative description of the area) concerning the specific areas and 

locations. If no records exist regarding structures and equipment where gauges were used or 

stored, licensees shall make all reasonable efforts to create such records based on historical 

information (e.g., employee recollections). In addition, if fixed gauge licensees have 

experienced unusual occurrences (e.g., leaking sources, other incidents that involve spread of 

contamination), they also need to maintain records about contamination that remains after 

cleanup or that may have spread to inaccessible areas.

For fixed gauge licensees whose sources have never leaked, acceptable records important to 

decommissioning are sketches or written descriptions of the specific locations where each 

gauge was used or stored.

Response from Applicant: No response is needed from most applicants. If financial 

assurance is required, submit the documentation required under 10 CFR 30.35.

Reference: See NUREG-1757, Volume 3, “Consolidated NMSS Decommissioning Guidance: 

Financial Assurance, Recordkeeping, and Timeliness.”

8.6 Item 6: Purpose(s) for Which Licensed Material Will Be Used

Regulation: 10 CFR 30.33(a)(1)

Criteria: An application for a license will be approved if the proposed activity is authorized by 

the Atomic Energy Act of 1954, as amended, and devices will be used only for the purposes for 

which they were designed and according to the manufacturer’s recommendations for use as 

specified in an approved SSD registration certificate.

Discussion: Uses other than those listed in the SSD registration certificate require review and 

approval by the NRC or an Agreement State. Requests to use fixed gauges for purposes not 

listed in the SSD registration certificate will be reviewed on a case-by-case basis. Applicants 

need to submit sufficient information to demonstrate that the proposed use will not compromise 

the source integrity or shielding, or other components of the device critical to radiation safety. 

The NRC will evaluate the radiation safety program for each type and use of gauge requested.

An NRC license does not relieve a licensee from complying with other applicable Federal, 

State, or local regulations.

Response from Applicant: Specifically describe how each device will be used. If the fixed 

gauge(s) will be used for the purposes listed on the SSD registration certificate, or as 

recommended by the manufacturer, the applicant may so state and provide a specific 

description of use for each type of gauge requested. If the fixed gauge will be used for 

purposes other than those listed on the SSD registration certificate, specify these other 

purposes and include a safety analysis supporting the request.
26

Notes:

• Allowed uses of fixed gauges normally include process control methods, such as 

measuring the thickness of paper, the density of coal, the level of material in vessels and 

tanks, etc.

• Unusual uses will be evaluated on a case-by-case basis, and the authorized use 

condition will reflect approved uses.

8.7 Item 7: Individual(s) Responsible for Radiation Safety Program and Their

Training and Experience

8.7.1 Radiation Safety Officer

Regulation: 10 CFR 30.33(a)(3)

Criteria: RSOs must have adequate training and experience. In the past, the NRC has found 

successful completion of one of the following to be evidence of adequate training and 

experience:

• fixed gauge manufacturer’s or distributor’s course for users or for RSOs

OR

• equivalent course that meets the criteria in Appendix D

Additional training is required for RSOs in programs that perform nonroutine operations. This 

includes repairs involving or potentially affecting components related to the radiological safety of

the gauge (e.g., the source, source holder, source drive mechanism, shutter, shutter control, or 

shielding) and any other activities during which personnel could receive radiation doses 

exceeding the NRC’s regulatory dose limits (e.g., installation, initial radiation survey, gauge 

relocation, and removal of the gauge from service). See Section 8.10.8, “Maintenance,” and 

Appendix J to this report for more information on nonroutine operations.

Discussion: The person responsible for the radiation protection program is the RSO. The 

RSO must have adequate training to understand the hazards associated with radioactive 

material and be familiar with all applicable regulatory requirements. The RSO needs 

independent authority to stop operations that he or she considers unsafe. He or she must have 

sufficient time and commitment from management to fulfill certain duties and responsibilities to 

ensure that radioactive materials are used in a safe manner. Typical RSO duties are illustrated 

in Figure 8.1 and described in Appendix C. The NRC requires the name of the RSO to be listed 

on the license to ensure that licensee management always has a responsible, qualified person 

appointed and that the named individual knows of his or her designation as RSO.
27

Figure 8.1 Typical duties and responsibilities of RSOs

Response from Applicant: Provide the following:

• name of the proposed RSO.

• documentation demonstrating that the proposed RSO is qualified by training and 

experience (i.e., certificate of completion of the RSO’s course or the authorized user’s 

course).

Notes: 

• Licensees must inform the NRC of changes in the designation of the RSO and request 

an amendment to the license to name a replacement RSO.

• The RSO may delegate responsibilities to another individual at any site listed on the 

license or at temporary job sites. If these responsibilities are delegated, then the 

delegate RSO should meet the same training and experience criteria as the RSO. 

Please note that only the primary RSO is named on an NRC license.
28

8.7.2 Authorized Users

Regulation: 10 CFR 30.33(a)(3)

Criteria: Authorized users (AUs) must have adequate training and experience in the use of 

fixed gauges. In the past, the NRC has found successful completion of one of the following to 

be evidence of adequate training and experience:

• fixed gauge manufacturer’s or distributor’s course for users

OR

• equivalent course that meets the criteria in Appendix D

Applicants requesting to perform nonroutine operations such as installation, initial radiation 

survey, repair, and maintenance of components related to the radiological safety of the gauge, 

gauge relocation, replacement and disposal of sealed sources, alignment, or removal of a 

gauge from service, must provide additional training. See Section 8.10.8, “Maintenance,” and 

Appendix J to this report for more information.

Discussion: AUs have the responsibility to ensure the proper use, security, and routine 

maintenance of fixed gauges containing licensed material. AUs must attend the training and 

instruction given at the time of installation or receive equivalent training and instruction.

An AU is considered to be supervising the use of licensed material when he or she directs 

personnel in operations involving the material. Although the AU may delegate specific tasks to 

supervised users (e.g., maintaining records), he or she is still responsible for safe use of 

licensed material.

Response from Applicant: Provide either of the following:

• the statement “Before using licensed materials, authorized users will have successfully 

completed one of the training courses described in the ‘Criteria’ part of the section 

entitled ‘Authorized Users’ in the current version of NUREG-1556, Volume 4, 

‘Consolidated Guidance about Materials Licenses: Program-Specific Guidance about 

Fixed Gauge Licenses.’”

OR

• a description of the training and experience for proposed AUs

Note: Records of the training of each AU must be maintained for at least 3 years after the last 

use of licensed material by the AU.
29

8.8 Item 8: Training for Individuals who in the Course of Employment Are

Likely To Receive Occupational Doses of Radiation in Excess of

1 Millisievert (100 Millirem) in a Year (Occupationally Exposed Workers) 

and Ancillary Personnel

Regulations: 10 CFR 19.11, 10 CFR 19.12, 10 CFR 30.7, 10 CFR 30.9, 10 CFR 30.10, and 

10 CFR 30.33

Criteria: Individuals who in the course of employment are likely to receive occupational doses 

of radiation in excess of 1 millisievert (mSv) (100 millirem (mrem)) in a year must receive 

training according to 10 CFR 19.12, “Instruction to workers.” The extent of this training must be 

commensurate with potential radiological health protection problems present in the work place.

Discussion: Licensees need to perform a prospective evaluation to determine radiation doses 

likely to be received by different individuals or groups (see Section 8.10.4, “Occupational 

Dose”). AUs and individuals performing installations, relocations, nonroutine maintenance, or 

repairs would be most likely to receive doses in excess of 1 mSv (100 mrem) in a year. See the 

previous section for a discussion of training and experience for AUs.

Licensee personnel who work in the vicinity of a fixed gauge but do not use gauges (ancillary 

staff) are not required to have radiation safety training as long as they are not likely to receive a 

dose of 1 mSv (100 mrem) in a year. However, to minimize potential radiation exposure when 

ancillary staff are working in the vicinity of a fixed gauge, it is prudent for them to work under the 

supervision and in the physical presence of an AU or to be provided some basic radiation safety 

training. Such ancillary staff should be informed of the nature and location of the gauge and the 

meaning of the radiation symbol and should be instructed not to touch the gauge and to keep 

away from it as much as their work permits.

Some ancillary staff, although not likely to receive doses over 1 mSv (100 mrem), should 

receive training to ensure adequate security and control of licensed material. To ensure the 

control and security of licensed material, licensees may provide these individuals with training 

commensurate with their assignments in the vicinity of the gauge.

Response from Applicant: No response is required from the applicant as part of the license 

application. The applicant’s training program for individuals who in the course of employment 

are likely to receive occupational doses of radiation in excess of 1 mSv (100 mrem) in a year 

(occupationally exposed workers) and ancillary personnel will be examined during inspections.

8.9 Item 9: Facilities and Equipment

Regulations: 10 CFR 20.1101(b), 10 CFR 20.1201, 10 CFR 20.1301, 10 CFR 20.1801, and 

10 CFR 30.33(a)(2)

Criteria: Facilities and equipment must be adequate to protect health and to minimize danger 

to life or property. This may be demonstrated by the following:

• the location of the gauge is compatible with the “Conditions of Normal Use” and 

“Limitations and/or Other Considerations of Use” on the SSD registration certificate,

AND
30

• the fixed gauge is secured to prevent unauthorized removal or access (e.g., located in a 

locked room, permanently mounted, or chained and locked to a storage rack).

Discussion: Fixed gauges incorporate many engineering features to protect the user from 

unnecessary radiation exposure in a wide variety of environments. Fixed gauges may be 

located in harsh environments involving variables such as pressure; vibration; mounting 

height/method; temperature; humidity; air quality; corrosive atmospheres; corrosive chemicals 

including process materials and cleaning agents; possible impact or puncture conditions; and 

fire, explosion, and flooding potentials. Applicants and/or licensees need to consult the sections 

on the SSD registration certificate entitled “Conditions of Normal Use” and “Limitations and/or 

Other Considerations of Use” to determine the appropriate gauge and location. In those 

instances when a proposed location is not consistent with the SSD registration certification, the 

applicant may ask the source or device manufacturer or distributor to request an amendment to 

modify the SSD registration certificate to include the new conditions. If the manufacturer or 

distributor does not request an amendment, the applicant must provide the NRC with specific 

information demonstrating that the proposed new conditions will not impact the safety or 

integrity of the source or device.

Response from Applicant: Provide one of the following:

• A statement that “We will ensure that the location of each fixed gauge meets the criteria 

in Section 8.9, ‘Facilities and Equipment,’ in the current version of NUREG-1556, 

Volume 4, ‘Consolidated Guidance about Materials Licenses: Program-Specific 

Guidance about Fixed Gauge Licenses.’”

OR

• Confirmation that the fixed gauge is secured to prevent unauthorized removal or access 

and submittal of specific information demonstrating that the proposed conditions will not 

impact the safety or integrity of the source or device. Address any instances where the 

proposed conditions exceed any conditions listed in the SSD registration certificate.

Note: Any deviations from an SSD registration certificate will require specific NRC or 

Agreement State approval.

8.10 Item 10: Radiation Safety Program

8.10.1 Audit Program

Regulations: 10 CFR 20.1101 and 10 CFR 20.2102

Criteria: Licensees must review the content and implementation of their radiation protection 

programs at least annually to ensure the following:

• Programs comply with NRC and DOT regulations and the terms and conditions of the 

license. 
31

• Occupational doses and doses to members of the public are ALARA (per 

10 CFR 20.1101, “Radiation protection programs”).

• Records of audits and other reviews of program content are maintained for at least 

3 years.

Discussion: Appendix E contains a suggested audit program that is specific to the use of fixed 

gauges and is acceptable to the NRC. All areas indicated in Appendix E may not be applicable 

to every licensee and may not need to be addressed during each audit.

Currently, the NRC’s emphasis in inspections is to perform actual observations of work in 

progress. As a part of their audit programs, applicants should consider performing 

unannounced audits of fixed gauge users to determine if, for example, operating and 

emergency procedures are available and are being followed, etc.

It is essential that once problems are identified comprehensive corrective action are taken in a 

timely manner; Information Notice (IN) 96-28, “Suggested Guidance Relating to Development 

and Implementation of Corrective Action,” dated May 1, 1996, provides guidance on this 

subject. The NRC will review the licensee’s audit results and determine if corrective actions are 

thorough, timely, and sufficient to prevent recurrence. If violations are identified by the licensee 

and these steps are taken, the NRC will normally exercise discretion and may elect not to cite a 

violation. The NRC’s goal is to encourage prompt identification and prompt, comprehensive 

correction of violations and deficiencies. For additional information on the NRC’s use of 

discretion in issuing violations, refer to the current version of the NRC Enforcement Policy.

With regard to audit records, 10 CFR 20.2102(a) requires, in part, that licensees maintain 

records of “audits and other reviews of program content and implementation” for 3 years from 

the date of the record. The NRC has found audit records that contain the following information 

to be acceptable: date of audit, name of person(s) who conducted audit, persons contacted by 

the auditor(s), areas audited, audit findings, corrective actions, and followup.

Response from Applicant: The applicant is not required to submit its audit program to the 

NRC for review during the licensing phase. The applicant’s audit program will be reviewed 

during NRC inspections.

References: For information about NRC enforcement, see the current version of the NRC 

Enforcement Policy found on the NRC’s public website. IN 96-28 is available electronically in 

the “NRC Library” on the NRC’s Home Page at http://www.nrc.gov.

8.10.2 Instruments

Regulations: 10 CFR 20.1501, 10 CFR 20.2103(a), and 10 CFR 30.33(a)(2)

Criteria: Licensees should possess, or have access to, radiation monitoring instruments, which 

are necessary to protect health and minimize danger to life or property. Instruments used for 

quantitative radiation measurements must be calibrated periodically for the radiation measured.

Discussion: Typically, fixed gauge licensees are not required to possess a survey meter; 

however, an applicant that plans to conduct nonroutine operations will need to conduct surveys 

in accordance with 10 CFR 20.1501, “General.” Nonroutine operations include installation, 

initial radiation surveys, relocation, removal from service, dismantling, alignment, replacement, 
32

disposal of the sealed source, and nonroutine maintenance and repair of components related to 

the radiological safety of the gauge. See Section 8.10.8, “Maintenance,” and Appendix J for 

more information on maintenance and nonroutine operations. Because some of these 

operations may increase the opportunity for radiation exposure, the areas where individuals will 

be performing these operations should be carefully monitored with a survey meter. Such survey 

meters must be properly calibrated. Proper calibration is particularly important for initial 

surveys, because the results can be used as a basis for public dose estimates. For those 

licensees requesting authorization to calibrate their own survey instruments, Appendix F

contains calibration procedures acceptable to the NRC.

Each year, a number of gauges experience equipment failures because of corrosion caused by 

the harsh environmental conditions in which many gauges are operated. Following discovery of 

an equipment failure, it is important to determine if the shielding and source are intact. The 

most efficient and effective way to do this is by measuring the radiation levels in the vicinity of 

the gauge with an appropriate, calibrated survey instrument. Because many fixed gauge 

licensees are not required to possess a survey meter, applicants should preplan how they will 

obtain assistance in performing a radiation survey in the event of an emergency (e.g., obtain a 

survey instrument from the manufacturer, a consultant, another NRC or Agreement State 

licensee, or a local emergency response organization).

Response from Applicant: Provide one of the following:

• A statement that “Surveys according to 10 CFR 20.1501 will be performed by a person 

specifically authorized by the NRC or an Agreement State to perform these surveys.”

OR

• A statement that “We will use survey instruments that meet the criteria in Section 8.10.2, 

‘Instruments,’ in the current version of NUREG-1556, Volume 4, ‘Consolidated Guidance 

about Materials Licenses: Program-Specific Guidance about Fixed Gauge Licenses,’

and one of the following three choices:

– “Each survey meter will be calibrated by the manufacturer or other person 

authorized by the NRC or an Agreement State to perform survey meter

calibrations.”

OR

– “We will implement the model survey meter calibration program published in 

Appendix F, ‘Model Survey Instrument Calibration Program’ in the current version 

of NUREG-1556, Volume 4, ‘Consolidated Guidance about Materials Licenses: 

Program-Specific Guidance about Fixed Gauge Licenses.’”

OR

– “We will submit alternative calibration procedures for NRC review.”

OR, IN LIEU OF ALL OF THE ABOVE

• Submit a description of an alternative method to perform surveys according to 

10 CFR 20.1501.
33

Notes:

• Alternative responses will be reviewed against the criteria listed above.

• Regardless of whether an applicant is authorized to calibrate survey meters or contracts 

an authorized firm to perform calibrations, the licensee must retain calibration records for 

at least 3 years in accordance with 10 CFR 20.2103(a).

8.10.3 Material Receipt and Accountability

Regulations: 10 CFR 20.1801, 10 CFR 20.1802, 10 CFR 20.2201, 10 CFR 20.2207, 

10 CFR 30.34(e), 10 CFR 30.41, 10 CFR 30.51

Criteria: Licensees must do the following:

• Maintain records of receipt, transfer, and disposal of fixed gauges.

• Conduct physical inventories at intervals not to exceed 6 months (or some other interval 

justified by the applicant and approved by the NRC) to account for all sealed sources.

• Update transactions in the NSTS and perform an NSTS annual inventory reconciliation, 

if applicable.

Discussion: Licensed materials must be tracked from “cradle to grave” in order to ensure 

gauge accountability; identify when gauges could be lost, stolen, or misplaced; and ensure that 

possession limits listed on the license are not exceeded. Significant problems can arise from 

failure to ensure the accountability of gauges (see NRC IN 88-02, “Lost or Stolen Gauges,”

dated February 2, 1988).

Receipt, inventory, transfer, and disposal records must be maintained for the times specified in 

Table 8.2. Typically, these records contain the following types of information:

• the radionuclide and the activity (in units of becquerels or curies) of byproduct material in 

each sealed source

• the manufacturer’s or distributor’s name, the model number, and the serial number (if 

appropriate) of each device containing byproduct material

• the location of each sealed source and device

• for inventories, the date of the inventory and signature of the individual conducting the 

inventory

• for materials transferred or disposed of, the date of the transfer or disposal, the name 

and license number of the recipient, and a description of the affected radioactive 

material (e.g., radionuclide, activity, manufacturer’s or distributor’s name and model 

number, serial number)
34

Table 8.2 Record Maintenance

Type of Record How Long Record Must Be Maintained

Receipt For as long as the material is possessed until 3 years after 

transfer or disposal

Inventory For 5 years from the date of the inventory

Transfer For 3 years after transfer

Disposal Until the NRC terminates the license

Important to Decommissioning* Until the site is released for unrestricted use

* See Section 8.5.2, “Financial Assurance and Recordkeeping for Decommissioning,” for more 

details.

Response from Applicant: Provide the following:

• A statement that “Physical inventories will be conducted at least every 6 months or at

other intervals approved by the NRC, to account for all sealed sources and devices

received and possessed under the license.”

• A statement that “We will develop, implement, and maintain procedures for ensuring 

accountability of licensed materials at all times.”

Note: The requirements of the NSTS apply to sealed sources containing a quantity equal to or 

greater than Category 1 or Category 2 levels of any radioactive material listed in Appendix E, 

“Nationally Tracked Source Thresholds,” to 10 CFR Part 20.

References: INs are available in the “NRC Library” on the NRC’s public website at

http://www.nrc.gov. 

8.10.4 Occupational Dose

Regulations: 10 CFR 19.13, 10 CFR 20.1201, 10 CFR 20.1207, 10 CFR 20.1208, 

10 CFR 20.1501, and 10 CFR 20.1502

Criteria: Applicants must do either of the following:

• Perform a prospective evaluation demonstrating that unmonitored individuals are not 

likely to receive, in one year, a radiation dose in excess of 10 percent of the allowable 

limits, as shown in Figure 8.2, and maintain the record for inspection by the NRC.

OR

• Provide dosimetry processed and evaluated by a National Voluntary Laboratory 

Accreditation Program (NVLAP) approved processor that is exchanged at a frequency 

recommended by the processor.
35

Figure 8.2 Annual dose limits for radiation workers

Discussion: Under conditions of routine use, the typical fixed gauge user does not require a 

personnel monitoring device (dosimetry). A gauge user also may not require dosimetry when 

proper emergency procedures are used. Appendix G provides guidance on performing a 

prospective evaluation demonstrating that fixed gauge users are not likely to exceed 10 percent 

of the limits shown in Figure 8.2 and thus are not required to have personnel dosimetry.

Individuals who perform nonroutine operations such as installation, initial radiation survey, 

repair, and maintenance of components related to the radiological safety of the gauge, gauge 

relocation, replacement, and disposal of sealed sources, alignment, or removal of a gauge from 

service are more likely to exceed 10 percent of the limits shown in Figure 8.2. Applicants must 

provide dosimetry (whole body and perhaps extremity monitors) to individuals who perform

nonroutine operations or must perform a prospective evaluation demonstrating that unmonitored 

individuals who perform nonroutine operations are not likely to receive, in one year, a radiation 

dose in excess of 10 percent of the allowable limits as shown in Figure 8.2.

When personnel monitoring is needed, most licensees use dosimeters supplied by an

NVLAP-approved processor. The exchange frequency for dosimeters is typically monthly or 

quarterly. Applicants should consult with their NVLAP-approved processor for its 

recommendations for exchange frequency and proper use of the dosimeter.

Response from Applicant: Provide any one of the following:

• A statement that “We will maintain, for inspection by the NRC, documentation 

demonstrating that unmonitored individuals are not likely to receive, in one year, a 

radiation dose in excess of 10 percent of the allowable limits in 10 CFR Part 20.”

OR

Dose Limits for

Radiation Worker

(10 CFR 20.1201)

Total effective dose equivalent 

(whole body) 0.05 Sv (5 rem)

Eyes (Lens)

0.15 Sv

(15 rem)

Skin

0.50 Sv

(50 rem)

Elbow

to hands

0.50 Sv

(50 rem)

Knees to feet

0.50 Sv

(50 rem)
36

• A statement that “We will provide dosimetry, processed and evaluated by an 

NVLAP-approved processor, that is exchanged at a frequency recommended by the 

processor.”

OR

• A description of an alternative method for demonstrating compliance with the referenced 

regulations.

Notes:

• Alternative responses will be evaluated against the criteria listed above.

• Some licensees choose to provide personnel dosimetry to their workers for reasons 

other than compliance with NRC requirements (e.g., to respond to worker requests or to 

maintain records of personal exposure).

Reference: The National Institute of Standards and Technology (NIST) maintains a directory of 

accredited laboratories at http://ts.nist.gov.

8.10.5 Public Dose

Regulations: 10 CFR 20.1003, 10 CFR 20.1301, 10 CFR 20.1302, 10 CFR 20.1801, 

10 CFR 20.1802, 10 CFR 20.2107

Criteria: Licensees must do the following:

• Ensure that fixed gauges will be used, transported, and stored in such a way that 

members of the public will not receive more than 1 mSv (100 mrem) in one year, and the 

dose in any unrestricted area will not exceed 0.02 mSv (2 mrem) in any one hour, from 

licensed operations.

AND

• Prevent unauthorized access, removal, or use of fixed gauges.

Discussion: Public dose is defined in 10 CFR Part 20 as “the dose received by a member of 

the public from exposure to radiation or to radioactive material released by a licensee, or to any 

other source of radiation under the control of a licensee.” Public dose excludes doses received 

from background radiation and medical procedures. Whether the dose to an individual is an 

occupational dose or a public dose depends on the individual’s assigned duties. It does not 

depend on the area (restricted, controlled, or unrestricted) where the individual is when he or 

she receives the dose.

In the case of fixed gauges, members of the public include persons who live, work, or may be 

near locations where fixed gauges are used or stored and employees whose assigned duties do 

not include the use of licensed materials and who work in the vicinity where gauges are used or 

stored. Since a fixed gauge presents a radiation field, the applicant must use methods to limit 

the public dose such that the radiation level in an unrestricted area (e.g., a nearby walkway or 
37

area near the gauge that requires frequent maintenance) does not exceed 1 mSv (100 mrem)

in a year or 0.02 mSv (2 mrem) in any one hour.

Because fixed gauges are generally permanently mounted (e.g., chained and locked to a 

storage rack), they may not need to be in a locked area to prevent loss, theft, or unauthorized 

removal. Procedures regarding security and lock-out procedures specified in Section 8.10.6, 

“Operating and Emergency Procedures,” of this document should be sufficient to limit the 

exposure to the public during use or storage and after accidents. IN 81-37, Revision 1,

“Unnecessary Radiation Exposures to the Public and Workers During Events Involving 

Thickness and Level Measuring Devices,” dated February 9, 1982, provides information about 

an event that resulted or may have resulted in unnecessary radiation exposure to members of 

the public. IN 88-02 provides information about several events where fixed gauges were lost or 

stolen.

Public dose is also affected by the location of the gauge. Use the concepts of time, distance, 

and shielding when developing a method to limit public dose. Decreasing the time spent near a 

gauge, increasing the distance from the gauge, and using shielding will reduce the radiation 

exposure. The most effective way to limit public dose is to prevent members of the public from 

entering areas where gauges are used or stored. This may be accomplished by administrative

or engineering controls.

Administrative controls include training and warning signs. In cases where gauges are located 

in harsh environmental conditions (e.g., high temperatures, caustic chemicals, etc.), warning 

signs may be difficult to maintain so requiring radiation awareness training to caution employees

would be a good practice.

Engineering controls reduce radiation levels in areas that are accessible to the public. Shielding 

the gauge with a protective barrier (e.g., using brick, concrete, lead, or other solid walls) or 

placing the gauge within an enclosure to prevent access to higher radiation levels are examples 

of engineering controls (see Figure 8.3).
38

Figure 8.3 Limiting Public Dose

When dose rates in an area are high enough that a member of the public could receive a dose 

in excess of 0.02 mSv (2 mrem) in any one hour or 1 mSv (100 mrem) in a year, licensees must 

take additional measures to prevent public access to these higher dose rates, such as building 

enclosures around the gauges.

Public dose can be estimated in areas near the gauge by using radiation levels determined 

during initial surveys and applying the “inverse square” law to evaluate the effect of distance on 

radiation levels and occupancy factors to account for the actual presence of members of the 

public (see Appendix G for an example).

If, after making a public dose estimate, the conditions used to make the evaluation change (e.g., 

there are changes to the location of gauges, changes to the type or frequency of gauge use, 

addition of new gauges, or changes to the occupancy of adjacent areas), then the licensee must 

perform a new evaluation to ensure that the public dose limits are not exceeded and take 

corrective action, as needed.

During NRC inspections, licensees must be able to provide documentation demonstrating, by 

measurement or calculation, that the dose to the individual likely to receive the highest dose 

from the licensed operation does not exceed the annual limit for individual members of the 

public (see Appendix G for examples of methods to demonstrate compliance).

Response from Applicant: No response is required from the applicant during the licensing 

phase. The NRC will examine this matter during inspection.

References: INs are available in the “NRC Library” on the NRC’s Home Page at 

http://www.nrc.gov.
39

8.10.6 Operating and Emergency Procedures

Regulations: 10 CFR 19.11(a)(3), 10 CFR 20.1101, 10 CFR 20.1801, 10 CFR 20.1802, 

10 CFR 20.2201–2203, 10 CFR 20.2207, 10 CFR 21.21, 10 CFR 30.34(e), 10 CFR 30.50

Criteria: Each applicant should do the following:

• Develop, implement, and maintain operating procedures that contain the following 

elements for each type of fixed gauge:

– instructions for using the gauge, including use at temporary job sites if requested 

and authorized on the license (see Section 8.10.10, “Fixed Gauges Used at 

Temporary Job Sites,” for more information on fixed gauges used at temporary 

job sites)

– instructions for performing routine maintenance according to the manufacturer’s 

recommendations and instructions

– instructions for testing each gauge for the proper operation of the on-off 

mechanism (shutter) and indicator, if any, at intervals not to exceed 6 months or 

as specified in the SSD registration certificate

– instructions for lock-out procedures, if applicable, that are adequate to ensure 

that no individual or portion of an individual’s body can enter the radiation beam

– steps to take to keep radiation exposures ALARA

– steps to maintain accountability of gauges (i.e., inventory)

– instructions to ensure that nonroutine operations such as installation, initial 

radiation survey, repair and maintenance of components related to the 

radiological safety of the gauge, gauge relocation, replacement and disposal of 

sealed sources, alignment, or removal of a gauge from service are performed by 

the manufacturer, distributor, or person specifically authorized by the NRC or an 

Agreement State

– steps to ensure that radiation warning signs are visible and legible

– instructions for transporting radioactive material to ensure compliance with DOT 

regulations, if the applicant will be transporting gauges to temporary job sites.

• Develop, implement, and maintain emergency procedures for gauge malfunction or 

damage that contain the following elements for each type of fixed gauge:

– instructions for how to stop the use of the gauge

– instructions for restricting access to the area

– contact information for safety/emergency personnel (telephone numbers for the 

RSO, AUs, the gauge manufacturer or distributor, fire department, or other 

emergency response organization)
40

– guidelines for reporting to the NRC pursuant to 10 CFR 20.2202, “Notification of 

incidents,” 10 CFR 20.2203, “Reports of exposures, radiation levels, and 

concentrations of radioactive material exceeding the constraints or limits,” 10 

CFR 21.21, “Notification of failure to comply or existence of a defect and its 

evaluation,” and/or 10 CFR 30.50, “Reporting requirements”

– descriptions of additional actions or precautions based on the specific situation

– guidelines to follow when a gauge is involved in a transportation accident, if the 

license authorizes temporary job sites.

• Develop, implement, and maintain security procedures to prevent theft, loss, or sabotage 

that contain the following elements:

– instructions to prevent unauthorized access, removal, or use of fixed gauges, 

including at temporary job sites if authorized on the license

– requirements for reporting to the NRC pursuant to 10 CFR 20.2201, “Reports of 

theft or loss of licensed material,” and 10 CFR 20.2207, “Reports of transactions 

involving nationally tracked sources,” as appropriate

– guidelines to meet the security requirements of 10 CFR 20.1801, “Security of 

stored material,” 10 CFR 20.1802, “Control of material not in storage,” and any 

other applicable security requirements

• Provide copies of operating, emergency, and security procedures to all gauge users and 

at each site.

• Post copies of operating and emergency procedures at each location of use, or if posting 

procedures is not practicable, post a notice that briefly describes the procedures and 

states where they may be examined.

Discussion: Applicants must develop, implement, and maintain operating and emergency 

procedures. The NRC will permit an applicant greater flexibility when licensing certain types of 

gauges. For each gauge that is requested, if one or more of the following safety conditions are 

met, the applicant need not submit these procedures for NRC review during the licensing 

process:

• The air gap between the radiation source and detector of the device is less than 45 

centimeters (18 inches).

• The air gap of the device would not allow insertion of a 30-centimeters (12-inch)

diameter sphere into the radiation beam of the device without removal of a barrier.

• The radiation dose rate in the radiation beam of the device at 45 centimeters (18 inches)

from the radiation source with the device shutters, if any, in the open position does not 

exceed 1 mSv per hour (0.1 rem per hour).
41

• Entry into vessels (e.g., bins, tanks, hoppers, or pipes) with a gauge installed is not 

necessary under any foreseeable circumstances and is prohibited.

If the requested gauge does not meet any of the safety conditions above, then the applicant 

must submit its procedures to the NRC for review during the licensing process.

Operating, emergency, and security procedures should be developed, maintained, and 

implemented to ensure that gauges are used only as they were designed to be used, control 

and accountability are maintained, and radiation doses received by occupational workers and 

members of the public are ALARA.

Improper operation could lead to the damage or malfunction of a gauge and elevated exposure 

rates in the gauge’s immediate vicinity. A list of specific items that should be addressed in 

operating, emergency, and security procedures is contained in Appendix H. Figure 8.4

illustrates a proper response to a fire involving a fixed gauge. Emergency procedures should be 

developed to address a spectrum of incidents (e.g., fire, explosion, mechanical damage, flood, 

or earthquake).
42

Figure 8.4 Proper handling of incident

Licensee personnel implement emergency procedures when a fire melts the lead shielding of a 

gauge producing the potential for elevated exposure levels.
43

The NRC considers the security of licensed material to be extremely important, and lack of 

security is a significant violation for which licensees may be fined. Although most fixed gauges 

are difficult to move, the licensee must prevent unauthorized access, removal, or use of the 

gauge. Licensees are responsible for ensuring that gauges are secured and accounted for at all 

times (e.g., during plant modifications, change in ownership, staffing changes, or after 

termination of activities at a particular location).

The NRC must be notified when gauges are lost or stolen or certain other conditions occur.

The RSO should be proactive in evaluating whether NRC notification is required. Refer to 

Appendix L to this report and the regulations (i.e., 10 CFR 20.2201–20.2203, 10 CFR 20.2207, 

10 CFR 21.21, and 10 CFR 30.50) for a description of when and where notifications are 

required.

Figure 8.5 Lock-out procedures

Typical lock-out procedures include locking the shutter into the “off” position and tagging the 

shutter control mechanism to indicate the gauge is locked out.

When the distance or air gap between the source and detector permits entry of all or a portion of 

a person’s body into the primary radiation beam, licensees must develop lock-out procedures. 

Lock-out procedures encompass locking the on-off or shutter mechanism into the off position or 

otherwise controlling the radiation beam or using any other means of preventing an individual or 

a portion of an individual’s body from entering the radiation beam during maintenance, repairs, 

or work in, on, or around the process line (e.g., bin, tank, hopper, pipe, or conveyor belt) where 

the device is mounted. The on-off or shutter control mechanism should be tagged to indicate 

that the gauge is locked out. A warning sign should be posted at each entryway to an area 

where it is possible to be exposed to the radiation beam. In addition to providing a warning, the 

sign should give safety instructions (e.g., “contact the RSO before entering this vessel”). Lockout procedures should specify who is responsible for performing them.
44

Response from Applicant: Provide one of the following:

• If the gauge meets one or more of the safety conditions specified in “Discussion,” 

provide the following

 A statement that “Operating, emergency, and security procedures will be developed, 

implemented, maintained, and distributed, and will meet the criteria in Section 8.10.6, 

’Operating, and Emergency Procedures’ in the current version of NUREG-1556, 

Volume 4, ’Consolidated Guidance about Materials Licenses: Program-Specific 

Guidance about Fixed Gauge Licenses.’”

OR

• If the gauge does not meet any of the safety conditions specified in “Discussion,” provide 

your operating, emergency, security, and lock-out (if applicable) procedures to the NRC 

for review during the licensing process.

Note: When developing operating, emergency, and security procedures, applicants should 

consider the information contained in the following documents describing incidents involving 

fixed gauges:

• IN 81-37, Revision 1, “Unnecessary Radiation Exposures to the Public and Workers 

During Events Involving Thickness and Level Measuring Devices,” dated 

February 9, 1982

• IN 86-31, “Unauthorized Transfer and Loss of Control of Industrial Nuclear Gauges,” 

dated May 5, 1986

• IN 88-02, “Lost or Stolen Gauges,” dated February 2, 1988

• IN 88-90, “Unauthorized Removal of Industrial Nuclear Gauges,” dated 

November 22, 1988

• IN 94-15, “Radiation Exposures during an Event Involving a Fixed Nuclear Gauge,” 

dated March 2, 1994 

References: INs are available in the “NRC Library” on the NRC’s Home Page at 

http://www.nrc.gov.

8.10.7 Leak Tests

Regulations: 10 CFR 20.1501, 10 CFR 20.2103, and 10 CFR 30.53

Criteria: The NRC requires testing to determine whether there is any radioactive leakage from 

the source in the fixed gauge. Records of the test results must be maintained for 3 years.

Discussion: Based on the conditions of the license, a licensee is required to perform leak tests 

at an interval approved by the NRC or an Agreement State in the device’s SSD registration 

certificate or at a more frequent interval that the licensee committed to in its license application.
45

The measurement of the leak test sample is a quantitative analysis that requires instrumentation 

capable of detecting 185 becquerels (0.005 microcuries) of radioactivity.

Manufacturers, distributors, consultants, and other organizations may be authorized by the NRC 

or an Agreement State to either perform the entire leak test sequence on behalf of licensees or 

provide leak test kits to licensees. In the latter case, the licensee takes the leak test sample 

according to the manufacturer’s and/or the kit supplier’s instructions and returns it to the leak 

test service provider for evaluation and reporting results. Leak test samples should be collected 

at the most accessible area where contamination would accumulate if the sealed source were 

leaking. Licensees may also be authorized to conduct the entire leak test sequence 

themselves. Appendix I contains information to support a request to perform leak testing and 

sample analysis.

Response from Applicant: Provide one of the following:

• A statement that “Leak tests will be performed at intervals approved by the NRC or an 

Agreement State and specified in the Sealed Source and Device registration certificate. 

Leak tests will be performed by an organization authorized by the NRC or an Agreement 

State to provide leak testing services to other licensees or using a leak test kit supplied 

by an organization authorized by the NRC or an Agreement State to provide leak test 

kits to other licensees and according to the kit supplier’s instructions. Records of leak

test results will be maintained.”

OR

• A statement that “We will implement the model leak test program published in Appendix I

to the current version of NUREG-1556, Volume 4, ‘Consolidated Guidance about 

Materials Licenses: Program-Specific Guidance about Fixed Gauge Licenses.’”

OR

• A description of alternative equipment and/or procedures for determining whether there 

is any radioactive leakage from sources contained in gauges.

Note: Requests for authorization to perform leak testing and sample analysis will be reviewed 

and, if approved, will be authorized via a license condition.

8.10.8 Maintenance

Regulations: 10 CFR 20.1101, 10 CFR 30.33(a)(3), and 10 CFR 30.34(e)

Criteria: Licensees should routinely clean and maintain gauges according to the 

manufacturer’s or distributor’s written recommendations and instructions. Individuals performing 

routine maintenance must have adequate training and experience. Radiation safety procedures 

for routine cleaning and maintenance (e.g., removal of exterior residues from the gauge 

housing, shutter operation checks, external lubrication of shutter mechanism, calibration, and 

electronic repairs) must consider ALARA principles and ensure that the gauge functions as 

designed and source integrity is not compromised.
46

“Nonroutine maintenance or repair” (beyond routine cleaning, lubrication, calibration, and 

electronic repairs) means any maintenance or repair that involves or potentially affects 

components, including electronics, related to the radiological safety of the gauge (e.g., the 

source, source holder, source drive mechanism, shutter, shutter control or shielding) and any 

other activities during which personnel could receive radiation doses exceeding the NRC’s 

regulatory dose limits.

The fixed gauge manufacturer or distributor, or a person specifically authorized by the NRC or 

an Agreement State, must perform nonroutine maintenance or repair. Appendix J presents 

information to support requests for specific authorization to perform nonroutine maintenance or 

repair.

Figure 8.6 Maintenance

Licensees need to perform routine maintenance to ensure proper operation of the fixed gauge. 

For nonroutine maintenance, most licensees use or contract the services of the gauge 

manufacturer, distributor, or a service company.

Discussion: The NRC allows fixed gauge licensees to perform routine maintenance of the 

gauges. Licensees should follow the gauge manufacturer’s or distributor’s written 

recommendations and instructions provided with the gauge(s). Generally, before any 

maintenance or repair work is done, licensees need to determine (and assure themselves of the 

adequacy of) the following:

• the tasks to be performed;

• the protocol or procedures to be followed;

• the radiation safety procedures including possible need for compensatory measures 

(i.e., steps taken to compensate for lack of or reduced shielding);

• ALARA considerations;

• training and experience of personnel performing the work;
47

• instructions for lock-out procedures;

• the qualification of parts, components, other materials to be used in the gauge; and

• the tests to be performed before the gauge is returned to routine use to ensure that it 

functions as designed.

Although manufacturers or distributors may use different terms, “routine maintenance” includes, 

but is not limited to, cleaning, lubricating, and calibrating the gauge and making electronic 

repairs. Routine maintenance also includes testing the proper operation of the gauge’s on-off 

mechanism (i.e., shutter checks) and indicator at intervals not to exceed 6 months or as 

specified in the SSD registration certificate. A standard license condition in all fixed gauge 

licenses imposes this requirement. Two NRC Information Notices: (1) IN 2009-18, 

“Performance of Required Shutter Checks and Reporting of Gauge Shutter Failures,” dated 

September 18, 2009, and (2) IN 2011-09, “Fixed Gauge Shutter Failures Due to Operating in 

Harsh Working Environments,” dated May 18, 2011, discuss the importance of performing 

shutter checks.

Routine maintenance does not include any activities that involve:

• components, including electronics, related to the radiological safety of the gauge (e.g., the 

source, source holder, source drive mechanism, shutter, shutter control or shielding);

• installation, relocation, or alignment of the gauge;

• initial radiation surveys;

• replacement and disposal of sealed sources;

• removal of a gauge from service;

• a potential for any portion of the body to come into contact with the primary radiation beam; 

and

• any other activity during which personnel could receive radiation doses exceeding the 

NRC’s regulatory dose limits.

Mounting a gauge is unpacking or uncrating the gauge and fastening, hanging, or affixing the 

gauge into position before using. Mounting does not include electrical connection, activation, or 

operation of the gauge. Installing a gauge includes mounting, electrical connection, activation, 

and first use of the device. Specific NRC or Agreement State authorization on a license is 

required to install a gauge; however, a licensee may initially mount a gauge, without specific 

NRC or Agreement State authorization, if the gauge’s SSD registration certificate explicitly 

permits it. The conditions under which a licensee may initially mount a gauge will be specified 

on the license as a standard license condition if the licensee is not specifically authorized to 

install a gauge. The source must remain fully shielded and the gauge may not be used until it is 

installed and made operational by a person or entity specifically licensed by the NRC or an 

Agreement State to perform such operations.
48

A condition in the NRC license will state that operations such as installation, initial radiation 

survey, repair, and maintenance of components related to the radiological safety of the gauge, 

gauge relocation, replacement, and disposal of sealed sources, alignment, or removal of a 

gauge from service shall be performed only by the manufacturer, distributor, or other persons 

specifically licensed by the NRC or an Agreement State to perform such services. Most 

licensees do not perform nonroutine operations. Rather, these licensees use or contract the 

services of other licensees that have the specialized equipment and the technical expertise 

needed to perform these activities. Applicants seeking authorization to perform nonroutine 

operations must submit specific procedures for review. See Appendix J for more information.

Response from Applicant:

For routine maintenance: Submit either of the following:

• A statement that “We will implement and maintain procedures for routine maintenance of 

our gauges according to each manufacturer’s or distributor’s written recommendations 

and instructions.”

OR

• Alternative procedures for the NRC’s review.

For nonroutine operations: Submit either of the following:

• A statement that “The gauge manufacturer, distributor, or other person authorized by the 

NRC or an Agreement State will perform nonroutine operations such as installation, 

initial radiation survey, repair and maintenance of radiological safety components, 

relocation, replacement, alignment, removal from service, and disposal of sealed 

sources.”

OR

• Procedures using the information listed in Appendix J to support a request to perform 

nonroutine operations “in-house.”

Notes:

• Alternative procedures for performing routine maintenance will be evaluated using the 

criteria listed above.

• Requests for authorization to perform nonroutine operations will be reviewed and, if 

approved, will be authorized via a license condition.

References: INs are available in the “NRC Library” on the NRC’s Home Page at 

http://www.nrc.gov. 
49

8.10.9 Transportation

Regulations: 10 CFR 20.1101, 10 CFR 71.5, and 49 CFR Parts 171–178

Criteria: Applicants must do one of the following:

• Develop, implement, and maintain safety procedures for off-site transport of radioactive 

material to ensure compliance with DOT regulations.

OR

• Arrange for transportation of a gauge by the manufacturer, distributor, or other person 

specifically licensed to transport gauges by the NRC or Agreement State.

Discussion: Some fixed gauge licensees have the manufacturer, distributor, or other person 

specifically licensed to transport gauges by the NRC or Agreement State arrange for preparing 

and shipping licensed material. If licensees decide to transport their own gauges, they are 

responsible for compliance with DOT regulations including hazardous materials (hazmat)

employee training.

Figure 8.7 illustrates some important DOT requirements for gauge licensees, and Appendix K

lists applicable DOT regulations.

Figure 8.7 DOT transportation requirements

Response from Applicant: No response is needed from applicants during the licensing 

process; this issue will be reviewed during inspection.

Reference: “Radioactive Material Regulations Review,” published by DOT’s Pipeline and 

Hazardous Materials Safety Administration in December 2008, can be obtained at 

http://www.phmsa.dot.gov.
50

8.10.10 Fixed Gauges Used at Temporary Job Sites

Regulations: 10 CFR 20.1101, 10 CFR 20.1801, 10 CFR 20.1802, 10 CFR 20.2201–2203, 

10 CFR 30.34(e), and 10 CFR 30.50.

Criteria: Each applicant requesting authorization to perform work with fixed gauges at 

temporary job sites must develop, implement, and maintain operating, emergency, and security

procedures, as described in Section 8.10.6, ”Operating and Emergency Procedures,” and 

provide copies of those procedures to all gauge users and at each job site.

Discussion: A temporary job site is a location where work with licensed materials is conducted 

for a limited period of time. Temporary job sites are not specifically listed on a license. A gauge 

user may be dispatched to work at several temporary job sites in one day. A location is not 

considered a temporary job site if it is used to store and dispatch licensed material. The NRC 

considers such a location to be a field office. Licensees must apply for and receive a license 

amendment specifically listing each field office location.

There are two primary categories of fixed gauges used at temporary job sites: (1) gauges that 

are permanently mounted to vehicles or trailers and (2) gauges that are transported to industrial 

facilities and temporarily installed on process equipment to conduct short-term quality 

assurance and/or quality control studies (see Figure 8.8). The NRC also handles the licensing 

of sealed sources used for industrial diagnostic scanning at facilities and mobile nonintrusive 

scanning systems in the same manner as fixed gauges. These applications of radioactive 

materials may also be used at temporary job sites.

Figure 8.8 Examples of uses for fixed gauges at temporary job sites
51

When working at a temporary job site, licensees generally have to follow the rules and 

procedures of the organization that owns or controls the site. Thus, licensees may not be able 

to restrict access to areas in the same manner that they can at their own facilities. Furthermore, 

non-licensee personnel may not be familiar with fixed gauges or radioactive material. 

Therefore, to avoid lost or stolen gauges and to prevent unnecessary radiation exposures to 

members of the public, licensees must keep gauges under constant surveillance, or secured 

against unauthorized use or removal (see Figure 8.9).

Figure 8.9 Security

Examples of Methods Used To Secure Fixed Gauges at Temporary Job Sites.

While installing gauges at a temporary job site, personnel could receive radiation doses 

exceeding NRC limits if they do not follow proper radiation safety principles. Licensee 

personnel performing installations should be assigned and wear dosimetry and use a survey 

meter to monitor radiological conditions (see Figure 8.10).
52

Figure 8.10 Installation of fixed gauges at temporary job sites

Examples of the Additional Precautions Needed When Installing Fixed Gauges at Temporary 

Job Sites.

After installing a gauge at a temporary job site, a radiation survey must be conducted to ensure 

that dose rates in unrestricted areas will not exceed 0.02 mSv (2 mrem) in any one hour or 

1 mSv (100 mrem) in a year. If surveys indicate that a member of the public (e.g., client 

personnel) could receive a dose exceeding these limits, licensees would need to adopt 

additional measures to prevent public access, such as maintaining constant surveillance or 

erecting physical barriers.

Response from Applicant: Submit one of the following:

• A statement that “We will not use fixed gauges at temporary job sites.”

OR

• A statement that “We will develop, implement, maintain, and distribute procedures that 

meet the criteria in Section 8.10.10, ‘Fixed Gauges Used at Temporary Job Sites,’ of 

NUREG-1556, Volume 4, ‘Consolidated Guidance about Materials Licenses: ProgramSpecific Guidance about Fixed Gauge Licenses.’ Copies of these procedures will be 

provided to all gauge users and will be available at all temporary job sites.”

8.10.11 Security Program

Regulations: 10 CFR 20.1801, 10 CFR 20.1802, and 10 CFR 20.2207

Criteria: Licensees must ensure the security and control of licensed material.

Discussion: The regulations in 10 CFR 20.1801, and 10 CFR 20.1802 require licensees to 

secure radioactive materials from unauthorized removal or access while in storage and to 

control and maintain constant surveillance over licensed material that is not in storage. 

Depending on the quantity and type of radioactive materials used by fixed gauges, some 

licensees may be required to implement additional tracking and security requirements.
53

The regulations in 10 CFR 20.2207, “Reports of transactions involving nationally tracked 

sources,” require that each licensee who manufactures, transfers, receives, disassembles, or 

disposes of a nationally tracked source shall complete and submit a National Source Tracking 

Transaction Report (NSTS). The NSTS is a major security initiative of the NRC. The NSTS is a 

secure, accessible and easy-to-use computer system that tracks high-risk radioactive sources 

from the time they are manufactured or imported through the time of their disposal or export, or 

until they decay enough to no longer be of concern. 

In addition to the regulations, the NRC in 2005 began issuing Orders imposing security 

requirements on licensees who possessed radioactive materials quantities of concern 

(RAMQC). The Orders required licensees to implement enhanced security to control access to 

RAMQC and to protect sensitive security-related information. In 2007, the NRC issued 

additional Orders to the same licensees requiring fingerprinting and criminal history records

checks for unescorted access to RAMQC. The specific radionuclides and associated thresholds 

were based on the Category 1 and Category 2 quantities described in the IAEA’s “Code of 

Conduct on the Safety and Security of Radioactive Sources.”

Applicants and licensees should visit the NRC’s public Web site (www.nrc.gov) and select the 

Nuclear Security tab at the top of the NRC home page for additional information regarding the 

security of licensed material or any enhanced security requirements imposed by Orders. 

Please contact the appropriate regional office for questions regarding the security of licensed 

material or any enhanced security requirements imposed by the Orders.

8.11 Item 11: Waste Management—Gauge Disposal and Transfer

Regulations: 10 CFR 20.2001, 10 CFR 20.2207, 10 CFR 30.36, 10 CFR 30.41, and 

10 CFR 30.51

Criteria: Licensed materials must be disposed of in accordance with NRC requirements by 

transfer to an authorized recipient. Appropriate records must be maintained.

Discussion: When disposing of fixed gauges, licensees must transfer them to an authorized 

recipient. Authorized recipients are the original manufacturer or distributor of the device, a 

commercial firm licensed by the NRC or an Agreement State to accept radioactive waste from 

other persons, or another specific licensee authorized to possess the licensed material (i.e., its 

license specifically authorizes the same radionuclide, form, and use).

Before transferring radioactive material, the licensee must verify that the recipient is properly 

authorized to receive the licensed material using one of the methods described in 

10 CFR 30.41, “Transfer of byproduct material.” In addition, all packages containing radioactive 

sources must be prepared and shipped in accordance with NRC and DOT regulations. Records 

of the transfer must be maintained by the licensee as required by 10 CFR 30.51, “Records.”

If source activity exceeds activities listed in Appendix E to 10 CFR Part 20, the transfer 

transaction must be reported in accordance with 10 CFR 20.2207.

Response from Applicant: The applicant does not need to provide a response to this item 

during the licensing process. However, the licensee should establish and include waste 
54

disposal procedures in its radiation safety program.

References: See IN 86-31 and IN 88-02. INs are available in the “NRC Library” on the NRC’s 

Home Page at http://www.nrc.gov.

The next two items on NRC Form 313 should be completed on the form itself.

8.12 Item 12: License Fees

On NRC Form 313, enter the appropriate fee category from 10 CFR 170.31 and the amount of 

the fee enclosed with the application.

Direct all questions about the NRC’s fees or the completion of Item 12 of NRC Form 313 to the 

Office of the Chief Financial Officer at NRC Headquarters in Rockville, MD, (301) 415-7554. 

Information about fees may also be obtained by calling the NRC’s toll-free number, (800) 368-

5642, extension 415-7554. The e-mail address for fees questions is Fees.Resource@nrc.gov.

8.13 Item 13: Certification

Individuals acting in a private capacity are required to date and sign NRC Form 313. Otherwise, 

a representative of the corporation or legal entity filing the application must sign and date NRC 

Form 313 and include his or her title. The representative signing the application must be 

authorized to make binding commitments and to sign official documents on behalf of the 

applicant. As discussed previously in Chapter 3, “Management Responsibility,” signing the 

application acknowledges management’s commitment to and responsibility for the radiation

protection program. The NRC will return all unsigned applications for proper signature.

Notes:

• It is a criminal offense to make a willful false statement or representation on applications 

or correspondence (18 U.S.C. 1001).

• When the application references commitments, those items become binding and are part

of the license conditions and regulatory requirements
55

9. AMENDMENTS AND RENEWALS TO A LICENSE

It is the licensee’s obligation to keep the license current. If any of the information provided in 

the original application is to be modified or changed, the licensee must submit an application for 

a license amendment before the change takes place. The change is not in effect until the 

amendment has been issued. Also, to continue the license after its expiration date, the licensee 

must submit an application for a license renewal at least 30 days before the expiration date 

(10 CFR 2.109(a), 10 CFR 30.36(a)).

Applicants for license amendment or renewal should do the following:

• Use the most recent guidance in preparing an amendment or renewal request.

• Submit either an NRC Form 313 or a letter requesting amendment or renewal.

• Provide the license number and docket number.

• For renewals, provide a complete and up-to-date application if many outdated 

documents are referenced or there have been significant changes in regulatory 

requirements, the NRC’s guidance, the licensee’s organization, or the licensee’s

radiation protection program. Alternatively, describe clearly the exact nature of the 

changes, additions, and deletions.

9.1 Timely Notification of Transfer of Control

Regulation: 10 CFR 30.34(b)

Criteria: Licensees must provide full information and obtain the NRC’s prior, written consent

before transferring control of the license, or, as some licensees call it, “change of ownership” 

and/or “transferring the license.”

Discussion: Transferring control may be the result of mergers, buyouts, or majority stock 

transfers. Although it is not the NRC’s intent to interfere with the business decisions of 

licensees, it is necessary for licensees to obtain prior NRC written consent to ensure the 

following:

• Radioactive materials are possessed, used, or controlled only by persons who have 

valid NRC licenses or Agreement State licenses.

• Materials are properly handled and secured.

• Persons using these materials are competent and committed to implementing 

appropriate radiological controls.

• A clear chain of custody is established to identify who is responsible for disposition of 

records and licensed material.

• Public health and safety are not compromised by the use of such materials.
56

Response from Applicant: No response is required from an applicant for a new license. 

However, current licensees should refer to NUREG-1556, Volume 15, for more information 

about transfer of ownership.
57

10. APPLICATIONS FOR EXEMPTIONS

Regulations: 10 CFR 19.31, 10 CFR 20.2301, 10 CFR 30.11

Criteria: Licensees may request exemptions from regulations. The licensee must demonstrate 

that the exemption is authorized by law; will not endanger life, property, or the common defense 

and security; and is otherwise in the public interest.

Discussion: Various sections of the NRC’s regulations address requests for exemptions 

(e.g., 10 CFR 19.31, “Application for exemptions”; 10 CFR 20.2301, “Applications for 

exemptions”; 10 CFR 30.11, “Specific exemptions”). These regulations state that the NRC may 

grant an exemption, acting on its own initiative or on an application from an interested person.

Exemptions are not intended to revise regulations or apply to large classes of licensees and are 

generally limited to unique situations. Exemption requests must be accompanied by 

descriptions of the following:

• Exemption requested, basis, and justification for the requested exemption.

• Proposed compensatory safety measures intended to provide a level of health and 

safety equivalent to the regulation for which the exemption is being requested.

• Alternative methods for complying with the regulation and an explanation of why 

compliance with the existing regulation is not feasible.

Until the NRC has granted an exemption in writing, the agency expects strict compliance with 

all applicable regulations.

59

11. TERMINATION OF ACTIVITIES

Regulations: 10 CFR 30.34(b), 10 CFR 30.35(g), 10 CFR 30.36(d), 10 CFR 30.36(j)(1),

10 CFR 30.51(f)

Criteria: The licensee must do the following:

• Notify the NRC, in writing, within 60 days of the occurrence of any of the following:

– Expiration of its license.

– A decision to cease licensed activities permanently at the entire site.

– A decision to cease licensed activities permanently in any separate building or 

outdoor area that contains residual radioactivity such that the building or area is 

unsuitable for release according to NRC requirements.

– No principal activities under the license have been conducted for a period of 

24 months.

 

– No principal activities have been conducted for a period of 24 months in any 

separate building or outdoor area that contains residual radioactivity such that 

the building or area is unsuitable for release according to NRC requirements.

• Submit to the appropriate NRC regional office a completed NRC Form 314, “Certificate 

of Disposition of Materials” (or equivalent information), and a demonstration that the 

premises are suitable for release for unrestricted use (e.g., results of final leak tests).

• Before a license is terminated, send the records important to decommissioning to the 

appropriate NRC regional office. If licensed activities are transferred or assigned in 

accordance with 10 CFR 30.34(b), transfer records important to decommissioning to the 

new licensee in accordance with 10 CFR 30.35(g).

Discussion: Typically, a fixed gauge termination request will meet the above criteria if the 

licensee has performed the following steps:

• Dispose of or transfer the gauges to an NRC or Agreement State licensee authorized to 

possess these devices as described in Section 8.11, “Waste Management—Gauge 

Disposal and Transfer.”

• Copy applicable decommissioning records as described in Section 8.5.2 “Financial 

Assurance and Recordkeeping for Decommissioning.” Typically, this includes transfer 

records and final leak test records. See Section 8.5.2 for additional recordkeeping 

requirements if leaking sealed sources or other incidents that involve the spreading of 

contamination have occurred.

• Submit a completed NRC Form 314 and a copy of the applicable decommissioning 

records to the appropriate NRC regional office.
60

Response from Applicant: The applicant is not required to submit a response to the NRC 

during the initial application. The licensee’s obligations in this matter begin when the license 

expires or at the time the licensee ceases operations, whichever is earlier. These obligations 

are to undertake the necessary decommissioning activities, to submit NRC Form 314 or 

equivalent information, and to perform any other actions summarized in “Criteria” above.

Reference: NRC Form 314 is available at http://www.nrc.gov/reading-rm/doc-collections/forms.
APPENDIX A

U.S. NUCLEAR REGULATORY COMMISSION FORM 313

A-1

United States Nuclear Regulatory Commission Form 313

APPENDIX B

SUGGESTED FORMAT FOR PROVIDING INFORMATION REQUESTED 

IN ITEMS 5 THROUGH 11 OF U.S. NUCLEAR REGULATORY 

COMMISSION FORM 313

B-1

Items 5 & 6: Materials to be Possessed and Proposed Uses

Yes No Radionuclide Manufacturer

or Distributor 

Model No.

Quantity Use as Listed on 

SSD Registration 

Certificate

Specify Other 

Uses Not Listed 

on SSD 

Registration 

Certificate

Cobalt-60 Sealed source 

manufacturer or 

distributor and 

model number:

Device 

manufacturer or 

distributor and 

model number:

______________

Require 

maximum 

possession limits 

per source or 

device and 

include the total 

possession limit.

Yes ❒

Specific 

description of the 

gauge use:

______________

______________

______________

______________

______________

______________

______________

❒ Not applicable

❒ Uses are:

(Submit safety 

analysis 

supporting safe 

use)

Krypton-85 Sealed source 

manufacturer or 

distributor and 

model number:

Device 

manufacturer or 

distributor and 

model number:

______________

Require 

maximum 

possession limits 

per source or 

device and 

include the total 

possession limit.

Yes ❒

Specific 

description of the 

gauge use:

______________

______________

______________

______________

______________

______________

______________

❒ Not applicable

❒ Uses are:

(Submit safety 

analysis 

supporting safe 

use)

Strontium-90 Sealed source 

manufacturer or 

distributor and 

model number:

Device 

manufacturer or 

distributor and 

model number:

______________

Require 

maximum

possession limits 

per source or 

device and 

include the total 

possession limit.

Yes ❒

Specific 

description of the 

gauge use:

______________

______________

______________

______________

______________

______________

______________

❒ Not applicable

❒ Uses are:

(Submit safety 

analysis 

supporting safe 

use)
B-2

Yes No Radionuclide Manufacturer

or Distributor 

Model No.

Quantity Use as Listed on 

SSD Registration 

Certificate

Specify Other 

Uses Not Listed 

on SSD 

Registration 

Certificate

Cesium-137 Sealed source 

manufacturer or 

distributor and 

model number:

Device 

manufacturer or 

distributor and 

model number:

______________

Require 

maximum 

possession limits 

per source or 

device and 

include the total 

possession limit..

Yes ❒

Specific 

description of the 

gauge use:

______________

______________

______________

______________

______________

______________

______________

❒ Not applicable

❒ Uses are:

(Submit safety 

analysis 

supporting safe 

use)

Americium241Sealed source 

manufacturer or 

distributor and 

model number:

Device 

manufacturer or 

distributor and 

model number:

______________

Require 

maximum 

possession limits 

per source or 

device and 

include the total 

possession limit.

Yes ❒

Specific 

description of the 

gauge use:

______________

______________

______________

______________

______________

______________

______________

❒ Not applicable

❒ Uses are:

(Submit safety 

analysis 

supporting safe 

use)

Other Isotope 

(Specify):

Sealed source 

manufacturer or 

distributor and 

model number:

Device 

manufacturer or 

distributor and 

model number:

______________

Require 

maximum 

possession limits 

per source or 

device and 

include the total 

possession limit.

Yes ❒

Specific 

description of the 

gauge use:

______________

______________

______________

______________

______________

______________

______________

❒ Not applicable

❒ Uses are:

(Submit safety 

analysis 

supporting safe 

use)

Is financial assurance required? If yes, submit evidence of financial assurance
B-3

Items 7 Through 11: Training and Experience, 

Facilities and Equipment, Radiation Safety Program, 

and Waste Disposal

Item No. and Title Suggested Response Yes Alternative 

Procedures Attached

7. Individual(s) 

Responsible For 

Radiation Safety 

Program and 

Their Training 

and Experience

7.1 Radiation safety 

officer

Name: 

Provide documentation of the proposed 

radiation safety officer’s training and 

experience.

Submit applicable documentation.

7. Individual(s) 

Responsible For 

Radiation Safety 

Program and 

Their Training 

and Experience

7.2 Authorized 

users

Before using licensed materials, authorized 

users will have successfully completed one of 

the training courses described in the “Criteria” 

part of the section entitled “Authorized Users” 

in the current version of NUREG-1556, 

Volume 4, “Consolidated Guidance about 

Materials Licenses: Program-Specific 

Guidance about Fixed Gauge Licenses.”

❒ ❒

8. Training for 

Individuals 

Who in the 

Course of 

Employment 

are Likely to 

Receive 

Occupational 

Doses of 

Radiation in 

Excess of 

1 mSv 

(100 mrem) in 

a Year 

(Occupationally 

Exposed 

Workers) and 

Ancillary 

Personnel

The applicant is not required to, and should 

not, submit its training program for individuals 

who in the course of employment are likely to 

receive occupational doses of radiation in 

excess of 1 mSv (100 mrem) in a year 

(occupationally exposed workers) and 

ancillary personnel to the NRC for review 

during the licensing phase.

Need not be submitted with 

application.
B-4

Item No. and Title Suggested Response Yes Alternative 

Procedures 

Attached

9. Facilities and 

Equipment

We will ensure that the location of each fixed gauge 

meets the criteria in Section 8.9, “Facilities and 

Equipment,” in the current version of NUREG-1556, 

Volume 4, “Consolidated Guidance about Materials 

Licenses: Program-Specific Guidance about Fixed 

Gauge Licenses.”

OR

Provide confirmation that the fixed gauge is 

secured to prevent unauthorized removal or access 

and submit specific information demonstrating that

the proposed conditions will not impact the safety 

or integrity of the source or device. Address any 

instances in which the proposed conditions exceed 

any conditions listed in the SSD registration 

certificate.

❒

❒

10. Radiation Safety 

Program – Audit

Program

The applicant is not required to, and should not, 

submit its audit program to the NRC for review 

during the licensing phase. The audit program will 

be reviewed during NRC inspections.

Need not be submitted 

with application.

10. Radiation Safety 

Program –

Instruments

Surveys according to 10 CFR 20.1501 will be 

performed by a person specifically authorized by 

the NRC or an Agreement State to perform these 

surveys.

OR

We will use instruments that meet the criteria in 

Section 8.10.2, “ Instruments,” in the current 

version of NUREG-1556, Volume 4, “Consolidated 

Guidance about Materials Licenses: ProgramSpecific Guidance about Fixed Gauge Licenses”

and one of the following:

Each survey meter will be calibrated by the 

manufacturer or other person authorized by the 

NRC or an Agreement State to perform survey 

meter calibrations.

OR

We will implement the model survey instrument 

calibration program in Appendix F, “Model 

Survey Instrument Calibration Program” in the 

current version of NUREG-1556, Volume 4.

OR

We will submit alternative calibration 

procedures for NRC review. (Attach 

procedures).

❒

❒

❒

❒

❒

❒
B-5

Item No. and Title Suggested Response Yes Alternative 

Procedures 

Attached

10. Radiation Safety 

Program –

Material Receipt 

and Accountability

Physical inventories will be conducted at intervals 

not to exceed 6 months or at other intervals 

approved by the NRC to account for all sealed 

sources and devices received and possessed 

under the license.

❒ ❒

10. Radiation Safety 

Program –

Occupational Dose

We will maintain, for inspection by the NRC, 

documentation demonstrating that unmonitored 

individuals are not likely to receive, in one year, a 

radiation dose in excess of 10 percent of the 

allowable limits in 10 CFR Part 20.

OR

We will provide dosimetry processed and evaluated 

by an NVLAP-approved processor that is 

exchanged at a frequency recommended by the 

processor.

❒

❒

❒

10. Radiation Safety 

Program – Public

Dose

The applicant is not required to submit a response 

to the public dose section in a license application. 

This matter will be examined during NRC 

inspections.

Need not be submitted 

with application.

10. Radiation Safety 

Program –

Operating and 

Emergency

Procedures

If the gauge meets one or more of the safety 

conditions specified in the “Discussion” part of 

Section 8.10.6, “Operating and Emergency

Procedures,” in the current version of 

NUREG-1556, Volume 4, “Consolidated Guidance 

about Materials Licenses: Program-Specific 

Guidance about Fixed Gauge Licenses,” state the 

following:

Operating, emergency, and security procedures 

will be developed, implemented, maintained, and 

distributed, and will meet the criteria in 

Section 8.10.6, “Operating and Emergency

Procedures,” in the current version of 

NUREG-1556, Volume 4.

AND/OR

For each gauge requested that does not meet one 

or more of the safety conditions specified in the 

“Discussion” part of Section 8.10.6, “Operating and 

Emergency Procedures,” in the current version of 

NUREG-1556, Volume 4, provide your operating, 

emergency, and lock-out (if applicable) procedures 

to the NRC for review.

❒

❒

Procedures 

Attached

❒
B-6

Item No. and Title Suggested Response Yes Alternative 

Procedures 

Attached

10. Radiation Safety 

Program – Leak

Tests

Leak tests will be performed at intervals approved 

by the NRC or an Agreement State and specified in 

the SSD registration certificate. Leak tests will be 

performed by an organization authorized by the 

NRC or an Agreement State to provide leak testing 

services to other licensees or using a leak test kit 

supplied by an organization authorized by the NRC 

or an Agreement State to provide leak test kits to 

other licensees and according to the kit supplier’s 

instructions.

OR

We will implement the model leak test program 

published in Appendix I of the current version of 

NUREG-1556, Volume 4, “Consolidated Guidance 

about Materials Licenses: Program-Specific 

Guidance about Fixed Gauge Licenses.”

❒

❒

❒

10. Radiation Safety 

Program –

Maintenance

We will implement and maintain procedures for 

routine maintenance of our fixed gauges according 

to each manufacturer’s or distributor’s written 

recommendations and instructions.

ROUTINE MAINTENANCE

The gauge manufacturer, distributor, or other 

person authorized by the NRC or an Agreement 

State will perform nonroutine operations, such as 

installation, initial radiation survey, repair and 

maintenance of radiological safety components, 

relocation, replacement, alignment, removal from 

service, and disposal of sealed sources.

NONROUTINE MAINTENANCE OPERATIONS

❒

❒

❒

❒

The 

information 

listed in 

Appendix J, 

“Information 

Needed to 

Support 

Applicant’s 

Request to 

Perform 

Nonroutine 

Operations,”

is attached.

10. Radiation Safety 

Program –

Transportation

The applicant is not required to submit its response 

about transportation during the licensing process. 

This issue will be reviewed during NRC 

inspections.

Need not be submitted 

with application.
B-7

Item No. and Title Suggested Response Yes Alternative 

Procedures 

Attached

10. Radiation Safety 

Program – Fixed

Gauges Used at 

Temporary Job 

Sites

We will not use fixed gauges at temporary job sites.

OR

We will develop, implement, maintain, and 

distribute procedures that meet the criteria in 

Section 8.10.10, “Fixed Gauges Used at 

Temporary Job Sites,” in the current version of 

NUREG-1556, Volume 4, “Consolidated Guidance 

about Materials Licenses: Program-Specific 

Guidance about Fixed Gauge Licenses,” and 

provide copies of these procedures to all gauge 

users and make them available at all temporary job 

sites.

❒

❒ ❒

10. Radiation Safety 

Program –

Security Program

Licensees must ensure the security and control of 

licensed material. The regulations in 

10 CFR 20.1801 and 20.1802 require licensees to 

secure radioactive materials from unauthorized 

removal or access while in storage and to control 

and maintain constant surveillance over licensed 

material that is not in storage.

Need not be submitted 

with application.

11. Waste 

Management –

Gauge Disposal & 

Transfer

The applicant is not required to submit a response 

to waste management during the licensing 

process; however, the licensee should develop, 

implement, and maintain gauge transfer and 

disposal procedures in its radiation protection 

program.

Need not be submitted 

with application.

APPENDIX C

TYPICAL DUTIES AND RESPONSIBILITIES OF THE RADIATION

SAFETY OFFICER

C-1

The RSO’s duties and responsibilities (illustrated in Figure 8.2) typically include ensuring the 

following:

• Licensed activities that the RSO considers unsafe are stopped.

• Radiation exposures are kept as low as is reasonably achievable (ALARA).

• Up-to-date operating and emergency procedures are developed, implemented, 

maintained, and distributed.

• Individuals who use fixed gauges are properly trained.

• Possession, installation, relocation, use, storage, routine maintenance, and nonroutine 

operations of fixed gauges are consistent with the limitations in the license, the Sealed 

Source and Device registration certificate(s), and the manufacturer’s or distributor’s 

recommendations and instructions.

• Safety consequences of nonroutine operations are analyzed before conducting any such 

activities that have not been previously analyzed.

• Nonroutine operations are performed by the manufacturer, distributor, or person 

specifically authorized by the NRC or an Agreement State.

• Prospective evaluations are performed to demonstrate that unmonitored individuals are 

not likely to receive, in one year, a radiation dose in excess of 10 percent of the 

allowable limits or that personnel monitoring devices are provided.

• When necessary, personnel monitoring devices are used and exchanged at the proper 

intervals, and records of the results of such monitoring are maintained.

• Documentation is maintained to demonstrate, by measurement or calculation, that the 

dose to the individual member of the public likely to receive the highest dose from the 

licensed operation does not exceed the annual limit in 10 CFR 20.1301, “Dose limits for 

individual members of the public.”

• Fixed gauges are properly secured.

• Proper authorities are notified in case of accident, damage to, or malfunction of 

fixed-gauges, fire, loss, or theft.

• Unusual occurrences involving the fixed gauge (e.g., malfunctions or damage) are 

investigated, cause(s) are determined, and appropriate corrective action(s) are taken.

• Audits are performed and documented at least annually, and corrective actions are 

taken.

• When the licensee identifies violation(s) of regulations or license conditions or program 

weaknesses, corrective action(s) are developed, implemented, and documented.
C-2

• Licensed material is transported in accordance with all applicable NRC and DOT 

requirements.

• Licensed material is disposed of properly.

• All required records are maintained.

• An up-to-date license is maintained, and amendment and renewal requests are 

submitted in a timely manner.

• Posting of documents required by 10 CFR 19.11, “Posting of notices to workers,” and 10 

CFR 21.6, “Posting Requirements,” or posting a notice indicating where these 

documents can be examined. Required documents include the following:

− 10 CFR Part 19;

− 10 CFR Part 20;

− license documents;

− operating procedures;

− any notice of violation involving radiological working conditions, proposed 

imposition of civil penalty, order, or any licensee response;

− NRC Form 3, “Notice to Employees”;

− 10 CFR Part 21;

− Section 206 of the Energy Reorganization Act of 1974; and

− procedures adopted under 10 CFR Part 21.
C-3

Model Delegation of Authority to RSO

Memo To: Radiation Safety Officer

From: Chief Executive Officer

Subject: Delegation of Authority

You, _______________________________, have been appointed radiation safety officer and 

are responsible for ensuring the safe use of radiation. You are responsible for managing the 

Radiation Protection Program, identifying radiation protection problems, initiating, 

recommending, or providing corrective actions, verifying implementation of corrective actions, 

stopping unsafe activities, and ensuring compliance with regulations. You are hereby delegated 

the authority necessary to meet those responsibilities, including prohibiting the use of byproduct 

material by employees who do not meet the necessary requirements and shutting down 

operations, when justified, to maintain radiation safety. You are required to notify management 

if staff does not cooperate and does not address radiation safety issues. In addition, you are 

free to raise issues with the U.S. Nuclear Regulatory Commission at any time. It is estimated 

that you will spend _____ hours per week conducting radiation protection activities.

____________________________________ ____________________

Signature of Management Representative Date

I accept the above responsibilities,

_____________________________________ ____________________

Signature of Radiation Safety Officer Date

cc: Affected department heads

APPENDIX D

CRITERIA FOR ACCEPTABLE TRAINING FOR AUTHORIZED USERS 

AND RADIATION SAFETY OFFICERS

D-1

Course Content

Classroom training may be in the form of lectures, videos, computer-based sessions, or 

self-study lessons that emphasize practical subjects important to the safe use of the gauge

including:

Radiation Safety:

• radiation versus contamination

• internal versus external exposure

• biological effects of radiation

• types and relative hazards of radioactive material possessed

• ALARA concept

• use of time, distance, and shielding to minimize exposure

• location of sealed source within the gauge

Regulatory Requirements:

• applicable regulations

• license conditions, amendments, and renewals

• locations of use and storage of radioactive materials

• material control and accountability

• annual audit of radiation safety program

• transfer and disposal

• recordkeeping

• prior events involving fixed gauges

• handling incidents

• recognizing and ensuring that radiation warning signs are visible and legible

• licensing and inspection by regulatory agency

• need for complete and accurate information

• employee protection

• deliberate misconduct

Practical Explanation of the Theory and Operation for Each Gauge Possessed by the Licensee:

• operating and emergency procedures

• routine versus nonroutine maintenance

• lock-out procedures

Supervised, Hands-On Experience (On-the-Job Training) Involving:

• operating procedures

• test runs of emergency procedures

• routine maintenance

• lock-out procedures

Note: On-the-job training must be done under the supervision of an AU or RSO.
D-2

Training Assessment

Management will ensure that proposed AUs are qualified to work independently with each type 

of gauge with which they may work. Management will ensure that proposed RSOs are qualified 

to work independently with and are knowledgeable of the radiation safety aspects of all types of 

gauges that may be possessed by the applicant.

Course Instructor Qualifications

Instructors should have, at a minimum, the following:

• successful completion of a fixed gauge manufacturer’s or distributor’s course for users 

(or equivalent);

• successful completion of an 8-hour radiation safety course or radiation safety officer 

training course; and

• documentation of 8 hours of hands-on experience with fixed gauges.

OR

• The applicant may submit a description of alternative training and experience for the 

course instructor.

Note: Additional training is required for those applicants intending to perform nonroutine 

operations, such as installation, initial radiation survey, repair, and maintenance of components 

related to the radiological safety of the gauge, gauge relocation, replacement, and disposal of 

sealed sources, alignment, or removal of a gauge from service. See Appendix J, “Information 

Needed to Support Applicant’s Request to Perform Nonroutine Operations.”
APPENDIX E

Suggested Fixed Gauge Audit Checklist

E-1

Note: All areas indicated in audit notes may not be applicable to every license and may not 

need to be addressed during each audit. For example, licensees do not need to address areas 

that do not apply to their activities, and activities that have not occurred since the last audit need 

not be reviewed during the next audit.

Licensee’s name License No. 

Date of This Audit Date of Last Audit 

Audit Date Range 

Date 

(Auditor Signature)

Date 

(Management Signature)

1. Audit History

A. Last audit of this location conducted on (date) ______________________

B. Were previous audits conducted periodically (at least annually)? (10 CFR

20.1101)

C. Were records of previous audits maintained? (10 CFR 20.2102)

D. Were any deficiencies identified during the last two audits or two years, 

whichever is longer?

E. Were corrective actions taken? (Look for repeated deficiencies).

2. Organization and Scope of Program

A. If the mailing address or places of use changed, was the license amended?

B. If ownership changed or bankruptcy was filed, did the licensee obtain prior NRC 

consent or notify the NRC?

C. If the licensee changed the RSO, was the license amended?

D. Sealed Sources and Devices

1. Does the license authorize all of the NRC-regulated radionuclides 

contained in the gauges possessed?

2. Are the gauges as described in the SSD registration certificate?

3. Are copies of (or access to) SSD registration certificates available?

4. Are manufacturers’ or distributors’ manuals for operation and 
E-2

maintenance available? (10 CFR 32.210)

5. Are the actual uses of gauges consistent with the authorized uses listed 

on the license?

6. Are the locations of the gauges compatible with the “Conditions of Normal 

Use” and “Limitations and/or Other Considerations of Use” on the SSD 

registration certificates?

3. Training and Instructions to Workers

A. Were all workers who are likely to exceed 1 mSv (100 mrem) in a year instructed 

per 10 CFR 19.12? Was refresher training provided, as needed? Were records 

maintained?

B. Did each AU receive training and instruction given at the time of gauge 

installation or equivalent training and instruction before using gauges?

C. Are training records maintained for each AU?

D. Did individuals who perform nonroutine operations receive training before 

performing these operations?

E. Did interviews with AUs reveal that they know the operating, emergency, and 

security procedures?

F. Did this audit include observations of AUs using the gauge?

G. Did this audit include observations of workers performing routine cleaning and 

lubrication on the gauge?

H. Was DOT hazardous materials (HAZMAT) training provided, if required? (49 

CFR 172.700, 172.701, 172.702, 172.704)

4. Radiation Survey Instruments

A. If the licensee is required to possess a survey meter, does it meet the NRC’s 

criteria? (10 CFR 20.1501)

B. Are calibration records maintained, if applicable? (10 CFR 20.2103(a))

C. If the licensee does not possess a survey meter, are specific plans made to have 

one available?

5. Gauge Inventory

A. Is a record kept showing the receipt of each gauge? (10 CFR 30.51(a)(1))

B. Are all gauges physically inventoried every 6 months?

C. Are records of inventory results with appropriate information maintained?
E-3

6. Personnel Radiation Protection

A. Are considerations for keeping doses as low as is reasonably achievable 

(ALARA) incorporated into the radiation protection program? 

(10 CFR 20.1101(b))

B. Were prospective evaluations performed showing that unmonitored individuals 

receive ≤10 percent of the NRC limits in 10 CFR 20.1201(a)? (10 CFR 

20.1502(a))

C. Did unmonitored individuals’ activities change during the year in a way that could 

put them over 10 percent of the NRC limit? If yes, was a new evaluation 

performed?

D. If external dosimetry is required (i.e., when individuals are likely to receive 

>10 percent of the NRC limit), is dosimetry provided to these individuals? If yes, 

address the following:

1. Is the dosimetry supplier approved by the National Voluntary Laboratory 

Accreditation Program? (10 CFR 20.1501(c))

2. Are the dosimeters exchanged at the appropriate frequency?

3. Are dosimetry reports reviewed and signed by the RSO when they are 

received?

4. Are the records based on NRC Forms or the equivalent? 

(10 CFR 20.2104(d) and 20.2106(c))

a. Is NRC Form 4, “Cumulative Occupational Exposure History,” 

completed?

b. Is NRC Form 5, “Occupational Dose Record for a Monitoring 

Period,” completed?

5. Are there any declared pregnant workers?

a. If a worker declared her pregnancy, did the licensee comply with 

10 CFR 20.1208, “Dose equivalent to an embryo/fetus”?

b. Were records kept of doses to an embryo/fetus per 

10 CFR 20.2106(e)?

E. Are records of exposures, surveys, monitoring, and evaluations maintained? 

(10 CFR 20.2102, “Records of radiation protection programs”; 10 CFR 20.2103, 

“Records of surveys”; 10 CFR 20.2106, “Records of individual monitoring 

results”)
E-4

7. Public Dose

A. Is public access to gauges controlled in a manner to keep doses below 1 mSv

(100 mrem) in a year? (10 CFR 20.1301(a)(1))

B. Has a survey or evaluation been performed per 10 CFR 20.1501(a)? Have there 

been any additions or changes to the storage, security, or use of the surrounding 

areas that would necessitate a new survey or evaluation?

C. Do unrestricted area radiation levels exceed 0.02 mSv (2 mrem) in any one 

hour? (10 CFR 20.1301(a)(2))

D. Is gauge access controlled in a manner that would prevent unauthorized use or 

removal? (10 CFR 20.1801)

E. Are records of surveys maintained? (10 CFR 20.2103, 10 CFR 20.2107)

8. Operating and Emergency Procedures

Note: An ideal way to assess the adequacy and adherence to operating procedures is 

by observing work in progress.

A. Have operating and emergency procedures been updated to incorporate any 

new elements, practices, or requirements?

B. Does each individual working with the gauges have current copies of the 

operating and emergency procedures (including lock-out procedures and 

emergency telephone numbers)?

C. Is a lock-out warning sign posted at each entryway to an area where it is possible 

to be exposed to the beam?

D. Did any emergencies occur?

1. If so, were they handled properly?

2. Were appropriate corrective actions taken?

9. Leak Tests

A. Were sealed source leak tests performed at the prescribed intervals?

B. Was the leak test performed in accordance with license requirements?

C. Are records of leak test results retained with all of the required information 

included?

D. Were any sources found to be leaking and if yes, was the NRC notified?
E-5

10. Maintenance of Gauges

A. Are manufacturers’ or distributors’ procedures followed for routine cleaning and 

lubrication of gauges?

B. Was each on-off mechanism tested for proper operation every 6 months or at 

other approved intervals?

C. Are repair and maintenance of components related to the radiological safety of 

the gauge performed by the manufacturer, distributor, or person specifically 

authorized by the NRC or an Agreement State and according to license 

requirements (e.g., extent of work, procedures, dosimetry, survey instrument, 

compliance with 10 CFR 20.1301 limits)?

D. Are labels, signs, and postings identifying gauges containing radioactive material,

radiation areas, and lock-out procedures and warnings clean and legible?

11. Transportation

Note: This section will not apply if you have not transported gauges during the period covered 

by this audit.

A. Were DOT-7A or other authorized packages used? (49 CFR 173.415, 

“Authorized Type A Packages”; 49 CFR 173.416, “Authorized Type B Packages”)

B. Are package performance test records on file?

C. For any special form source, is the International Atomic Energy Agency 

Certificate of Competent Authority or other safety analysis documentation 

maintained on file? (49 CFR 173.476(a))

D. Were packages properly labeled? (49 CFR 172.403, “Class 7 (radioactive) 

material”)

E. Were packages properly marked? (49 CFR 172.301, “General marking 

requirements for non-bulk packagings”; 49 CFR 172.304, “Marking 

requirements”; 49 CFR 172.310, “Class 7 (radioactive) materials”; 49 CFR 

172.324, “Hazardous substances in non-bulk packagings”)

F. Were packages closed and sealed before transport? (49 CFR 173.475(f))

G. Were shipping papers prepared and used? (49 CFR 172.200(a))

H. Did shipping papers contain proper entries (i.e., identification number (United 

Nations (UN) number), shipping name, hazard class, total quantity, package 

type, nuclide, reportable quantity (RQ), if applicable, physical and chemical form, 

activity (International System of Units required), category of label, Transportation 

Index (TI), shipper’s name, certification and signature, emergency response 

phone number, cargo aircraft only (if applicable)? (49 CFR 172.200, 

“Applicability”; 49 CFR 172.201, “Preparation and retention of shipping papers”;

49 CFR 172.202, “Description of hazardous material on shipping papers”;
E-6

49 CFR 172.203, “Additional description requirements”; 49 CFR 172.204, 

“Shipper’s certification”; 49 CFR 172.604, “Emergency response telephone 

number”)

I. Were shipping papers within the driver’s reach and readily accessible during 

transport? (49 CFR 177.817(e))

J. Were packages secured against movement? (49 CFR 177.834,”General 

requirements”)

K. Were placards on the vehicle, if needed? (49 CFR 172.504, “General placarding 

requirements”)

L. If applicable, were proper overpacks used? (49 CFR 173.25, “Authorized 

packagings and overpacks”)

M. Were any incidents reported to DOT? (49 CFR 171.15, “Immediate notice of 

certain hazardous materials incidents”; 49 CFR 171.16, “Detailed hazardous 

materials incident reports”)

12. Auditor’s Independent Survey Measurements (If Made)

Describe the type, location, and results of measurements. Does any radiation level 

exceed regulatory limits? (10 CFR 20.1501(a), 10 CFR 20.1502(a))

13. Notifications and Reports

A. Was any radioactive material lost or stolen? Were reports made? 

(10 CFR 20.2201)

B. Did any reportable incidents occur? Were reports made? (10 CFR 20.2202, 

“Notification of incidents”; 10 CFR 21.21, “Notification of failure to comply or 

existence of a defect and its evaluation”; 10 CFR 30.34, “Terms and conditions of

licenses”; 10 CFR 30.36, “Expiration and termination of licenses and 

decommissioning of sites and separate buildings or outdoor areas”; 10 CFR 

30.50, “Reporting requirements”)

C. Did any overexposures or high radiation levels occur? Were reports made? 

(10 CFR 20.2203)

D. If any events (as described in items a through c above) did occur, what was the 

root cause? Were corrective actions appropriate?

E. Is the licensee aware of the telephone number for the NRC Emergency 

Operations Center, which is 301-816-5100?

14. Posting and Labeling

A. Is NRC Form 3, “Notice to Workers,” posted? (10 CFR 19.11)

B. Are NRC regulations and license documents posted, or is a notice posted stating 
E-7

where these documents are located? (10 CFR 19.11, 10 CFR 21.6)

C. Is there any other posting and labeling? (10 CFR 20.1902, “Posting 

requirements”; 10 CFR 20.1904, “Labeling containers”)

15. Record Keeping for Decommissioning

A. Are records kept of information important to decommissioning? 

(10 CFR 30.35(g))

B. Do records include all information outlined in 10 CFR 30.35(g)?

16. Bulletins and Information Notices

A. Are NRC Bulletins, NRC Information Notices (INs), and FSME Newsletters

received?

B. Is appropriate training and action taken in response to these?

17. Special License Conditions or Issues

Did the auditor review special license conditions or other issues (e.g., nonroutine 

maintenance)?

18. Evaluation of Other Factors

A. Is senior licensee management appropriately involved with the radiation 

protection program and/or RSO oversight?

B. Does the RSO have sufficient time to perform his or her radiation safety duties?

C. Does the licensee have sufficient staff to support the radiation protection 

program?

19. Deficiencies Identified in Audit and Corrective Actions

A. Summarize problems/deficiencies identified during the audit.

B. If problems/deficiencies are identified in this audit, describe corrective actions 

planned or taken. Are corrective actions planned or taken at all licensed 

locations (not just the location audited)? Include date(s) when corrective actions 

are implemented.

C. Provide any other recommendations for improvement.

APPENDIX F

MODEL SURVEY INSTRUMENT CALIBRATION PROGRAM

F-1

Training

Before independently calibrating survey instruments, an individual should complete both 

classroom and on-the-job training as follows:

• Classroom training may be in the form of lecture, video, computer-based, or self-study 

and will cover the following subject areas:

− principles and practices of radiation protection;

− radioactivity measurements, monitoring techniques, and the use of radiation 

detection instruments;

− mathematics related to the use and measurement of radioactivity; and

− biological effects of radiation.

• On-the-job training will be considered complete if the individual has completed both of 

the following:

- observed authorized personnel performing survey instrument calibration; and

- conducted survey meter calibrations under the supervision and in the physical 

presence of an individual already authorized to perform calibrations.

Facilities and Equipment

• To reduce doses received by individuals not calibrating instruments, calibrations will be 

conducted in an isolated area of the facility or at times when no one else is present.

• Individuals conducting calibrations will wear assigned dosimetry.

• Individuals conducting calibrations will use a calibrated and operable survey instrument 

to ensure that unexpected changes in exposure rates are identified and corrected.

Model Procedure for Calibrating Survey Instruments

• A radioactive sealed source(s) will be used for calibrating survey instruments, and this 

source will incorporate the following:

- approximate a point source;

- have its apparent source activity or the exposure rate at a given distance 

traceable by documented measurements to a standard certified to be within 

± 5 percent accuracy by the National Institutes of Standards and Technology 

(NIST);

- contain a radionuclide that emits radiation of identical or similar type and energy 

as the sealed sources that the instrument will normally measure; and
F-2

- be strong enough to emit a radiation field that is representative of the field being 

emitted by the gauge. For calibration of instruments intended to measure 

gamma radiation, the exposure rate should be at least 7.7 microcoulomb per 

kilogram per hour (30 milliroentgen per hour) at 100 centimeters (e.g., 3.1 

gigabecquerels (85 millicuries) of cesium-137 or 780 megabecquerels 

(21 millicuries) of cobalt-60).

• Inverse square and radioactive decay laws must be used to correct changes in exposure 

rate due to changes in distance or source decay.

• A record must be made of each survey meter calibration.

• A single point on a survey meter scale may be considered satisfactorily calibrated if the 

indicated exposure rate differs from the calculated exposure rate by less than 

± 20 percent.

• There are three kinds of scales frequently used on radiation survey meters. They are 

calibrated either as described in American National Standards Institute (ANSI) 

N323A-1997, “American National Standard Radiation Protection Instrumentation Test 

and Calibration – Portable Survey Instruments,” or as follows:

- Meters on which the user selects a linear scale must be calibrated at not fewer 

than two points on each scale. The points will be at approximately 1/3 and 2/3 of 

the decade.

- Meters that have a multidecade logarithmic scale must be calibrated at one point 

(at the least) on each decade and not fewer than two points on one of the

decades. Those points will be approximately 1/3 and 2/3 of the decade.

- Meters that have an automatically ranging digital display device for indicating 

exposure rates must be calibrated at one point (at the least) on each decade and 

at no fewer than two points on one of the decades. Those points should be at 

approximately 1/3 and 2/3 of the decade.

• Readings above 50 microcoulomb per kilogram per hour (200 milliroentgen per hour)

need not be calibrated; however, higher scales should be checked for operation and 

approximately correct response.

• Survey meter calibration reports will indicate the procedure used and the results of the 

calibration. The reports will include the following:

- the owner or user of the instrument;

- a description of the instrument that includes the manufacturer’s name, model 

number, serial number, and type of detector;

- a description of the calibration source, including the exposure rate at a specified 

distance on a specified date, and the calibration procedure;
F-3

- for each calibration point, the calculated exposure rate, the indicated exposure 

rate, the deduced correction factor (the calculated exposure rate divided by the 

indicated exposure rate), and the scale selected on the instrument;

- the exposure reading indicated with the instrument in the “battery check” mode (if 

available on the instrument);

- for instruments with external detectors, the angle between the radiation flux field 

and the detector (i.e., parallel or perpendicular);

- for instruments with internal detectors, the angle between radiation flux field and 

a specified surface of the instrument;

- for detectors with removable shielding, an indication whether the shielding was in 

place or removed during the calibration procedure;

- the exposure rate from a check source, if used; and

- the signature of the individual who performed the calibration and the date on 

which the calibration was performed.

The following information will be attached to the instrument as a calibration sticker or tag:

• the source used to calibrate the instrument;

• the proper deflection in the battery check mode (unless this is clearly indicated on the 

instrument);

• for each scale or decade not calibrated, an indication that the scale or decade was 

checked only for function but not calibrated;

• the date of calibration and the next calibration due date; and

• the apparent exposure rate from the check source, if used.

References: Detailed information about survey instrument calibration may be obtained by 

referring to ANSI N323A-1997, “American National Standard Radiation Protection 

Instrumentation Test and Calibration – Portable Survey Instruments.” Copies may be ordered 

online at http://www.ansi.org.

APPENDIX G

DOSIMETRY-RELATED GUIDANCE

G-1

Part 1: Guidance for Demonstrating that Unmonitored Individuals Are Not Likely To 

Exceed 10 Percent of the Allowable Limits

Dosimetry is required for individuals likely to receive, in 1 year from sources external to the 

body, a dose in excess of 10 percent of the applicable regulatory limits in Title 10 of the Code of 

Federal Regulations (10 CFR) 20.1201, “Occupational dose limits for adults.” Therefore, a 

licensee should evaluate the doses its workers receive in performing their duties to assess 

whether dosimetry is required.

Example

Note: The examples in this appendix use conventional units. The conversions to International 

System of Units are as follows: 1 foot = 0.305 meter; 0.01 mSv = 1 mrem.

A gauge manufacturer has estimated the doses to the extremities and whole body of a person 

replacing the assay plate on one of its series of gauges. Each gauge in the series is authorized 

to contain up to 7.4 gigabecquerels (200 millicuries) of Cs-137. The manufacturer based its 

estimate on observations of individuals performing the recommended procedure according to 

good radiation safety practices. The manufacturer had the following information:

• time needed to perform the entire procedure (e.g., 15 minutes)

• expected dose rate received by the whole body of the individual, associated with the 

shielded source and determined using measured or manufacturer-determined data 

(e.g., 0.02 mSv per hour (2 mrem per hour) at 46 centimeters (18.1 inches) from the 

shield)

• time the hands were exposed to the shielded source (e.g., 6 minutes)

• expected dose rate received by the extremities of the individual, associated with the 

shielded source and determined using measured or manufacturer-determined data on 

contact with the shield (e.g., 0.15 mSv per hour (15 mrem per hour)).

From this information, the manufacturer estimated that the individual performing each routine 

cleaning and lubrication could receive the following:

• less than 0.005 mSv (0.5 mrem) to the whole body and

• 0.015 mSv (1.5 mrem) to the hands.

The applicable whole body dose limit is 50 mSv (5 rem) per year and 10 percent of that value is 

5 mSv (500 mrem) per year. If one of these procedures delivers 0.005 mSv (0.5 mrem), then an 

individual could perform 1,000 of these procedures each year and remain within 10 percent of 

the applicable limit.

The applicable extremity dose limit is 500 mSv (50 rem) per year and 10 percent of that value is 

50 mSv (5 rem or 5,000 mrem) per year. If one of these procedures delivers 0.015 mSv 

(1.5 mrem), then an individual could perform 3,333 of these procedures each year and remain 

within 10 percent of the applicable limit.

Based on the above specific situation, no dosimetry is required if a worker performs fewer than 

1,000 routine maintenance procedures per year.
G-2

Guidance to Licensees

Licensees who wish to demonstrate that they are not required to provide dosimetry to their 

workers must prepare a written evaluation similar to that shown in the example above. The 

expected dose rates, times, and distances used in the above example may not be appropriate 

to individual licensee situations. In their evaluations, licensees must use information 

appropriate to the various types of gauges on which they will perform routine cleaning and 

lubrication. This information is generally available from gauge manufacturers or the SSD 

registration certificate maintained by the NRC or Agreement State.

Table G.1 may be helpful in documenting a licensee’s evaluation.

Licensees should review evaluations periodically and revise them as needed. They should 

check assumptions used in their evaluations to ensure that the assumptions are up-to-date and 

accurate. For example, if workers became lax in following good radiation safety practices in the 

example used above, the extremities could be closer to the unshielded source, and the workers 

would receive more exposure than 0.15 mSv (15 mrem) per minute. Alternatively, workers 

could perform the task more slowly than the estimated 10 minutes total and 3 minutes with their

hands near the unshielded source. Also, using new gauges containing sources of different 

activities, different radionuclides, or different cleaning and lubrication procedures requires a new 

evaluation.

Table G.1 Dosimetry Evaluation

Dosimetry Evaluation for ___________________ Model ___________ Fixed Gauge

A. Time needed to perform the entire routine 

cleaning and lubrication procedure on the gauge

_________ minutes _________ hour

(divide # of minutes 

by 60)

B. Expected whole-body dose rate that the individual 

will encounter, determined using measured or 

manufacturer-provided data

____________ mrem/hour

C. Time the hands were exposed to the unshielded 

source

_________ minutes _________ hour

D. Expected extremity dose rate that the individual 

will encounter, determined using measured or 

manufacturer-provided data for the unshielded 

source at the typical distance from the hands to 

the unshielded source

____________ mrem/hour

Estimated Whole Body Dose Equivalent*

Formula: (_____hours in Row A) x (______mrem/hour in Row B) = (_____estimated mrem) x (_____# 

of cleaning and lubrications conducted each year) = _______ Whole Body Dose mrem

Estimated Extremity Dose Equivalent**

Formula: (_____hours in Row C) x (______mrem/hour in Row D) = (_____estimated mrem) x (_____# 

of cleaning and lubrications conducted each year) = _______ Extremity Dose mrem

* An expected Whole Body Dose Equivalent less than 500 mrem requires no dosimetry.

** An expected Extremity Dose Equivalent less than 5000 mrem requires no dosimetry.
G-3

Part 2: Guidance for Demonstrating that Individual Members of the Public Will Not 

Receive Doses Exceeding the Allowable Limits

Licensees must ensure the following:

The radiation dose received by individual members of the public does not exceed 

1 mSv (100 mrem) in 1 calendar year resulting from the licensee’s possession 

and/or use of licensed materials.

Members of the public include persons who live, work, or may be near locations where 

fixed gauges are used or stored. Employees whose assigned duties do not include the 

use of licensed materials and who work in the vicinity where gauges are used or stored

are also considered members of the public.

The radiation dose in unrestricted areas does not exceed 0.02 mSv (2 mrem) in 

any 1 hour.

Typical unrestricted areas may include offices, shops, laboratories, a nearby walkway, 

areas outside buildings, and non-radioactive equipment storage areas. The licensee 

does not control access to these areas for purposes of controlling exposure to radiation 

or radioactive materials. However, the licensee may control access to these areas for 

other reasons, such as security.

Licensees must show compliance with both of these portions of the regulations. 

Calculations and measurements (e.g., using a dosimeter to monitor an area) are often 

used to prove compliance.

Calculational Method

The calculational method takes a tiered approach, using a three-part process starting with a 

worst case situation and moving toward more realistic situations. It makes the following 

simplifications: (1) each gauge is a point source, (2) typical radiation levels encountered when 

the source is in the shielded position are taken from either the SSD registration certificate or the 

manufacturer’s literature, and (3) no credit is taken for any shielding found between the gauges 

and the unrestricted areas.

Part 1 of the calculational method is simple but conservative. It assumes that an affected 

member of the public is present 24 hours a day and uses only the “inverse square law” to 

determine if the distance between the gauge and the affected member of the public is sufficient 

to show compliance with the public dose limits. Part 2 considers not only distance, but also the 

time that the affected member of the public is actually in the area under consideration. Part 3 

considers distance and the amount of time that both the gauge and the affected member of the 

public are present. Using this approach, licensees make only those calculations that are 

needed to demonstrate compliance. In many cases, licensees will need to use the calculational 

method through Part 1 or Part 2. The results of these calculations typically result in higher 

radiation levels than would exist at typical facilities but provide a method for estimating 

conservative doses that could be received.
G-4

Note: If calculated potential doses are not within regulatory limits initially, the licensee should 

take into account any shielding from the radiation beam provided by equipment or structures 

between the beam and workers or consider adding other shielding materials to reduce potential 

doses. In order to calculate the effects of any shielding, the licensee should seek the 

assistance of an individual knowledgeable in health physics.

Example 1

To better understand the calculational method, we will look at ABC Bottling, Inc., a fixed gauge 

licensee. Yesterday, the company’s president noted that three new gauges will be very close to 

a bottling control panel where a quality control supervisor—a worker who does not work with 

fixed gauges—works. The company’s president asked Joe, the RSO, to determine if the 

company is complying with the NRC’s regulations.

Joe measures the distances from each gauge to the bottling control panel and looks up in the 

manufacturer’s literature the radiation levels that individuals would encounter for each gauge. 

Figure G.1 is Joe’s sketch of the areas in question, and Table G.2 summarizes the information 

Joe has on each gauge.

Figure G.1 Diagram of bottling line and fixed gauges

This sketch shows the areas described in Examples 1 and 2.
G-5

Table G.2 Information Known about Each Gauge

DESCRIPTION OF KNOWN

INFORMATION

GAUGE 1 GAUGE 2 GAUGE 3

Where gauge is located Gauge on bottling 

line

Gauge on main feed 

line

Gauge on tank

Dose rate in mrem/hour encountered 

at specified distance from the gauge 

(from manufacturer’s literature)

2 mrem/hour at 1 ft 8 mrem/hour

at 1 ft

2 mrem/hour

at 3 ft

Distance in feet to bottling control 

panel

8 ft 12 ft 15 ft

Example 1: Part 1

Joe’s first thought is that the distance between the gauges and the bottling control panel may be 

sufficient to show compliance with the regulation in 10 CFR 20.1301. So, taking a “worst case” 

approach, he assumes that: (1) the gauges are constantly present (i.e., 24 hours per day), 

(2) all three gauges are on (i.e., shutters are open), and (3) a quality control (QC) supervisor—a 

worker who does not work with the fixed gauges—is constantly sitting at the control panel (i.e., 

24 hours per day). Joe proceeds to calculate the dose the QC supervisor might receive hourly 

and yearly from each gauge as shown in Tables G.3, G.4, and G.5 below.

Table G.3 Calculational Method, Part 1 –

Hourly and Annual Dose Received from Gauge 1

Step No. Description Gauge 1

Input Data Results

1 Dose received in an hour at known distance from gauge (e.g., from 

manufacturer’s data), in mrem per hour

2 2

2 Square of the distance (ft) at which the Step 1 rate was measured, 

in ft2 (1)2 1

3 Square of the distance (ft) from the gauge to the bottling control 

panel in an unrestricted area, in ft2

(8)2 64

4 Multiply the results of Step 1 by the results of Step 2 (this is an 

intermediate result)

2 x 1 =2

5 Divide the result of Step 4 by the result of Step 3 to calculate dose 

received by the worker at the bottling control panel = HOURLY 

DOSE RECEIVED FROM GAUGE 1, in mrem per hour

2/64 = 0.031

6 Multiply the result of Step 5 by 24 hours per day x 365 days per 

year = MAXIMUM ANNUAL DOSE RECEIVED FROM GAUGE 1, 

in mrem per year

0.031 x 24 x 365 = 

0.031 x 8760 = 272
G-6

Table G.4 Calculational Method, Part 1 –

Hourly and Annual Dose Received from Gauge 2

Gauge 2

Step No. Description Input Data Results

1 Dose received in an hour at known distance from gauge 

(e.g., from manufacturer’s data), in mrem per hour

8 8

2 Square of the distance (ft) at which the Step 1 rate was measured, 

in ft2

(1) 2 1

3 Square of the distance (ft) from the gauge to the bottling control 

panel in an unrestricted area, in ft2

(12) 2 144

4 Multiply the results of Step 1 by the results of Step 2 (this is an 

intermediate result)

8 x 1 = 8

5 Divide the result of Step 4 by the result of Step 3 to calculate dose 

received in an hour by the worker at the bottling control panel = 

HOURLY DOSE RECEIVED FROM GAUGE 2, in mrem per hour

8/144 = .056

6 Multiply the result of Step 5 by 24 hour per day x 365 days per 

year = MAXIMUM ANNUAL DOSE RECEIVED FROM GAUGE 2, 

in mrem per year

0.056 x 24 x 365 = 

0.056 x 8760 = 491

Table G.5 Calculational Method, Part 1 –

Hourly and Annual Dose Received from Gauge 3

Gauge 3

Step No. Description Input Data Results

1 Dose received in an hour at known distance from gauge 

(e.g., from manufacturer’s data), in mrem per hour

2 2

2 Square of the distance (ft) at which the Step 1 rate was measured, 

in ft2 (3)2 9

3 Square of the distance (ft) from the gauge to bottling control panel 

in an unrestricted area, in ft2 (15)2 225

4 Multiply the results of Step 1 by the results of Step 2 (this is an 

intermediate result)

2 x 9 =18

5 Divide the result of Step 4 by the result of Step 3 to calculate dose 

received by the worker at the bottling control panel = HOURLY 

DOSE RECEIVED FROM GAUGE 3, in mrem per hour

18/225 = 0.08

6 Multiply the result of Step 5 by 24 hour per day x 365 days per 

year = MAXIMUM ANNUAL DOSE RECEIVED FROM GAUGE 3, 

in mrem per year

0.08 x 24 x 365 = 0.08 

x 8760 = 701

To determine the total hourly and total annual dose received, Joe adds the pertinent data from 

the preceding tables (see Table G.6).

Table G.6 Calculational Method, Part 1 –

Total Hourly and Annual Dose Received from Gauges 1, 2, and 3

Step 

No.

Description Gauge 1 Gauge 2 Gauge 3 Sum

7 TOTAL HOURLY DOSE 

RECEIVED from Step 5 of Tables 

G.3, G.4, and G.5, in mrem per hour

0.031 0.056 0.08 0.031 + 0.056 + 0.08 = 

0.167

8 TOTAL ANNUAL DOSE RECEIVED 

from Step 6 of Tables G.3, G.4, and 

G.5, in mrem per year

272 491 701 272 + 491 + 701 

= 1,464
G-7

Note: The sum in Step 7 demonstrates compliance with the 2-mrem-in-any-one-hour limit. 

Reevaluate if assumptions change. If the sum in Step 8 exceeds 100 mrem per year, proceed 

to Part 2 of the calculational method.

At this point, Joe is pleased to see that the total dose that an individual could receive in any one 

hour is only 0.167 mrem, but he notes that an individual could receive a dose of 1,464 mrem in 

a year, much higher than the 100-mrem regulatory limit.

Example 1: Part 2

Joe reviews his assumptions and recognizes that the QC supervisor is not at the bottling control 

panel 24 hours per day. He decides to make a realistic estimate of the number of hours the QC 

supervisor would be present at the bottling control panel, keeping his other assumptions 

constant (i.e., the gauges are constantly present 24 hours per day and all three gauges remain 

on (i.e., shutter is open)). He then recalculates the annual dose received (see Table G.7).

Table G.7 Calculational Method, Part 2 –

Annual Dose Received from Gauges 1, 2, and 3

Step 

No.

Description Results

9 A. Average number of hours per day that individual spends in area of 

concern (e.g., worker present at bottling control panel 5 hours per day; 

the rest of the day the worker is away from the area performing other 

duties not in the vicinity of gauges)

B. Average number of days per week in area (e.g., worker is part time 

and works 3 days a week)

C. Average number of weeks per year in area (e.g., worker works all year)

5

3

52

10 Multiply the results of Step 9A by the results of Step 9B by the results of 

Step 9C = AVERAGE NUMBER OF HOURS IN AREA OF CONCERN PER 

YEAR

5 x 3 x 52 = 780

11 Multiply the sum in Step 7 by the results of Step 10 = ANNUAL DOSE 

RECEIVED FROM GAUGES CONSIDERING REALISTIC ESTIMATE OF 

TIME SPENT IN AREA OF CONCERN, in mrem per year

0.167 x 780 = 130

Note: If Step 11 exceeds 100 mrem in a year, proceed to Part 3 of the calculational method.

Although Joe is pleased to note that the calculated annual dose received is significantly lower, 

he realizes it still exceeds the 100 mrem in a year limit.

Example 1, Part 3

Again Joe reviews his assumptions and recognizes that Gauge 3 will only be used on the 

process line during product changeovers and Gauge 2 has different radiation levels depending 

on whether the gauge is in the on or off position (i.e., shutter is open or closed). As he 

examines the situation, he realizes he must consider each gauge individually.
G-8

Table G.8 Calculational Method, Part 3— Summary of Information

INFORMATION ON GAUGES:

• GAUGE 1: operates continuously (24 hours per day) on the bottling line.

• GAUGE 2: operates (in the “on” position) while the tank is being filled, approximately 1 hour 

during the time the worker is present. When the pipe is not filling the tank, the gauge is in the 

“off” position. While in the “off” position, the radiation level around the gauge drops to 2 mrem per 

hour at 1 foot, 1/4 of the radiation level when the gauge is in the “on” position.

• GAUGE 3: is only used on the process line during product changeovers, 4 weeks per year.

While affixed, it operates continuously (24 hours per day).

INFORMATION FROM EXAMPLE 1, PART 2, ON WHEN THE WORKER IS PRESENT AT THE 

BOTTLING CONTROL PANEL:

• 5 hours per day

• 3 days per week

• 52 weeks per year

Table G.9 Calculational Method, Part 3 –

Annual Dose Received from Gauges 1, 2, and 3

Step 

No.

Description Gauge 1 Gauge 2 “On” Gauge 2 “Off” Gauge 3

12 Average number of hours per 

day gauge operates when 

worker is present at the bottling 

control panel

5 1 4 5

13 Average number of days per 

week gauge operates when 

worker is present at the bottling 

control panel

3 3 3 3

14 Average number of weeks per 

year gauge operates when 

worker is present at the bottling 

control panel

52 52 52 4

15 Multiply the results of Step 12 

by the results of Step 13 by the 

results of Step 14 = TOTAL 

HOURS EACH GAUGE 

OPERATED PER YEAR 

WHILE WORKER IS 

PRESENT AT BOTTLING 

CONTROL PANEL

5 x 3 x 52 =

780

1 x 3 x 52 =

156

4 x 3 x 52 =

624

5 x 3 x 4 = 60

16 Multiply the results of Step 15 

by the results of Step 7 (for 

Gauge 2 in the “off” position, 

the radiation level drops to 

1/4th, so divide the results of 

Step 7 by 4) = ANNUAL DOSE 

RECEIVED FROM EACH 

GAUGE, in mrem in a year

780 x 0.031 = 

24

156 x 0.056 = 

8.7

624 x (0.056/4) 

= 8.7

60 x 0.08 = 4.8 

in mrem in a 

year 
G-9

Step 

No.

Description Gauge 1 Gauge 2 “On” Gauge 2 “Off” Gauge 3

17 Sum the results of Step 16 for 

each gauge = TOTAL ANNUAL 

DOSE RECEIVED 

CONSIDERING REALISTIC 

ESTIMATE OF TIME SPENT 

IN AREA OF CONCERN AND 

TIME GAUGE OPERATES, in 

mrem per year

24 + 8.7 + 8.7 + 4.8 = 46.2

Note: If the result in Step 17 is greater than 100 mrem per year, the licensee must take 

corrective actions.

Joe is pleased that the result in Step 17 shows compliance with the 100-mrem-per-year limit. 

Had the result in Step 17 been higher than 100 mrem per year, then Joe could have done one 

or more of the following:

• Consider whether the assumptions used to determine occupancy and the time each 

gauge operates are accurate, revise the assumptions as needed, and recalculate using 

the new assumptions.

• Calculate the effect of any shielding located between the gauges and the bottling control 

panel (such calculation is beyond the scope of this appendix).

• Take corrective action (e.g., add shielding, move the bottling control panel) and perform 

new calculations to demonstrate compliance.

• Designate the area outside the storage area as a restricted area and the QC supervisor 

as an occupationally exposed individual. This would require controlling access to the 

area for purposes of radiation protection and training of the QC supervisor as required 

by 10 CFR 19.12, “Instruction to workers.”

Note that in the example, Joe evaluated the unrestricted area at the bottling control panel. 

Licensees also need to make similar evaluations for other unrestricted areas and to keep in 

mind the “as low as is reasonably achievable” (ALARA) principle, taking reasonable steps to 

keep radiation dose received below regulatory limits. In addition, licensees need to be alert to 

changes in situations (e.g., adding a gauge to the process line, changing the QC supervisor’s 

schedule, or changing the estimate of the portion of time spent at the bottling control panel) and 

to perform additional evaluations, as needed.

RECORDKEEPING: 10 CFR 20.2107 requires licensees to maintain records demonstrating 

compliance with the dose limits for individual members of the public.

Combination Measurement – Calculational Method

This method, which allows the licensee to take credit for shielding between the gauge and the 

area in question, begins by measuring radiation levels in the areas, as opposed to using 

manufacturer-supplied rates at a specified distance from each gauge. These measurements 
G-10

must be made with calibrated survey meters sufficiently sensitive to measure background levels 

of radiation. A maximum dose of 1 mSv (100 mrem) received by an individual over a period of 

2,080 hours (i.e., a work year of 40 hours per week for 52 weeks per year) is equal to less than 

0.5 microsievert (0.05 mrem) per hour.

This rate is well below the minimum sensitivity of most commonly available survey instruments 

(i.e., Geiger-Mueller or G-M). Instruments used to make measurements for calculations must 

be sufficiently sensitive. An instrument equipped with a scintillation-type detector (e.g., a 

micro-Roentgen or micro-R meter) should be adequate.

Licensees may also choose to use dosimeters specifically engineered for measuring 

environmental dose rates. Dosimeters used for personnel monitoring may not have sufficient 

sensitivity for this purpose. Generally, the minimum reportable dose received by a personnel 

dosimeter is 0.1 mSv (10 mrem). Suppose a personnel dosimeter monitors dose received and 

is changed once a month. If the measurements are at the minimum reportable level, the annual 

dose received could have been about 1.2 mSv (120 mrem), a value in excess of the 100-

millirem-per-year limit. If licensees use dosimeters to evaluate compliance with the public dose 

limits, they should consult with their dosimetry supplier and choose more sensitive dosimeters.

This direct measurement method would provide a definitive measurement of actual radiation 

levels in unrestricted areas without any restrictive assumptions. Records of these 

measurements can then be evaluated to ensure that rates in unrestricted areas do not exceed 

the 100-mrem-per-year limit.

Example 2

As in Example 1, Joe is the RSO for ABC Bottling, Inc., a fixed gauge licensee. The company 

has three gauges located near a bottling control panel that is operated by a worker who does 

not work with the fixed gauges. See Figure G.1 and Table G.2 for information. Joe wants to 

see if the company complies with the public dose limits at the bottling control panel. Joe placed 

an environmental dosimeter at the bottling control panel for 30 days. The dosimetry processor 

sent Joe a report indicating the environmental dosimeter received 100 mrem in those 30 days.

Table G.10 Combination Measurement – Calculational Method

Step 

No.

Description Input Data and Results

Part 1

1 Dose received by dosimeter, in mrem 100

2 Total hours that dosimeter was exposed 24 hours per day x 30 days = 720

3 Divide the results of Step 1 by the results of Step 2 to 

determine HOURLY DOSE RECEIVED, in mrem per hour

0.14

4 Multiply the results of Step 3 by 8,760 hours in one year = 

MAXIMUM ANNUAL DOSE RECEIVED FROM GAUGES, 

in mrem per year

0.14 x 8,760 = 1,226

Note: For the conditions described above, Step 3 indicates that the dose received in any one

hour is less than the 2-mrem-in-any-one-hour limit. However, if there are any changes, then the 

licensee would need to reevaluate the potential doses that could be received in any one hour. 
G-11

Step 4 indicates that the annual dose received would be much greater than the 100 mrem per 

year allowed by the regulations.

Part 2

At this point Joe can adjust for a realistic estimate of the time the worker spends at the bottling control 

panel, as he did in Part 2 of Example 1.

Part 3

If the results of Joe’s evaluation in Part 2 show that the annual dose received in a year exceeds 

100 mrem, then he can make adjustments for realistic estimates of the time of exposure in the area of 

concern, as in Part 3 of Example 1. Recall that the dosimeter measurement was made while all the 

gauges were operating (i.e., 24 hours a day for the 30 days that the dosimeter was in place).

APPENDIX H

OPERATING AND EMERGENCY PROCEDURES

H-1

Operating Procedures

• If personnel dosimetry is provided, do the following:

– Always wear your assigned dosimeter when using the gauge.

– Never wear another person’s dosimeter.

– Never store your dosimeter near a gauge.

• Use the gauge according to the manufacturer’s or distributor’s instructions and 

recommendations. Perform routine cleaning and maintenance according to the 

manufacturer’s or distributor’s instructions and recommendations.

• Test each gauge for the proper operation of the on-off mechanism (shutter) and 

indicator, if any, at intervals not to exceed 6 months or as specified in the SSD 

registration certificate.

• Do not touch the unshielded source with your fingers, hands, or any part of your body.

• Do not place hands, fingers, feet, or other body parts in the radiation field from an 

unshielded source.

• Post a radiation warning sign at each entryway to an area where it is possible to be 

exposed to the radiation beam.

• Prevent employees from entering the radiation beam during maintenance, repairs, or 

work in, on, or around the equipment on which the device is mounted by developing 

lock-out procedures. These procedures should specify who will be responsible for 

ensuring that the lock-out procedures are followed.

• Prevent unauthorized access, removal, or use of the gauge.

• After making changes affecting the gauge (e.g., changing the location of gauges, 

removing shielding, adding gauges, changing the occupancy of adjacent areas, etc.) 

reevaluate compliance with public dose limits and ensure proper security of gauges.

• Conduct a physical inventory every 6 months to account for all devices.

• Provide instructions to prevent unauthorized access, removal, or use of fixed gauges, 

including at temporary job sites if authorized on the license.

• Require timely reporting to the NRC pursuant to 10 CFR 20.2201 and 10 CFR 20.2207, 

as appropriate.

• Address guidelines to meet the security requirements of 10 CFR 20.1801, 10 CFR 

20.1802, and any other applicable security requirements.
H-2

Emergency Procedures

• If a gauge becomes damaged, a source becomes dislodged, a source is leaking, the 

device fails to function properly, or if any other non-radiological emergency (e.g., fire or 

medical emergency) or unusual situation arises:

– Stop use of the gauge.

– Immediately secure the area and keep people away from the gauge until the 

situation is assessed and radiation levels are known. Perform first aid for any 

injured individuals and remove them from the area when it is medically safe to do 

so.

– If any equipment is involved, isolate the equipment until it is determined there is 

no contamination present.

– Gauge users and other potentially contaminated individuals should not leave the 

scene until emergency assistance arrives.

– Notify the persons, in the order listed below, of the situation:

NAME1 WORK PHONE NUMBER1 HOME PHONE NUMBER1

____________________ _______________________ ________________________

____________________ _______________________ ________________________

____________________ _______________________ ________________________

• Follow the directions provided by the person contacted above.

Radiation Safety Officer and Licensee Management

• Arrange for a radiation survey to be conducted as soon as possible by a knowledgeable 

person using appropriate radiation detection instrumentation. This person could be a 

licensee employee using a survey meter, a local emergency responder, or a consultant. 

To accurately assess the radiation danger, it is essential that the person performing the 

survey be competent in the use of the survey meter.

 1 Fill in with (and update, as needed) the names and telephone numbers of appropriate personnel (e.g., the RSO,

 AUs, or other knowledgeable licensee staff, licensee’s consultant, gauge manufacturer, distributor, or

 representative, fire department, or other emergency response organization, as appropriate, and the NRC) to be

 contacted in case of emergency.
H-3

• Make necessary notifications to local authorities as well as the NRC as required. 

Appendix L contains typical NRC incident notifications required for fixed gauge 

licensees.

(Even if not required to do so, you may report ANY incident to the NRC by calling the 

NRC’s Operations Center at (301) 816-5100, which is staffed 24 hours a day and 

accepts collect calls.) NRC notification is required when gauges containing licensed 

material are lost or stolen and when gauges are damaged or involved in incidents that 

result in doses in excess of 10 CFR 20.2203 limits. Reporting requirements are found in 

10 CFR 20.2201-2203 and in 10 CFR 30.50.

APPENDIX I

MODEL LEAK TEST PROGRAM

I-1

Training

Before allowing an individual to perform leak testing, the RSO will ensure that he or she has 

sufficient classroom and on-the-job training to show competency in performing leak tests 

independently.

Classroom training may be in the form of lecture, video, computer-based, or self-study and will 

cover the following subject areas:

• principles and practices of radiation protection;

• radioactivity measurements, monitoring techniques, and the use of instruments;

• mathematics and calculations used for measuring radioactivity; and

• biological effects of radiation.

Appropriate on-the-job training consists of:

• observing authorized personnel collecting and analyzing leak test samples; and

• collecting and analyzing leak test samples under the supervision and in the physical 

presence of an individual authorized to perform leak tests.

Facilities and Equipment

• To ensure achieving the required sensitivity of measurements, analyze leak tests in a 

low-background area.

• Use a calibrated and operable survey instrument to check leak test samples for gross 

contamination before they are analyzed.

• Analyze the leak test sample using an instrument that is appropriate for the type of 

radiation to be measured (e.g., a well-counter system for gamma emitters, liquid 

scintillation for beta emitters, and a gas-flow proportional counter for alpha emitters).

• If the sensitivity of the counting system is unknown, the minimum detectable activity 

(MDA) should be determined. The MDA may be determined using the following formula:

MDA = 

 t x E

2.71 + 4.65 √(bkg x t)

where: MDA = minimum detectable activity in disintegrations per minute (dpm)

bkg = background count rate in counts per minute (cpm)

t = background counting time in minutes

E = detector efficiency in counts per disintegration
I-2

Example

Assume: bkg = 200 counts per minute (cpm)

E = 0.1 counts per disintegration (10 percent efficient)

T = 2 minutes

MDA = 2.71 + 4.65 √(200 cpm x 2 minutes) =

 2 x 0.1 0.2

2.71 + 4.65√(400)

= 2.71 + 4.65 (20) = 2.71 + 93 =

 0.2 0.2 0.2

95.71

=

 minute

478.55 disintegrations

becquerels (Bq) =

 second

1 disintegration

MDA = 478.55 disintegration x 1 minute 

 minutes 60 seconds

= 7.976 Bq

Frequency for Conducting Leak Tests of Sealed Sources

Leak tests will be conducted at the frequency specified in the respective Sealed Source and 

Device (SSD) registration certificate.

Procedure for Performing Leak Testing and Analysis

• For each source to be tested, list identifying information such as the manufacturer’s 

name, model number, serial number, radionuclide, and activity.

• If available, use a survey meter to monitor exposure.

• Prepare a separate wipe sample (e.g., cotton swab or filter paper) for each source.

• Number each wipe to correlate with identifying information for each source.

• Wipe the most accessible area where contamination would accumulate if the sealed 

source were leaking (see manufacturer’s instructions).

• Select an instrument that is sensitive enough to detect 185 bequerel (Bq) (0.005 

microcurie) of the radionuclide contained in the gauge.

• Count and record background count rate using the selected instrument.

• Check the instrument’s counting efficiency using a standard source of the same 

radionuclide as the source being tested or one with similar energy characteristics. 

Accuracy of standards should be within ± 5 percent of the stated value and traceable to 

a primary radiation standard such as those maintained by the NIST.
I-3

• Calculate the counting efficiency of the detector. The counting efficiency may be 

determined using the following formula:

Efficiency in cpm/Bq =

 activity of standard source in Bq

[(cpm from std) - (cpm from bkg)]

where: cpm = counts per minute

bkg = background

std = standard

Bq = becquerel

• Count each wipe sample; determine net count rate.

• For each sample, calculate and record estimated activity in becquerels (or microcuries). 

The activity of the sample in becquerels may be calculated using the following formula:

Activity of sample =

 efficiency in cpm/Bq

[(cpm from wipe sample) - (cpm from bkg)]

• Sign and date the list of sources, data, and calculations. Retain records for 3 years.

• If the wipe test activity is 185 becquerels (0.005 microcuries) or greater, notify the RSO 

so that the source can be withdrawn from use and disposed of properly. Also notify the 

NRC.

APPENDIX J

INFORMATION NEEDED TO SUPPORT APPLICANT’S REQUEST TO 

PERFORM NONROUTINE OPERATIONS

J-1

Applicants should review Section 8.10.8, “Maintenance,” which discusses, in general, licensee 

responsibilities before any maintenance or repair is performed.

Nonroutine operations include installation, initial radiation survey, repair and maintenance of 

radiological safety components, relocation, replacement, alignment, removal from service, and 

disposal of sealed sources. See Figure 8.6 in Section 8.10.8.

Any replacement components, parts, or other materials (e.g., lubricants) other than those 

supplied, specified, or recommended by the manufacturer or distributor need to be evaluated to 

ensure that they do not degrade the engineering safety analysis performed and accepted as 

part of the device’s SSD registration certificate. Licensees also need to ensure that, after 

maintenance or repair is completed, the gauge is tested and functions as designed before the 

unit is returned to routine use.

If nonroutine operations are not performed properly with attention to good radiation safety 

principles, the gauge may not operate as designed, and personnel performing these tasks could 

receive radiation doses that exceed the NRC’s regulatory limits. Radionuclides and activities in 

fixed gauges vary widely. For illustrative purposes, in less than one minute, an unshielded 

cesium-137 source with an activity of 3.7 gigabequerels (100 millicuries) can deliver 0.05 Sv 

(5 rem) to a worker’s hands or fingers (i.e., extremities), assuming the extremities are 

1 centimeter from the source. This dose corresponds to the threshold for extremity monitoring. 

Some gauges may contain sources of even higher activities with correspondingly higher dose 

rates.

Thus, applicants wishing to perform nonroutine operations must use personnel with special 

training and follow appropriate procedures consistent with the manufacturer’s or distributor’s 

instructions and recommendations that address radiation safety concerns (e.g., use of radiation 

survey meter, shielded container for the source, and personnel dosimetry (if required)).

Accordingly, applicants wishing to perform nonroutine operations must provide the following 

information with their license application:

• Describe the types of work, maintenance, cleaning, and/or repair that involve any of the 

following:

– installation, relocation, or alignment of the gauge;

– components, including electronics, related to the radiological safety of the gauge 

(e.g., the source, source holder, source drive mechanism, shutter, shutter 

control, or shielding);

– replacement and disposal of sealed sources;

– removal of a gauge from service;

– a potential for any portion of the body to come into contact with the primary 

radiation beam; or

– any other activity during which personnel could receive radiation doses

exceeding NRC limits.
J-2

• Identify who will perform nonroutine operations and describe their training and 

experience. Acceptable training includes manufacturers’ or distributors’ courses for 

nonroutine operations or an equivalent.

• Submit procedures for nonroutine operations. These procedures should ensure the 

following:

– Doses to personnel and members of the public are within regulatory limits and 

are kept as low as is reasonably achievable (ALARA) (e.g., use of shielded 

containers or shielding).

– The source is secured against unauthorized removal or access or is under 

constant surveillance.

– Appropriate labels and signs are used.

– Lock-out procedures are adequate to ensure that no individual or portion of an 

individual’s body can enter the radiation beam.

– Manufacturer’s or distributor’s instructions and recommendations are followed.

– Replacement components, parts, or other materials (e.g., lubricants) other than 

those supplied, specified, or recommended by the manufacturer or distributor are 

evaluated to ensure that they do not degrade the engineering safety analysis 

performed and accepted as part of the SSD registration certificate.

– The gauge, before being returned to routine use, is tested to verify that it 

functions as designed and source integrity is not compromised.

• Confirm that individuals performing nonroutine operations on gauges will wear both 

whole body and extremity monitoring devices or perform a prospective evaluation 

demonstrating that unmonitored individuals performing nonroutine operations are not 

likely to receive, in one year, a radiation dose in excess of 10 percent of the NRC’s 

regulatory limits.

• Confirm possession of at least one survey instrument that is appropriate for measuring 

the types of radiation and expected dose rates from the fixed gauge(s).

• Describe steps to be taken to ensure that radiation levels in areas where nonroutine 

operations will take place do not exceed limits set in 10 CFR 20.1301(e.g., surveys, 

calculations, etc.).
APPENDIX K

APPLICABLE U.S. DEPARTMENT OF TRANSPORTATION 

REGULATIONS

K-1

Note: The reference guides included at the end of this appendix are for reference only and are 

not a substitute for transportation regulations.

In 10 CFR 71.5, “Transportation of licensed material,” the U.S. Nuclear Regulatory Commission 

(NRC) requires compliance with U.S. Department of Transportation (DOT) regulations in 49 

CFR, “Transportation.” The major areas in the DOT regulations that are most relevant for 

transportation of typical fixed gauges shipped as Type A quantities are as follows:

• 49 CFR Part 172, “Hazardous Materials Table, Special Provisions, Hazardous Materials 

Communications, Emergency Response Information, Training Requirements, and 

Security Plans”

(1) Table of Hazardous Materials and Special Provisions (Subpart B)

– Purpose and use of hazardous materials table (49 CFR 172.101)

– List of Hazardous Substances and Reportable Quantities (49 CFR 

172.101, Appendix A, Table 2)

(2) Shipping Papers (Subpart C)

– Preparation and retention of shipping papers (49 CFR 172.201)

– Description of hazardous material on shipping papers (49 CFR 172.202)

– Additional description requirements (49 CFR 172.203)

– Shipper’s certification (49 CFR 172.204)

(3) Marking (Subpart D)

– General marking requirements for non-bulk packagings (49 CFR 172.301)

– Prohibited marking (49 CFR 172.303)

– Marking requirements (49 CFR 172.304)

– Class 7 (radioactive) materials (49 CFR 172.310)

– Hazardous substances in non-bulk packagings (49 CFR 172.324)

(4) Labeling (Subpart E)

– General labeling requirements (49 CFR 172.400)

– Prohibited labeling (49 CFR 172.401)

– Class 7 (radioactive) material (49 CFR 172.403)

– Placement of labels (49 CFR 172.406)

– Label specifications (49 CFR 172.407)

– RADIOACTIVE WHITE-I label (49 CFR 172.436)

– RADIOACTIVE YELLOW-II label (49 CFR 172.438)

– RADIOACTIVE YELLOW-III label (49 CFR 172.440)

(5) Placarding (Subpart F)

– Applicability of placarding requirements (49 CFR 172.500)

– Prohibited and permissive placarding (49 CFR 172.502)

– General placarding requirements (49 CFR 172.504)

– Providing and affixing placards: highway (49 CFR 172.506)

– Visibility and display of placards (49 CFR 172.516)

– General specifications for placards (49 CFR 172.519)

– RADIOACTIVE placard (49 CFR 172.556)

(6) Emergency Response Information (Subpart G)

– Applicability and general requirements (49 CFR 172.600)
K-2

– Emergency response information (49 CFR 172.602)

– Emergency response telephone number (49 CFR 172.604)

(7) Training (Subpart H)

− Applicability and responsibility for training and testing (49 CFR 172.702)

− Training requirements (49 CFR 172.704)

• 49 CFR Part 173, “Shippers – General Requirements for Shipments and Packagings,” 

Class 7 (Radioactive) Materials (Subpart I)

– Definitions (49 CFR 173.403)

– General design requirements (49 CFR 173.410)

– Additional design requirements for Type A packages (49 CFR 173.412)

– Authorized Type A packages (49 CFR 173.415)

– Requirements for determining basic radionuclide values, and for the listing of 

radionuclides on shipping papers and labels (49 CFR 173.433)

– Table of A1 and A2 values for radionuclides (49 CFR 173.435)

– Radiation level limitations and exclusive use provisions (49 CFR 173.441)

– Quality control requirements prior to each shipment of Class 7 (radioactive) 

materials (49 CFR 173.475)

– Approval of special form Class 7 (radioactive) materials (49 CFR 173.476)

• 49 CFR Part 177, “Carriage by Public Highway”

(1) General Information and Regulations (Subpart A)

– Driver training (49 CFR 177.816)

– Shipping papers (49 CFR 177.817)

(2) Loading and Unloading (Subpart B)

– General requirements (packages secured in a motor vehicle against 

movement) (49 CFR 177.834(a))

– Class 7 (radioactive) material (49 CFR 177.842)
K-3
K-4
K-5
K-6
K-7
K-8
K-9
K-10
APPENDIX L

NRC INCIDENT NOTIFICATIONS

L-1

Table L.1 Typical NRC Incident Notifications Required for Fixed Gauge Licensees

Event Telephone 

Notification

Written 

Report

Regulatory Requirement

Theft or loss of material Immediate 30 days 10 CFR 20.2201(a)(1)(i)

Whole body dose of 0.25 Sv (25 rem) or 

more

Immediate 30 days 10 CFR 20.2202(a)(1)(i)

Extremity dose of 2.5 Gy (250 rads) or 

more

Immediate 30 days 10 CFR 20.2202(a)(1)(iii)

Whole body dose greater than 0.05 Sv 

(5 rem) in 24 hours

24 hours 30 days 10 CFR 20.2202(b)(1)(i)

Extremity dose greater than 0.5 Sv 

(50 rem) in 24 hours

24 hours 30 days 10 CFR 20.2202(b)(1)(iii)

Whole body dose greater than 0.05 Sv 

(5 rem)

None 30 days 10 CFR 20.2203(a)(2)(i)

Dose to individual member of public 

greater than 1 mSv (100 mrem)

None 30 days 10 CFR 20.2203(a)(2)(iv)

Defect in equipment that could create a 

substantial safety hazard

2 days 30 days 10 CFR 21.21(d)(3)(i)

Filing petition for bankruptcy under any 

chapter of title11 U.S.C.

None Immediately

after filing 

petition

10 CFR 30.34(h)

Expiration of license None 60 days 10 CFR 30.36(d)

Decision to permanently cease licensed 

activities at entire site

None 60 days 10 CFR 30.36(d)

Decision to permanently cease licensed 

activities in any separate building or 

outdoor area that is unsuitable for 

release for unrestricted use

None 60 days 10 CFR 30.36(d)

No principal activities conducted for 

24 months at the entire site

None 60 days 10 CFR 30.36(d)

No principal activities conducted for 

24 months in any separate building or 

outdoor area that is unsuitable for 

release for unrestricted use

None 60 days 10 CFR 30.36(d)

Event that prevents immediate protective 

actions necessary to avoid exposures to 

radioactive materials that could exceed 

regulatory limits

Immediate 30 days 10 CFR 30.50(a)

Equipment is disabled or fails to function 

as designed when required to prevent 

radiation exposure in excess of 

regulatory limits 

24 hours 30 days 10 CFR 30.50(b)(2)

Unplanned fire or explosion that affects 

the integrity of any licensed material or 

device, container, or equipment with 

licensed material

24 hours 30 days 10 CFR 30.50(b)(4)

Note: Telephone notifications shall be made to the NRC Operations Center at (301) 816-5100.

APPENDIX M

INTERIM STAFF GUIDANCE ON CONSTRUCTION

M-1

INTERIM STAFF GUIDANCE TO NUREG-1556 AND NUREG-1520: 

COMMENCEMENT OF CONSTRUCTION AT EXISTING AND PROPOSED SOURCE, 

BYPRODUCT, AND SPECIAL NUCLEAR MATERIAL FACILITIES AND IRRADIATORS WITH 

SIGNIFICANT ENVIRONMENTAL IMPACTS

PURPOSE AND SCOPE

This Interim Staff Guidance (ISG) provides guidance to U.S. Nuclear Regulatory Commission 

(NRC) staff on the new definition of construction and the consideration of activities that can be 

performed by materials license applicants and potential applicants (hereinafter collectively 

referred to as “applicants”), and licensees before the NRC staff has concluded its environmental 

review of the proposed licensing action.

This ISG applies to the review of licensing actions related to the receipt and possession of 

licensable source, byproduct, and special nuclear material (SNM) for the conduct of any activity 

which the NRC determines will significantly affect the quality of the environment. This ISG is 

intended to provide guidance to NRC staff but may also be instructive to all holders of operating 

licenses for source, byproduct, and SNM facilities and irradiators, and all persons that have 

submitted applications to construct source, byproduct, and SNM facilities or irradiators, or have 

submitted letters of intent to submit such applications under Title 10 of the Code of Federal 

Regulations (10 CFR) Parts 30, 36, 40, and 70.

This ISG applies to all Part 30, 36, 40 and 70 materials facilities other than uranium recovery 

facilities. Site preparation activities at uranium recovery facilities are addressed in Regulatory 

Issue Summary 2009-12, Uranium Recovery Policy Regarding Site Preparation Activities at 

Proposed, Unlicensed Uranium Recovery Facilities, September 23, 2009, ML092090353. 

If a licensing action initiated pursuant to 10 CFR Parts 30, 40, or 70 meets any of the criteria in 

10 CFR 51.20 or 51.21, then commencement of construction of a facility before the NRC staff 

has completed its environmental review process is grounds for denial of the license application, 

in accordance with 10 CFR 30.33(a)(5), 40.32(e), and 70.23(a)(7). However, if the licensing 

action meets the criteria in 10 CFR 51.22(c) for a categorical exclusion, and the NRC has not 

determined that an environmental assessment or an environmental impact statement is required

in accordance with 10 CFR 51.22(b), then commencement of construction before the NRC staff 

concludes the environmental process should not be the sole basis for denial of the license 

application, as the NRC has already determined that this category of actions does not have a 

significant impact on the environment. In accordance with 10 CFR 36.15, commencement of 

construction of an irradiator will only be grounds for denial if the licensee or applicant has not 

submitted both an application and the requisite licensing fee. 

BACKGROUND

The NRC amended its regulations in September 2011, by revising certain provisions applicable 

to the licensing and approval processes for byproduct, source and SNMs licenses, and 

irradiators in the final rule, “Licenses, Certifications, and Approvals for Materials Licensees” (76 

FR 56951; September 15, 2011) (Material Licenses Construction Rule). The revisions 

contained in the Material Licenses Construction Rule revised the definitions of “construction” 

and “commencement of construction” with respect to materials licensing actions conducted 

under the NRC's regulations. The NRC adopted these changes to further improve the 

effectiveness and efficiency of the licensing and approval processes for future materials license 
M-2

applications, as well as to eliminate certain inconsistencies that existed within the NRC's 

regulations with respect to the use and definition of the terms “construction” or “commencement 

of construction” for certain materials licensees for purposes of its environmental reviews.

The new definitions of “commencement of construction” in 10 CFR 30.4, 36.2, 40.4, and 70.4 

are identical. 

Commencement of construction means taking any action defined as “construction” or any other 

activity at the site of a facility subject to the regulations in this part that has a reasonable nexus 

to:

1. Radiological health and safety; or

2. Common defense and security.

In 10 CFR 150.31, commencement of construction means taking any action defined as 

“construction” or any other activity at the site of a facility subject to the regulations in this part 

that has a reasonable nexus to radiological health and safety. The regulations in 10 CFR 

150.31 address the requirement for Agreement State regulation of byproduct material. Although 

Agreement State licensees may find this ISG informative, they should also communicate with 

the pertinent Agreement State agency for that agency’s applicable requirements and guidance.

The new definitions of “construction” in 10 CFR 30.4, 36.2, and 70.4 are also identical. 

Construction means the installation of foundations, or in-place assembly, erection, fabrication, 

or testing for any structure, system, or component of a facility or activity subject to the

regulations in this part that are related to radiological safety or security. The term “construction” 

does not include: 

(1) Changes for temporary use of the land for public recreational purposes;

(2) Site exploration, including necessary borings to determine foundation conditions or

other preconstruction monitoring to establish background information related to the

suitability of the site, the environmental impacts of construction or operation, or the

protection of environmental values;

(3) Preparation of the site for construction of the facility, including clearing of the site,

grading, installation of drainage, erosion and other environmental mitigation measures,

and construction of temporary roads and borrow areas;

(4) Erection of fences and other access control measures that are not related to the safe

use of, or security of, radiological materials subject to this part;

(5) Excavation;

(6) Erection of support buildings (e.g., construction equipment storage sheds, warehouse

and shop facilities, utilities, concrete mixing plants, docking and unloading facilities,

and office buildings) for use in connection with the construction of the facility;

(7) Building of service facilities (e.g., paved roads, parking lots, railroad spurs, exterior

utility and lighting systems, potable water systems, sanitary sewerage treatment

facilities, and transmission lines);

(8) Procurement or fabrication of components or portions of the proposed facility occurring

at other than the final, in-place location at the facility; or

(9) Taking any other action that has no reasonable nexus to:

(i) Radiological health and safety, or

(ii) Common defense and security.
M-3

“Construction,” as defined in 10 CFR 40.4, also includes the installation of wells associated with 

radiological operations (e.g., production, injection, or monitoring well networks associated with 

in-situ recovery or other facilities). 

The Atomic Energy Act of 1954, as amended, expressly limits the NRC’s regulatory authority to 

matters concerning the radiological public health and safety or common defense and security 

and non-radiological hazards to the extent such hazards result from the actual processing of byproduct material. The NRC has determined that this authority does not extend to site 

preparation activities that do not have a nexus to radiological health and safety or common 

defense and security.

This guidance provides criteria for NRC staff to use in evaluating whether a particular 

construction activity has a nexus to radiological health and safety, and thus falls under the 

jurisdiction of the NRC for licensing purposes. An activity or action has a reasonable nexus to 

radiological health and safety or the common defense and security if that activity or action has a 

rational, direct link to ensuring that a materials facility is operating, or will operate, in accordance 

with the NRC’s regulations and in a manner that protects the public health and safety or the 

common defense and security from radiological hazards. The revised definition of construction 

in 10 CFR 30.4, 36.2, 40.4, 70.4, and 150.31 list activities that are not considered “construction.”

This guidance provides examples of activities that fall under each of the excepted activities that 

do not constitute construction. This guidance addresses some important considerations for 

materials licensees and applicants that were emphasized in the response to comments on the 

proposed Material Licenses Construction Rule. For example, site preparation activities that are 

not considered “construction,” while not under NRC jurisdiction may be subject to the regulatory 

authority of another Federal, State, or local agency which may require National Environmental 

Policy Act or state environmental review. NRC’s responsibilities under the National Historic 

Preservation Act of 1966, as amended (NHPA), must also be satisfied before a license is 

issued. Specifically, as noted in the SOC to the final Material Licenses Construction Rule, 

under certain circumstances the NRC may be required to deny a license application if the NRC 

determines that the applicant intentionally significantly adversely affected, or allowed to be 

affected, a historic property with intent to avoid the requirements of §106 of the NHPA. 

DISCUSSION OF EXAMPLES

In addition to the background discussion provided above, the following examples clarify the

delineation of site preparation activities and construction activities. It is important to recognize 

that the NRC may have regulatory authority over activities that can occur before construction 

begins, such as procurement of basic components as defined in 10 CFR Part 21, the process 

of dedicating commercial grade items or basic components, or procurement of items relied on 

for safety (IROFS) as defined in 10 CFR Part 70. It should also be noted that, while site 

preparation activities may not require prior NRC approval, various local, State, or other Federal 

permits may be required.

BYPRODUCT MATERIAL (10 CFR PART 30)

Prior to the conclusion of the environmental review process, applicants for byproduct material 

licenses or license amendments should not perform construction activities that have a nexus to 

radiological health and safety or the common defense and security. An activity or action has a 

reasonable nexus to radiological health and safety or the common defense and security if that 

activity or action has a rational, direct link to ensuring that a licensed materials facility is 

operating, or will operate, in accordance with the NRC’s regulations and in a manner that 
M-4

protects the public health and safety or the common defense and security from radiological 

hazards.

Installation of foundations or in-place assembly, erection, fabrication, or testing for any structure, 

system, or component of a facility or activity subject to 10 CFR Part 30 that are related to 

radiological health and safety or common defense and security should not be performed prior to 

the conclusion of the environmental review of a license application or amendment. Byproduct 

material license applicants subject to 10 CFR Part 30 may perform those site preparation

activities identified in revised 10 CFR 30.4 before the NRC has completed its environmental 

review of the license application.

Excavation and other site preparation activities that do not have a reasonable nexus to 

radiological public health and safety or common defense and security, whether permanent or 

temporary, are not “construction” activities. For example, piles driven to support the erection of 

a bridge for a temporary or permanent access road to a new facility would not be considered as 

construction and may be performed prior to the NRC staff concluding its environmental review 

of a proposed action.

The installation of a temporary feature within an excavation for a building in which materials 

license activities will be conducted and that will be removed during construction is a site 

preparation activity. Such features include retaining walls, dewatering systems, ramps, and 

other structures that will have no physical presence following construction.

Construction includes installation of the foundation, including soil compaction; the installation of 

permanent drainage systems and geofabric; the placement of backfill, concrete (e.g., mudmats), 

or other materials that will not be removed before placement of the foundation of a structure; the 

placement and compaction of a subbase; the installation of reinforcing bars to be incorporated 

into the foundation of the structure; the erection of concrete forms for the foundations that will 

remain in place permanently (even if nonstructural); and the placement of concrete or other 

material constituting the foundation of any safety-related feature.

The term “permanent” in this context includes anything that will exist in its final, in-place facility

location after commencement of operations with licensed material. Construction also includes 

the “onsite, in-place” fabrication, erection, integration, or testing activities for any in-scope 

safety-related equipment. The terms “onsite, in place, fabrication, erection, integration, or 

testing” describe the process of constructing a facility in its final, onsite plant location, where 

components or modules are integrated into the final, in-plant location. The fabrication, 

assembly, and testing of components and modules in a shop building, warehouse, or laydown 

area, even if located onsite, is not construction. However, the installation or integration of the 

safety-related equipment into its final plant location is construction.

Construction also includes driving piles for safety-related equipment. Hence, an applicant must 

obtain a license before driving piles for safety-related equipment. However, driving piles that do 

not ensure the structural stability or integrity of a safety-related structure (e.g., piles driven to 

support the erection of a bridge for a temporary or permanent access road) is not construction; 

therefore, those piles may be driven prior to the NRC staff concluding its environmental review 

of a proposed action.
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IRRADIATORS (10 CFR 36)

An applicant for a new irradiator license under 10 CFR Part 36 may perform the nonconstruction activities identified in revised 10 CFR 36.2 at any time. However, installation of 

foundations or in-place assembly, erection, fabrication, or testing for any structure, system, or 

component of a facility or activity subject to 10 CFR Part 36 that have a reasonable nexus to 

radiological safety or security should not be performed prior to the submission of an application 

for a license and the fee required by 10 CFR 170.31. An activity or action has a reasonable 

nexus to radiological health and safety or the common defense and security if that activity or 

action has a rational, direct link to ensuring that a licensed materials facility is operating, or will 

operate, in accordance with the NRC’s regulations and in a manner that protects the public 

health and safety or the common defense and security from radiological hazards. Activities that 

have a reasonable nexus to radiological health and safety or common defense and security 

include, but are not limited to, construction of systems subject to 10 CFR Part 36, Subpart C, 

and the following: 

• Earthwork

• Pool excavation

• Footings and foundation for pool

• Irradiator foundations and walls

• Backfill pool

• Install pool liner

• Mechanical rough-in

• Electrical rough-in

• Shoring for roof

• Form and place roof

• Slab on grade

Subpart C of 10 CFR Part 36 currently lists the systems that have a nexus to radiological health 

and safety and defines the related engineering and safety concerns associated with each

system:

• Access Control: Adequacy of access control systems using interlocks and radiation 

monitors to prevent inadvertent entry to areas where radiation sources are unshielded; 

to provide emergency exits; and to ensure compliance with all the requirements of 10 

CFR 36.23. For computer-controlled access-control systems, licensing staff should 

consider expert evaluation of the software/system logic before operational testing.

• Site: Potential need for protection against flooding and earth slides.

• Base (soil, rock) for the Pool and Shielding Structures: Strength, settlement, 

liquefaction, ground water, soil compaction.

• Footers and Foundations for the Pool and Shielding Structures: Strength and 

reinforcement, alignment with pool and shielding structures.

• Pool and Shielding Structures: Strength and reinforcement, proper density of shielding 

materials, correct dimensions, minimization of voids in concrete or other shielding.

• Pool Liner: Contact with pool structure, penetrations in the liner, leak-tight welds.

• Pool Plumbing: Makeup water system; water cleanup system; effect of construction 

materials on pool-water chemistry; drainage system (potentially contaminated spilled 

water should flow into the pool); siphon breakers; radiation detection and alarm systems.

• Penetrations Through Shielding: Any significant effect on structural strength, shielding, 

or both.
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• Source Rack Protection: If the product to be irradiated moves on a product conveyor 

system, the source rack and the mechanism that moves the rack must be protected by a 

barrier or guides to prevent products and product carriers from hitting or touching the 

rack or mechanism.

• Source-Rack Mechanical Positioning System: Strength and stiffness of the rack and 

positioning cables or chains, source shroud will not interfere with source positioning, 

adequacy of motive power, potential for jamming.

• Source-Rack Movement and Position-Sensing System: Structural attachments for 

electrical and mechanical transducers, adequacy of transducers for interacting with the 

source-rack control system.

• Source-Rack Electrical Control System: Adequacy of the design of logistical and 

operational electrical circuitry and electromechanical components, to ensure 

unambiguous response of the system, which includes programmable controllers or 

computers and their interaction with operations, interlocks, doors, signals, and alarms.

• Source-Leak Detection: Adequacy of systems for detecting and isolating leaking 

sources.

• Hard Wiring: Adequacy of wire gauge and insulation to safely carry design currents and 

to withstand radiation and ozone damage if exposed; locating and attaching wiring to 

prevent fretting, wear, and exposure to potential fire hazards; accessibility to wiring for 

inspection and repair.

• Uninterruptable Electrical Power Supply: Adequate and reliable power capability to 

operate all electrical systems that are important to safety (including backup power 

sources); compatibility of the power supply with the electrical system.

• Fire Protection System: Adequacy to detect fire and smoke and to be manually as well 

as automatically initiated; must ensure that raised sources are immediately lowered into 

the pool.

• Emergency Systems for Returning an Up-stuck Source Rack to the Pool: Capability of 

the electrical control system to sense and signal the occurrence of an up-stuck sourcerack; adequacy of mechanical or electrical means for personnel to safely release and 

lower the rack; need for, and adequacy of, a system to cool the source-rack until it can 

be released and lowered.

• Ozone Ventilation System: Capability of the system to be properly initiated and to 

provide adequate volume flow rate of air to protect personnel and components.

• System for Transferring Sources from and to Transport Vehicles: Adequately sized 

openings in the shield-structure roof if sources are roof-loaded; structural adequacy of 

the roof-shield plug and its supports for its removal and replacement; structural and 

mechanical adequacy of systems for moving shipping containers into and out of the pool 

area.

URANIUM CONVERSION FACILITIES, ENRICHMENT FACILITIES, FUEL FABRICATION 

FACILITIES, AND URANIUM HEXAFLUORIDE (UF6) DECONVERSION FACILITIES (10 CFR 

PART 40 and 10 CFR PART 70)

If any of the following actions are performed before the NRC staff has completed its 

environmental review process, then the NRC has grounds for denial of a license application, in 

accordance with 10 CFR 40.32(e), and 70.23(a)(7):

1. Procurement or construction of engineered items that are items relied on for safety 

(IROFS) required to meet the performance requirements of 10 CFR 70.61.
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2. Construction of guard stations, fences, vehicle barriers, or other features that are, or will 

become, components of physical security systems required by regulations or orders.

3. Construction or installation of equipment whose purpose is the detection of radioactive 

material accidents or mitigation of the consequences of radioactive material accidents.

4. Installation of storage tanks that contain chemicals that could affect the safety of 

licensed material. 

5. Construction of facilities or warehouses that will be used for operations involving 

licensed material.

6. Driving of piles; subsurface preparation; placement of backfill, concrete, or permanent 

retaining walls within an excavation; installation of foundations; or in-place assembly, 

erection, fabrication, or testing, which are for IROFS and on-site emergency facilities.

7. Erection of buildings, offices, construction trailers and warehouses that will become part 

of a Standard Practice Procedures Plan for Protection of Classified Information.

Construction includes the onsite, in-place fabrication, erection, integration, or testing activities 

for any safety related item. The terms “onsite, in place, fabrication, erection, integration, or 

testing” describe the process of constructing a fuel cycle facility in its final, onsite plant location, 

where components or modules are integrated into the final, in-plant location. Under the 

definition of “construction” applicants and existing licensees may be able to fabricate, assemble, 

and test components and modules in a shop building, warehouse, or laydown area, even if 

these facilities are located onsite. However, the installation or integration of that safety related 

equipment into its final plant location is a construction activity and should not be performed until 

after the NRC staff concludes its environmental review of the license application.

Excavation includes the removal of any soil, rock, gravel, or other material below the final 

ground elevation to the final parent material, and may be conducted prior to the conclusion of 

the NRC staff’s environmental review. However, placing permanent, nonstructural dewatering 

materials, mudmats, or engineered backfill in advance of placing the foundation and associated 

permanent retaining walls for buildings or structures that will contain licensed materials are 

construction activities and should not be performed prior to the conclusion of the NRC staff’s 

environmental review.

Construction includes driving piles for buildings or structures that will contain licensed materials. 

Hence the driving of piles for such buildings or structures should not be performed before the 

NRC staff concludes its environmental review. Driving piles that do not ensure the structural 

stability or integrity of buildings or structures within the scope of the definition of “construction” 

(e.g., piles driven to support the erection of a bridge for a temporary or permanent access road) 

is not “construction”; therefore, those piles may be driven prior to the conclusion of the NRC 

staff’s environmental review.

In addition to 10 CFR 40.4, 51.4, and 70.4 criteria that are used to determine the scope of 

activities that fall within the definition of construction, construction includes the necessary 

excavation for safety related items. A necessary excavation is the portion of an excavation that 

provides sufficient construction access to the structures that are within the definition of 

construction. Applicants should ensure, and NRC staff will confirm, that these construction 

activities are separate from, and do not result in, adverse interactions with construction-related

safety related item including influence on the stability (static and dynamic) analyses.

Construction includes any change made to the parent material in which the excavation occurs 

(e.g., soil compaction, rock grouting); the driving of piles; the installation of foundations; the 

installation of permanent drainage systems and geofabric; the placement of backfill, concrete 
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(e.g., mudmats) or other materials that will not be removed before placement of the foundation 

of a structure; the placement and compaction of a subbase; and the installation of reinforcing 

bars to be incorporated into the foundation of any safety related items that fall within the 

definition of construction. The foregoing items fall within the definition of construction because 

they have a rational, direct link to ensuring that a licensed materials facility is operating, or will 

operate, in accordance with the NRC’s regulations and in a manner that protects the public 

health and safety from radiological hazards.

ACTIVITIES WHICH HAVE NO REASONABLE NEXUS TO RADIOLOGICAL SAFETY OR 

SECURITY

The NRC has determined that, in general, the following activities at source, byproduct, and SNM 

facilities and irradiators listed in 10 CFR 30.4, 36.2, 40.4, and 70.4, do not have a reasonable 

nexus to radiological health and safety and the common defense and security may be 

performed by a licensee or applicant at any time. Note that in some circumstances, based on 

the specific licensing proposal, any of these activities could be determined to have a reasonable 

nexus to radiological health and safety or common defense and security and, based on that 

determination, these activities would be construction:

(1) Changes for temporary use of the land for public recreational purposes;

(2) Site exploration, including necessary borings to determine foundation conditions or

other preconstruction monitoring to establish background information related to the

suitability of the site, the environmental impacts of construction or operation, or the

protection of environmental values;

(3) Preparation of the site for construction of the facility, including clearing of the site,

grading, installation of drainage, erosion and other environmental mitigation measures,

and construction of temporary roads and borrow areas;

(4) Erection of fences and other access control measures that are not related to the safe

use of, or security of, radiological materials subject to 10 CFR Parts 30, 36, 40, or 70;

(5) Excavation;

(6) Erection of support buildings (e.g., construction equipment storage sheds, warehouse

and shop facilities, utilities, concrete mixing plants, docking and unloading facilities,

and office buildings) for use in connection with the construction of the facility;

(7) Building of service facilities (e.g., paved roads, parking lots, railroad spurs, exterior

utility and lighting systems, potable water systems, sanitary sewerage treatment

facilities, and transmission lines);

(8) Procurement or fabrication of components or portions of the proposed facility occurring

at other than the final, in-place location at the facility; or

(9) Taking any other action that has no reasonable nexus to:

(i) Radiological health and safety, or

(ii) Common defense and security.

While the above site preparation activities may not require prior NRC approval, other Federal, 

State, or Local permits may be required. 

FINAL RESOLUTION

This interim staff guidance will be incorporated into the next revisions of NUREG-1556, and 

NUREG-1520.
M-9

APPLICABILITY

This ISG is applicable to all 10 CFR Parts 30, 36, 40, and 70 license applicants and existing 

licensees considering site preparation activities or construction activities at a facility that is 

subject to, or will be subject to, the licensing requirements of these parts.

REFERENCES 

1) NUREG-1556, Volume 6, “Consolidated Guidance About Material Facilities: Program-

 Specific Guidance About 10 CFR Part 36 Irradiator Licenses,” January 1999.

2) NUREG-1556, Volume 12, “Consolidated Guidance About Materials Licenses: Program

 Specific Guidance About Possession Licenses for Manufacturing and Distribution,”

 December 2000.

3) NUREG-1520, “Standard Review Plan for the Review of a License Application for a Fuel

Cycle Facility,” Revision 1, May 2010.

4) Regulatory Issue Summary 2009-12, Uranium Recovery Policy Regarding Site Preparation

Activities at Proposed, Unlicensed Uranium Recovery Facilities, September 23, 2009,

ML092090353.

5) NUREG-1748, “Environmental Review Guidance for Licensing Actions Associated with

Materials Facilities,” August 2003.

6) DC/COL-ISG-4, “Interim Staff Guidance on the Definition of Construction and on Limited 

Work Authorizations,” February 9, 2009, ML082970729. 

7) Inspection Manual Chapter 2815, “Construction and Preoperational Inspection of Panoramic

Wet-Source-Storage Gamma Irradiators,” March 27, 2001, ML010990225. 

8) Docket No. 030-36974, Final Environmental Assessment Related to the Proposed Pa’ina

Hawaii, LLC, Underwater Irradiator in Honolulu, Hawaii; August 10, 2007; ML071150121.

9) Docket No. 70-7015, Environmental Assessment for an Exemption to 10 CFR Parts 30, 40,

and 70, Commencement of Construction Requirements, Areva Enrichment Services, Eagle

Rock Enrichment Facility, Bonneville County, Idaho, February 28, 2010, ML093220528.

10) NUREG-1811, “Environmental Impact Statement for an Early Site Permit at the North Anna

ESP Site,” December 2006.

11) NUREG-1947, “Final Supplemental Environmental Impact Statement for Combined License

(COLs) for Vogtle Electric Generating Plant Unit 3 and 4,” March 2011.

Technical Contact: Mary T. Adams

301-492-3113

E-mail: Mary.Adams@nrc.gov

APPENDIX N

SAFETY CULTURE STATEMENT OF POLICY

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Safety Culture

The safety culture policy statement was published in the Federal Register (76 FR 34773) on 

June 14, 2011, and can be found at: http://www.gpo.gov/fdsys/pkg/FR-2011-06-14/pdf/2011-

14656.pdf. It is also posted in the NRC’s Agencywide Documents Access and Management 

System (ADAMS) and can be found using Accession Number ML11146A047.

Safety Culture Policy Statement

The purpose of this Statement of Policy is to set forth the Commission’s expectation that 

individuals and organizations establish and maintain a positive safety culture commensurate

with the safety and security significance of their activities and the nature and complexity of their 

organizations and functions. This includes all licensees, certificate holders, permit holders,

authorization holders, holders of quality assurance program approvals, vendors and suppliers of 

safety-related components, and applicants for a license, certificate, permit, authorization, or 

quality assurance program approval, subject to NRC authority. The Commission encourages 

the Agreement States, Agreement State licensees and other organizations interested in nuclear 

safety to support the development and maintenance of a positive safety culture, as articulated in 

this Statement of Policy.

Nuclear Safety Culture is defined as the core values and behaviors resulting from a collective 

commitment by leaders and individuals to emphasize safety over competing goals to ensure 

protection of people and the environment. Individuals and organizations performing regulated 

activities bear the primary responsibility for safety and security. The performance of individuals 

and organizations can be monitored and trended and, therefore, may be used to determine 

compliance with requirements and commitments and may serve as an indicator of possible 

problem areas in an organization’s safety culture. The NRC will not monitor or trend values. 

These will be the organization’s responsibility as part of its safety culture program.

Organizations should ensure that personnel in the safety and security sectors have an 

appreciation for the importance of each, emphasizing the need for integration and balance to

achieve both safety and security in their activities. Safety and security activities are closely 

intertwined. While many safety and security activities complement each other, there may be

instances in which safety and security interests create competing goals. It is important that 

consideration of these activities be integrated so as not to diminish or adversely affect either; 

thus, mechanisms should be established to identify and resolve these differences. A safety 

culture that accomplishes this would include all nuclear safety and security issues associated 

with NRC-regulated activities.

Experience has shown that certain personal and organizational traits are present in a positive 

safety culture. A trait, in this case, is a pattern of thinking, feeling, and behaving that

emphasizes safety, particularly in goal conflict situations, e.g., production, schedule, and the 

cost of the effort versus safety. It should be noted that although the term ‘‘security’’ is not

expressly included in the following traits, safety and security are the primary pillars of the NRC’s 

regulatory mission. Consequently, consideration of both safety and security issues, 

commensurate with their significance, is an underlying principle of this Statement of Policy.
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The following are traits of a positive safety culture:

(1) Leadership Safety Values and Actions – Leaders demonstrate a commitment to safety in 

their decisions and behaviors;

(2) Problem Identification and Resolution – Issues potentially impacting safety are promptly

identified, fully evaluated, and promptly addressed and corrected commensurate with 

their significance;

(3) Personal Accountability – All individuals take personal responsibility for safety;

(4) Work Processes – The process of planning and controlling work activities is 

implemented so that safety is maintained;

(5) Continuous Learning – Opportunities to learn about ways to ensure safety are sought 

out and implemented;

(6) Environment for Raising Concerns – A safety-conscious work environment is maintained 

where personnel feel free to raise safety concerns without fear of retaliation, intimidation, 

harassment, or discrimination;

(7) Effective Safety Communication – Communications maintain a focus on safety;

(8) Respectful Work Environment – Trust and respect permeate the organization; and

(9) Questioning Attitude – Individuals avoid complacency and continuously challenge 

existing conditions and activities in order to identify discrepancies that might result in 

error or inappropriate action.

There may be traits not included in this Statement of Policy that are also important in a positive 

safety culture. It should be noted that these traits were not developed to be used for inspection

purposes.

It is the Commission’s expectation that all individuals and organizations, performing or 

overseeing regulated activities involving nuclear materials, should take the necessary steps to

promote a positive safety culture by fostering these traits as they apply to their organizational 

environments. The Commission recognizes the diversity of these organizations and 

acknowledges that some organizations have already spent significant time and resources in the 

development of a positive safety culture. The Commission will take this into consideration as 

the regulated community addresses the Statement of Policy.



UNITED STATES 

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

--------------------

OFFICIAL BUSINESS
 

 

NUREG-1556

Volume 4, Rev. 1 Draft

Consolidated Guidance about Materials Licenses:

Program-Specific Guidance about Fixed Gauge Licenses