NUREG–2116 

Safety Evaluation Report for the International 

Isotopes Fluorine Products, Inc. 

Fluorine Extraction Process and Depleted 

Uranium Deconversion Plant in Lea County, 

New Mexico 

ML113140271 

Docket No. 40-9086 

International Isotopes Fluorine Products, Inc. 

Manuscript Completed: May 2012 

Date Published: May 2012 

Office of Nuclear Material Safety and Safeguards 
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iii 

ABSTRACT 

This report documents the U.S. Nuclear Regulatory Commission’s (NRC’s) staff review and 

safety and safeguards evaluation of International Isotopes Fluorine Products, Inc.=s (IIFP), 

application for a license to construct a Fluorine Extraction Process & Depleted Uranium 

Deconversion Plant (FEP/DUP) and possess and use source materials. INIS proposes that the 

FEP/DUP be located in Lea County, New Mexico, approximately 24 kilometers (15 miles) west 

of the city of Hobbs. The facility will possess natural and depleted uranium, and significant 

quantities of hazardous chemicals derived from licensed material, e.g., hydrogen fluoride. 

The objective of this review is to evaluate the facility=s potential adverse impacts on worker and 

public health and safety, under both normal operations and accident conditions. The review 

also considers the management organization, administrative programs, and financial 

qualifications provided to ensure safe design and operation of the facility. 

The NRC staff concludes, in this safety evaluation report, that the applicant=s descriptions, 

specifications, and analyses provide an adequate basis for safety of facility operations and that 

operation of the facility does not pose an undue risk to the worker and public health and safety. 

Potential environmental impacts associated with the proposed facility and its reasonable 

alternatives are addressed in a separate document, “Environmental Impact Statement for the 

Proposed Fluorine Extraction Process and Depleted Uranium Deconversion Plant in Lea 

County, New Mexico,” NUREG-2116. The draft Environmental Impact Statement was published 

in the Federal Register, Vol. 77, dated Friday, January 13, 2012, see pages 2096-2098 and the 

final EIS is scheduled to be published in the fall of 2012. 
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v 

CONTENTS 

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

CONTENTS ................................................................................................................................ v

TABLES .................................................................................................................................... xi

EXECUTIVE SUMMARY .......................................................................................................... xiii

ABBREVIATIONS .................................................................................................................... xvii

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

1.1 Facility and Process Description ............................................................................ 1-1

1.1.1 Regulatory Requirements ........................................................................... 1-1

1.1.2 Regulatory Guidance and Acceptance Criteria ........................................... 1-1

1.1.3 Staff Review and Analysis .......................................................................... 1-1

1.2 Institutional Information .......................................................................................... 1-4

1.2.1 Regulatory Requirements ........................................................................... 1-4

1.2.2 Regulatory Guidance and Acceptance Criteria ........................................... 1-4

1.2.3 Staff Review and Analysis .......................................................................... 1-4

1.3 Site Description .................................................................................................... 1-11

1.3.1 Regulatory Requirements ......................................................................... 1-11

1.3.2 Regulatory Guidance and Acceptance Criteria ......................................... 1-11

1.3.3 Staff Review and Analysis ........................................................................ 1-11

1.3.4 Site Description Conclusions .................................................................... 1-29

1.4 References ........................................................................................................... 1-29

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

2.1 Regulatory Requirements ....................................................................................... 2-1

2.2 Regulatory Guidance and Acceptance Criteria ....................................................... 2-1

2.3 Staff Review and Analysis ...................................................................................... 2-1

2.3.1 Organization ............................................................................................... 2-2

2.3.2 Pre-operational Testing and Initial Startup .................................................. 2-2

2.3.3 Organizational Responsibilities and Qualifications ...................................... 2-3

2.3.4 Procedures ................................................................................................. 2-6

2.3.5 Incidents and Items of Concern .................................................................. 2-6

2.4 Evaluation Findings ................................................................................................ 2-7

2.5 References ............................................................................................................. 2-7

3.0 INTEGRATED SAFETY ANALYSIS AND INTEGRATED SAFETY ANALYSIS 

SUMMARY .................................................................................................................. 3-1

3.1 Regulatory Requirements ....................................................................................... 3-1

3.2 Regulatory Guidance and Acceptance Criteria ....................................................... 3-3

3.3 Staff Review and Analysis ...................................................................................... 3-3

3.3.1 Safety Program and Integrated Safety Analysis Commitments ................... 3-4

3.3.2 Integrated Safety Analysis Methodology ..................................................... 3-7

3.3.3 Integrated Safety Analysis Integration and Team ...................................... 3-13
vi 

3.4 Integrated Safety Analysis Summary Review Conclusion ..................................... 3-14

3.5 References ........................................................................................................... 3-16

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

4.1 Regulatory Requirements ....................................................................................... 4-1

4.2 Regulatory Guidance and Acceptance Criteria ....................................................... 4-1

4.3 Staff Review and Analysis ...................................................................................... 4-1

4.3.1 Radiation Protection Program Implementation ............................................ 4-1

4.3.2 As Low As Reasonably Achievable Program .............................................. 4-3

4.3.3 Organization and Personnel Qualifications ................................................. 4-5

4.3.4 Written Procedures ..................................................................................... 4-6

4.3.5 Training ...................................................................................................... 4-7

4.3.6 Ventilation and Respiratory Protection Programs ........................................ 4-9

4.3.7 Radiation Survey and Monitoring Programs .............................................. 4-11

4.3.8 Additional Program Requirements ............................................................ 4-16

4.3.9 Control of Radiological Risk Resulting from Accidents .............................. 4-17

4.4 Evaluation Findings .............................................................................................. 4-17

4.5 References ........................................................................................................... 4-18

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

5.1 Regulatory Requirements ....................................................................................... 5-1

5.2 Evaluation of Findings ............................................................................................ 5-1

5.3 References ............................................................................................................. 5-1

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

6.1 Regulatory Requirements ....................................................................................... 6-1

6.2 Regulatory Guidance and Acceptance Criteria ....................................................... 6-1

6.3 Staff Review and Analysis ...................................................................................... 6-1

6.3.1 Process Description .................................................................................... 6-2

6.3.2 Chemical Accident Sequences ................................................................. 6-12

6.3.3 Chemical Accident Consequences ........................................................... 6-12

6.3.4 Items Relied on for Safety and Management Measures ............................ 6-14

6.3.5 Emergency Management .......................................................................... 6-19

6.3.6 Baseline Design Criteria and Defense-in-Depth ........................................ 6-20

6.4 Evaluation Findings .............................................................................................. 6-21

6.5 References ........................................................................................................... 6-22

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

7.1 Regulatory Requirements ....................................................................................... 7-1

7.2 Regulatory Acceptance Criteria .............................................................................. 7-1

7.3 Staff Review and Analysis ...................................................................................... 7-1

7.3.1 Fire Safety Management Measures ............................................................ 7-2

7.3.2 Fire Hazards Analysis ................................................................................. 7-3

7.3.3 Facility Design ............................................................................................ 7-9

7.3.4 Process Fire Safety .................................................................................. 7-12
vii 

7.3.5 Fire and Emergency Response................................................................. 7-14

7.3.6 Baseline Design Criteria ........................................................................... 7-15

7.4 Evaluation Findings .............................................................................................. 7-16

7.5 References ........................................................................................................... 7-16

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

8.1 Regulatory Requirements ....................................................................................... 8-1

8.2 Regulatory Acceptance Criteria .............................................................................. 8-1

8.3 Staff Review and Analysis ...................................................................................... 8-1

8.3.1 Facility Description ...................................................................................... 8-1

8.3.2 Onsite and Offsite Emergency Facilities ..................................................... 8-2

8.3.3 Types of Accidents ..................................................................................... 8-3

8.3.4 Classification of Accidents .......................................................................... 8-4

8.3.5 Detection of Accidents ................................................................................ 8-4

8.3.6 Mitigation of Consequences ........................................................................ 8-5

8.3.7 Assessment of Releases ............................................................................ 8-5

8.3.8 Responsibilities ........................................................................................... 8-6

8.3.9 Notification and Coordination ...................................................................... 8-6

8.3.10 Information To Be Communicated .............................................................. 8-7

8.3.11 Training ...................................................................................................... 8-7

8.3.12 Recovery and Plant Restoration (Safe Shutdown) ...................................... 8-8

8.3.13 Exercises and Drills .................................................................................... 8-9

8.3.14 Responsibilities for Developing and Maintaining the Emergency Plan 

and its Procedures ...................................................................................... 8-9

8.4 Evaluation Findings .............................................................................................. 8-10

8.5 References ........................................................................................................... 8-10

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

9.1 Regulatory Requirements ....................................................................................... 9-1

9.2 Regulatory Acceptance Criteria .............................................................................. 9-1

9.3 Staff Review and Analysis ...................................................................................... 9-2

9.3.1 Personnel Training ...................................................................................... 9-2

9.3.2 Personnel Qualifications ............................................................................. 9-2

9.3.3 Radiation Safety ......................................................................................... 9-4

9.3.4 Effluent and Environmental Monitoring ..................................................... 9-10

9.3.5 Integrated Safety Analysis Summary ........................................................ 9-21

9.3.6 10 CFR 70.61(c)(3) ................................................................................... 9-22

9.4 Evaluation Findings .............................................................................................. 9-22

9.5 References ........................................................................................................... 9-23

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

10.1 Regulatory Requirements ..................................................................................... 10-1

10.2 Regulatory Guidance and Acceptance Criteria ..................................................... 10-1

10.3 Staff Review and Analysis .................................................................................... 10-2

10.3.1 Conceptual Decontamination and Decommissioning Plan ........................ 10-3
viii 

10.3.2 Decommissioning Strategy ....................................................................... 10-3

10.3.3 IIFP Facility Description ............................................................................ 10-4

10.3.4 Design Features ....................................................................................... 10-4

10.3.5 Worker Exposure and Waste Volume Control ........................................... 10-5

10.3.6 Management Organization ........................................................................ 10-6

10.3.7 Health and Safety ..................................................................................... 10-6

10.3.8 Waste Management .................................................................................. 10-6

10.3.9 Physical Security and Material Control and Accounting ............................ 10-6

10.3.10 Recordkeeping ......................................................................................... 10-7

10.3.11 Decommissioning Process ........................................................................ 10-7

10.3.12 Size Reduction and Packaging Facility Outfitting ...................................... 10-9

10.3.13 System Preparation .................................................................................. 10-9

10.3.14 Dismantling ............................................................................................... 10-9

10.3.15 Decontamination and Size Reduction ..................................................... 10-10

10.3.16 Salvage of Equipment and Materials....................................................... 10-10

10.3.17 Disposal .................................................................................................. 10-10

10.3.18 Final Radiation Survey ............................................................................ 10-10

10.3.19 Decommissioning Impact on Integrated Safety Analysis ......................... 10-11

10.3.20 NRC Staff Determination ........................................................................ 10-11

10.4 Site-Specific Cost Estimate ................................................................................ 10-12

10.4.1 Regulatory Requirements ....................................................................... 10-12

10.4.2 Summary of Staff Review ....................................................................... 10-12

10.4.3 Staff Review of Decommissioning Funding Plan and Cost Estimate ....... 10-12

10.4.4 Evaluation Findings ................................................................................ 10-14

10.5 References ......................................................................................................... 10-15

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

11.1 Regulatory Requirements ..................................................................................... 11-1

11.2 Regulatory Acceptance Criteria ............................................................................ 11-1

11.3 Staff Review and Analysis .................................................................................... 11-2

11.3.1 Configuration Management ....................................................................... 11-2

11.3.2 Maintenance ............................................................................................. 11-6

11.3.3 Training and Qualifications ....................................................................... 11-9

11.3.4 Procedures Development and Implementation ....................................... 11-12

11.3.5 Audits and Assessments......................................................................... 11-17

11.3.6 Incident Investigations and Corrective Action Process ............................ 11-20

11.3.7 Records Management and Document Control ........................................ 11-21

11.3.8 Quality Assurance Program Elements .................................................... 11-22

11.4 Evaluation Findings ............................................................................................ 11-23

11.4.1 Configuration Management ..................................................................... 11-23

11.4.2 Maintenance ........................................................................................... 11-23

11.4.3 Training and Qualification ....................................................................... 11-24

11.4.4 Procedures ............................................................................................. 11-24

11.4.5 Audits and Assessments......................................................................... 11-25

11.4.6 Incident Investigations ............................................................................ 11-25

11.4.7 Records Management ............................................................................ 11-25

11.4.8 Other Quality Assurance Elements ......................................................... 11-26

11.5 References ......................................................................................................... 11-27
ix 

12.0 QUALITY ASSURANCE PROGRAM DESCRIPTION ................................................ 12-1

12.1 Organization ......................................................................................................... 12-1

12.1.1 Program .................................................................................................... 12-2

12.1.2 Design Control .......................................................................................... 12-6

12.1.3 Procurement Document Control................................................................ 12-9

12.1.4 Instructions, Procedures, and Drawings .................................................. 12-10

12.1.5 Document Control ................................................................................... 12-11

12.1.6 Control of Purchased Items and Services ............................................... 12-12

12.1.7 Identification and Control of Materials, Parts, and Components .............. 12-14

12.1.8 Control of Special Processes .................................................................. 12-15

12.1.9 Inspection ............................................................................................... 12-16

12.1.10 Test Control ............................................................................................ 12-17

12.1.11 Control of Measuring and Test Equipment .............................................. 12-18

12.1.12 Inspection, Test, and Operating Status ................................................... 12-19

12.1.13 Control of Nonconforming Items ............................................................. 12-20

12.1.14 Corrective Action .................................................................................... 12-21

12.1.15 Quality Assurance Records .................................................................... 12-22

12.1.16 Audits ..................................................................................................... 12-23

12.2 Evaluation of Findings ........................................................................................ 12-25

12.3 References ......................................................................................................... 12-26

13.0 PHYSICAL PROTECTION AND MATERIAL CONTROL AND ACCOUNTING .......... 13-1

13.1 Physical Protection ............................................................................................... 13-1

13.1.1 Regulatory Guidance ................................................................................ 13-1

13.1.2 Staff Review and Analysis ........................................................................ 13-1

13.1.3 Evaluation of Findings for Physical Protection .......................................... 13-1

13.2 Material Control and Accounting .......................................................................... 13-2

13.2.1 Regulatory Requirements for Material Control and Accounting ................. 13-2

13.2.2 Regulatory Acceptance Criteria ................................................................ 13-2

13.2.3 Staff Review and Analysis ........................................................................ 13-3

13.2.4 Evaluation Findings for Material Control and Accounting .......................... 13-3

13.3 References ........................................................................................................... 13-4

14.0 Human Factors .......................................................................................................... 14-1

14.1 Regulatory Requirements ..................................................................................... 14-1

14.2 Regulatory Guidance and Acceptance Criteria ..................................................... 14-1

14.3 Staff Review and Analysis .................................................................................... 14-2

14.3.1 Identification of Personnel Activities .......................................................... 14-2

14.3.2 Human Factors Engineering Design Review Planning .............................. 14-3

14.3.3 Operating Experience ............................................................................... 14-6

14.3.4 Functional Allocation Analysis and Task Analysis ..................................... 14-7

14.3.5 Human System Interface Design, Inventory, and Characterization ........... 14-9

14.3.6 Staffing ................................................................................................... 14-11

14.3.7 Procedure Development ......................................................................... 14-12

14.3.8 Training Program Development .............................................................. 14-13

14.3.9 Human Factors Verification and Validation ............................................. 14-14

14.4 Evaluation Findings ............................................................................................ 14-16
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14.5 References ......................................................................................................... 14-16

15.0 SAFETY EVALUATION REPORT PREPARERS ....................................................... 15-1

Appendix A ACCIDENT ANALYSIS .............................................................................. A-1

Appendix B STRUCTURAL AND GEOTECHNICAL DESIGN ....................................... B-1

Appendix C INSTRUMENTATION AND CONTROL ...................................................... C-1

Appendix D ELECTRICAL SYSTEMS ........................................................................... D-1

 
xi 

TABLES 

Table 1-1 Types, Forms, Quantities, and Proposed Authorized Uses Of Licensed 

Materials .............................................................................................................. 1-11

Table 6-1 Chemical Consequence Levels and Values (IIFP, 2012a) ................................... 6-13

Table 6-2 Worker Chemical Exposure Values (10 Minutes) (IIFP, 2012b) ........................... 6-13

Table 6-3 Public Chemical Exposure Values (30 Minutes) (IIFP, 2012b) ............................. 6-14

Table 15-1 U.S. Nuclear Regulatory Commission Contributors ............................................. 15-1

 
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xiii 

EXECUTIVE SUMMARY 

On December 30, 2009, International Isotopes Fluorine Products, Inc. (IIFP), a subsidiary of 

International Isotopes, Inc. (INIS), submitted to the U.S. Nuclear Regulatory Commission (NRC) 

an application requesting a license, under Title 10 of the Code of Federal Regulations (10 CFR) 

Part 40, “Domestic Licensing of Source Material.” Subsequently, beginning February 9, 2011, 

and concluding October 17, 2011, IIFP also provided 26 separate responses to NRC’s requests 

for additional information (RAI). This Safety Evaluation Report (SER) primarily references the 

information which was incorporated into the Revision B of the IIFP application, submitted in April 

2012. When applicable, the SER directly references the appropriate RAIs directly. If approved, 

the license would authorize construction and operation of a fluoride extraction and depleted 

uranium deconversion facility to possess and use byproduct and source material. The applicant 

proposed to locate the facility in Lea County, NM, approximately 24 kilometers (15 miles) west 

of Hobbs, NM. The proposed facility would be capable of deconverting approximately 

3.3 million kilograms per year (7.3 million pounds per year) of depleted uranium hexafluoride 

into fluoride products and uranium oxide for disposal. The facility will possess significant 

quantities of depleted uranium hexafluoride, silicon tetrafluoride, boron trifluoride, and 

anhydrous hydrogen fluoride. 

The license application (LA) also indicates that IIFP anticipates expanding the capacity of the 

facility after 3–4 years of operating experience. The Environmental Impact Statement 

addresses this future expansion as a reasonably foreseeable action. However, the LA does not 

request or evaluate the expansion, which would require a separate licensing action and 

environmental review. 

IIFP submitted an application for a source material license regulated in accordance with 

10 CFR Part 40, Domestic Licensing of Source Material, among other regulations. The NRC 

staff conducted its safety review of the LA, which describes the proposed operations, safety 

commitments, and other required programs (e.g., decommissioning, environmental protection, 

management measures). The applicant submitted an integrated safety analysis (ISA), which 

the NRC staff reviewed in accordance with 10 CFR Part 70, “Domestic Licensing of Special 

Nuclear Material,” as directed by the Commission in SRM-SECY-07-0146, “Staff Requirements–

Regulatory Options for Licensing New Uranium Conversion and Depleted Uranium 

Deconversion Facilities,” dated October 10, 2007. The NRC staff notified the applicant that the 

10 CFR Part 70 ISA requirements would eventually be superseded by similar requirements in 

10 CFR Part 40, which is undergoing rulemaking to incorporate ISA requirements. The NRC 

staff also reviewed the applicant’s Quality Assurance Program Description, Emergency Plan 

(EP), and Security Plan. 

The NRC staff used several guidance documents to evaluate the LA and to complete the SER 

for the IIFP facility. The 10 CFR Part 40 deconversion application does not have a dedicated 

Standard Review Plan, but many of the regulatory requirements for 10 CFR Part 40 fuel cycle 

facilities are similar to 10 CFR Part 70 fuel cycle facilities. The NRC staff used primarily 

NUREG-1520, Revision 0, “A Standard Review Plan for the Review of a License Application for 

a Fuel Cycle Facility,” to review the application. The updated revision of the NUREG was not 

available until after IIFP had submitted its application in 2009. 

NUREG-1520 is the Standard Review Plan for 10 CFR Part 70 reviews, and the hazards at the 

proposed IIFP facility are similar to 10 CFR Part 70 fuel cycle facilities. These hazards include 

processing of depleted uranium hexafluoride, use of hazardous chemicals (e.g., hydrogen 
xiv 

fluoride), and the associated safety systems. Certain portions of the NUREG do not apply to 

10 CFR Part 40 facilities, such as criticality, because of the absence of special nuclear material. 

In addition, the NRC staff used Regulatory Guide (RG) 5.59, Revision 1, “Standard Format and 

Content for a Licensee Physical Security Plan for the Protection of Special Nuclear Material of 

Moderate or Low Strategic Significance,” issued February 1983, to review the security-related 

portions of the application. 

In conducting its safety review, the NRC staff assessed, among other things, whether the 

applicant’s proposed equipment, facilities, and procedures would adequately protect public 

health and safety and the environment. The NRC staff evaluated the applicant’s facility designs, 

procedures, and stated commitments. The Commission will only issue a license if it makes a 

determination that the proposed facility complies with the regulatory requirements, as 

demonstrated in the application. 

Once a license is granted, construction of the facility may begin. In accordance with the ISA 

annual update requirements, the applicant (then licensee) will submit to the NRC annual 

updates to the ISA Summary along with a brief summary of the nonsafety-related changes 

made throughout the year. The NRC reviews these submissions as well as any license 

amendment requests that may be submitted. 

The applicant also submitted an Environmental Report, which the NRC staff used to prepare 

NUREG-2113, “Draft Environmental Impact Statement for the Proposed Fluorine Extraction 

Process and Depleted Uranium Deconversion Plant” issued February 27, 2012 (NRC, 2012). 

The LA contains IIFP’s commitments which demonstrate compliance with the NRC’s regulatory 

requirements, and the entire application is tied down in Safety Condition S-1 of the license. A 

summary of NRC’s review and findings in each of the review areas follows. 

General Information

The applicant adequately described the facility and processes, providing the NRC staff with an 

overall understanding of the relationships of the facility features and their functions. The 

application properly explained and outlined the financial qualifications. The description of the 

site included important information about regional hydrology, geology, meteorology, the nearby 

population, and potential effects of natural phenomena at the facility. 

Organization and Administration

The applicant adequately described the responsibilities and associated resources for the 

design, construction, and operation of the facility and its plans for managing the project. The 

plans and commitments described in the application provide reasonable assurance that an 

acceptable organization; appropriate administrative policies; and sufficient resources have been 

established or committed for the design, construction, and safe operation of the facility. 
xv 

Integrated Safety Analysis and Summary

The applicant provided sufficient information about the site, facility processes, hazards, and 

types of accident sequences. The information provided addressed each credible event, the 

potential radiological and chemical consequences of the event, and the likelihood of the event. 

No mitigated event consequence exceeds the performance requirements of 10 CFR 70.61, 

“Performance Requirements.” The applicant also provided adequate information about items 

relied on for safety (IROFS). 

Radiation Protection

The applicant provided sufficient information to evaluate the Radiation Protection Program. The 

application adequately describes the (1) qualification requirements, (2) written radiation 

protection procedures, (3) radiation work permit program, (4) program for ensuring that worker 

and public doses are as low as reasonably achievable, and (5) necessary training for all 

personnel who have access to radiologically restricted areas. The radiation survey and 

monitoring program is adequate to protect workers and members of the public who may be 

potentially exposed to radiation. 

Chemical Process Safety

The applicant adequately described and assessed accident consequences that could result 

from the handling, storage, or processing of licensed materials and that could have potentially 

significant chemical consequences and effects. The applicant performed hazard analyses that 

identified and evaluated chemical process hazards and potential accidents and established 

safety controls that meet the regulatory requirements. 

Fire Safety

The applicant committed to reasonable engineered and administrative controls to adequately 

minimize the risk of fires and explosions. The IROFS and defense-in-depth protection 

discussed in the applicant’s ISA Summary, along with safety basis assumptions and the 

planned programmatic commitments in the LA, meet safety requirements and provide 

reasonable assurance that the facility is protected against fire hazards. 

Emergency Management

The applicant provided an adequate EP for the facility that meets the regulatory requirements. 

The applicant committed to maintaining and executing an EP for responding to the radiological 

and chemical hazards resulting from potential release of radioactive or chemically hazardous 

materials incident to the processing of licensed material. Approved, written procedures 

implement the requirements of the EP. 

Environmental Protection

The applicant committed to adequate environmental protection measures, including 

(1) environmental and effluent monitoring and (2) effluent controls to maintain public doses as 

low as reasonably achievable as part of the Radiation Protection Program. The applicant’s 

proposed controls are adequate to protect the environment and the health and safety of the 

public, and they comply with the regulatory requirements. 
xvi 

Decommissioning

The applicant provided a conceptual Decommissioning Plan for the facility that adequately 

addresses (1) contamination control, (2) control of worker exposures and waste volumes, 

(3) waste disposal, (4) the final radiation survey, (5) control of special nuclear material, 

(6) control of classified matter, and (7) recordkeeping for decommissioning. 

The applicant provided a Decommissioning Funding Plan (DFP) for the facility that adequately 

demonstrates that adequate funding will be available for decommissioning and that 

decommissioning will not pose a threat to public health and safety or the environment. The 

NRC staff will add a license condition, if a license is issued, to ensure that the applicant 

provides final copies of the proposed financial assurance instruments to the NRC for review at 

least 6 months before the planned date for possessing licensed material, as well as final 

executed copies of the reviewed financial assurance instruments before the receipt of licensed 

material. The DFP is acceptable because it provides sufficient funding to ensure that 

decommissioning and decontamination of the facility can be accomplished, even if the licensee 

is unable to meet its financial obligations. 

Management Measures

The applicant provided sufficient information about management measures that will be applied 

to the project. The information describes (1) the overall configuration management program 

and policy, (2) the maintenance program, (3) training, and (4) the process for the development, 

approval, and implementation of procedures. The applicant explained the audits and 

assessments program, as well as incident investigations and the records management system. 

The applicant committed to establishing and documenting surveillances, tests, and inspections 

to provide reasonable assurance of satisfactory performance of the IROFS. The proposed 

management measures are acceptable and meet the regulatory requirements in 

10 CFR 70.62(d). 

Materials Control and Accountability

The applicant provided adequate information describing the fundamental nuclear material 

control plan for the project. The plan describes the programs to be used to control and account 

for licensed material in the facility. The program meets the applicable regulatory requirements 

in 10 CFR Part 74, “Material Control and Accounting of Special Nuclear Material.” 

Physical Protection

The applicant provided adequate information on the policies, methods, and procedures to be 

implemented to protect the facility. This information is acceptable and meets the requirements 

in 10 CFR Part 73, “Physical Protection of Plants and Materials.” 

Reference

(NRC, 2012) U.S. Nuclear Regulatory Commission, “Draft Environmental Impact Statement for 

the Proposed Fluorine Extraction Process and Depleted Uranium Deconversion Plant,” 

NUREG-2113, 2012. 

 
xvii 

ABBREVIATIONS 

ADAMS Agencywide Documents Access and Management 

System 

AEGL acute exposure guideline level 

AHF anhydrous hydrogen fluoride 

ALARA as low as reasonably achievable 

ALOHA Area Locations of Hazardous Atmospheres 

(computer code) 

ANSI American National Standards Institute 

ASCE American Society of Civil Engineers 

ASME American Society of Mechanical Engineers 

ASTM American Society for Testing and Materials 

BDC baseline design criteria 

BF3 boron trifluoride 

B2O3 diboron trioxide 

Bq becquerels 

C Celsius 

CAA controlled access area 

CaF2 calcium fluoride 

CAP corrective action program 

CEDE committed effective dose equivalent 

CFR Code of Federal Regulations

Ci curie 

cm centimeter 

CM configuration management 

CEO Chief Executive Officer 

COO Chief Operations Officer 

DAC derived air concentration 

DB design and build 

DCE Decommissioning Cost Estimate 

DFP Decommissioning Funding Plan 

DOE U.S. Department of Energy 

DOT U.S. Department of Transportation 

DR damage ratio 

DUF4 depleted uranium tetrafluoride 

DUF6 depleted uranium hexafluoride 

DUO depleted uranium oxide 

EAL environmental assessment lead 

EMT emergency medical technician 

EOC emergency operations center 

EP Emergency Plan 

EPA U.S. Environmental Protection Agency 

EPIP emergency plan implementation procedure 

EPP environmental protection process 

ER environmental report 
xviii 

ERO emergency response organization 

ERPG emergency response planning guideline 

ESH environmental safety and health 

ESO environmental and safety officer 

F Fahrenheit 

FAA functional allocation analysis 

FEMA Federal Emergency Management Agency 

FEP fluorine extraction process 

FEP/DUP fluorine extraction process and depleted uranium 

deconversion plant 

FHA fire hazards analysis 

FNMCP fundamental nuclear material control plan 

FSRC Facility Safety Review Committee 

ft feet 

ft/s feet per second 

g acceleration of gravity 

gal gallon 

gpm gallons per minute 

H3BO3 boric acid 

HBF4 fluoroboric acid 

HDBK handbook 

HED human engineering discrepancies 

HF hydrogen fluoride 

HFE human factors engineering 

HPS Health Physics Society 

HS&E health, safety, and environment 

HVAC heating, ventilation, and air conditioning 

ICRP International Commission on Radiological 

Protection 

IEEE Institute for Electrical and Electronics Engineering 

IIFP International Isotopes Fluorine Products, Inc. 

in. inch 

LLW low-level radioactive waste 

INIS International Isotopes, Inc. 

lpm liters per minute 

IROFS item relied on for safety 

ISA integrated safety analysis 

I&C instrumentation and control 

KF potassium fluoride 

kg kilogram 

km kilometer 

km/h kilometer per hour 

KOH potassium hydroxide 

kPa kilopascal 

L liter 
xix 

LA license application 

lb pound 

m meter 

mb body wave magnitude 

ML local magnitude 

m3 cubic meter 

MC&A material control and accounting 

μCi microcurie 

MDC minimum detectable concentration 

MEI maximally exposed individual 

mi mile 

MIL military 

mg milligram 

mm millimeter 

MM Modified Mercalli 

MOA military operation area 

MOU memorandum of understanding 

mph miles per hour 

mrem millirem 

mrem/yr millirem per year 

m/s meter per second 

mSv millisievert 

mSv/yr millisiever per year 

MSDS material safety data sheet 

M&TE measuring and test equipment 

NCS Nuclear Criticality Cafety 

NELAC National Environmental Laboratory Accreditation 

Conference 

NFPA National Fire Protection Association 

NIST National Institute of Standards and Technology 

NMCBC New Mexico Commercial Building Code 

NMED State of New Mexico Environment Department 

NPDES National Pollutant Discharge Elimination System 

NOAA National Oceanic and Atmospheric Administration 

NRC U.S. Nuclear Regulatory Commission 

NVLAP National Voluntary Laboratory Accreditation 

Program 

OSHA U.S. Occupational Safety and Health Administration 

OER operational experience review 

P&E purge and evacuation 

PFD process flow diagram 

PFPE perfouropolyether 

PHA process hazards analysis 

PM plant manager 

PM preventive maintenance 

psi pounds per square inch 

psig pounds per square inch gauge 
xx 

PSB payment surety bond 

PSP physical security plan 

QA quality assurance 

QAPD quality assurance program description 

QL quality level 

QMS quality management system 

RAI request for additional information 

RCRA Resource Conservation and Recovery Act 

rem roentgen equivalent man 

REMP Radiological Environmental Monitoring Program 

RQAD Regulatory Affairs and Quality Assurance Director 

RPM radiation protection manager 

RWP radiation work permit 

SEI Structural Engineering Institute 

SER safety evaluation report 

SiF4 silicon tetrafluoride 

SiO2 silicon dioxide 

SNM special nuclear material 

SNM-LSS special nuclear material—low strategic significance 

SSC structure, system, and components 

STA standby trust agreement 

STD standard 

Sv sievert 

TA task analysis 

TEDE total effective dose equivalent 

TSV task support verification 

UF4 uranium tetrafluoride 

UF6 uranium hexafluoride 

UO2 uranium dioxide 

UO2F2 uranyl fluoride 

U3O8 triuranium octoxide 

USGS U.S. Geological Survey 

V&V verification and validation 

WIP work in process 

 
1-1 

1.0 GENERAL INFORMATION 

1.1 Facility and Process Description

The purpose of the U.S. Nuclear Regulatory Commission’s (NRC’s) review of the International 

Isotopes Fluorine Products, Inc.’s (IIFP or the applicant) facility and process description is to 

evaluate whether the license application (LA) (IIFP, 2012a) includes an overview of the facility 

layout and a summary description of the proposed processes. The Integrated Safety Analysis 

(ISA) Summary (IIFP, 2012b) contains a more detailed description of the facility and processes. 

Some information related to the ISA Summary for the site description has been marked by the 

applicant as "Security-Related Information," pursuant to Title 10 of the Code of Federal 

Regulations (10 CFR) 2.390, and is not included in the public version of this report. 

1.1.1 Regulatory Requirements 

The regulations in 10 CFR 40.32, entitled, “General Requirements for Issuance of Specific 

Licenses,” require each application for a license to include information on the proposed activity 

and the equipment and facilities that the applicant will use to protect health and minimize danger 

to life and property. In addition, the regulations in 10 CFR 70.65, “Additional Content of 

Applications,” require each application to include a general description of the facility, with 

emphasis on those areas that could affect safety, including identification of the controlled area 

boundaries. The LA, including license commitments, is incorporated directly into the license by 

reference as Safety Condition S-1. 

1.1.2 Regulatory Guidance and Acceptance Criteria 

Chapter 1 of NUREG-1520, Revision 0, “Standard Review Plan for the Review of a License 

Application for a Fuel Cycle Facility,” issued March 2002 (NRC, 2002), contains the guidance 

applicable to NRC’s review of the facility and process description section of the LA 

(IIFP, 2012a). This chapter of NUREG-1520 is applicable to the IIFP application in its entirety. 

In particular, NUREG-1520, Revision 0, Section 1.1.4.3, contains the acceptance criteria 

applicable to this portion of the review. 

1.1.3 Staff Review and Analysis 

The first section of Chapter 1 of the LA (IIFP, 2012a) provides a basic overview of the proposed 

IIFP facility with sufficient information to understand the process. The applicant plans to build 

the plant 24 kilometers (15 miles) west of Hobbs, NM, in Lea County. The facility will provide 

deconversion services to the uranium enrichment industry throughout the United States (U.S.) 

to convert depleted uranium hexafluoride (DUF6) into uranium oxide for disposal. In addition, 

the fluoride extracted in the deconversion of DUF6 will be used to produce specialty fluoride gas 

products, such as silicon tetrafluoride (SiF4), boron trifluoride (BF3), and anhydrous hydrogen 

fluoride (AHF). The plant is designed to process 3.3 million kilograms (kg) per year of DUF6, 

which will result in production of up to 0.68 million kg of SiF4, 0.23 million kg of BF3, and 

0.45 million kg of AHF annually. These products will be sold commercially for various industrial 

applications. 

The basic process involves receipt of the DUF6 via tractor trailers in 14-ton cylinders from any of 

the U.S. commercial enrichment facilities. The DUF6 is gasified in an autoclave. The DUF6 gas 
1-2 

then reacts with hydrogen gas to produce depleted uranium tetrafluoride (DUF4) and AHF. The 

DUF4 is transferred into a calciner with silicon dioxide or boron trioxide. As the calciner heats 

and mixes the materials, the mixture reacts to produce depleted uranium oxidize (DUO) powder 

and SiF4 or BF3 gas, depending on the process line. The DUO powder is collected in a disposal 

container for transportation to a licensed low-level waste disposal facility. The fluoride gases 

are purified, concentrated, and collected into storage cylinders for resale. These are stored 

temporarily, packaged for transportation, and sold commercially. 

The proposed facility is to be built on 40 acres of a 640-acre plot in a rural area of Lea 

County, NM. Figure 1-4 of the LA (IIFP, 2012a) provides an aerial photograph of the entire 

property, with the outline of the plant superimposed in the top left quadrant of the grid at its 

approximate location on the site. The terrain is primarily flat and underdeveloped, but overlays 

oil and gas geologic formations which are developed extensively in the area. Land use within 

5 miles of the facility consists primarily of cattle grazing and oil and gas production. The nearest 

industrial facility is a gas-fired power plant on the west boundary of the site, and the nearest 

resident lives 1.6 miles northwest of the facility. Chapter 1 of the ISA Summary (IIFP, 2012b) 

contains an expanded description of the facility. 

The IIFP facility, as described in the LA (IIFP, 2012a) and ISA Summary (IIFP, 2012b), consists 

of multiple buildings; each one of which performs a specific function. A list of the primary 

buildings and their specific functions is provided below in the order in which the material flows 

through the plant. 

• The DUF6 storage pad is used to store 14-ton cylinders full of DUF6, which arrive at the 

IIFP facility via tractor trailers. These cylinders are inspected; assayed; and, if accepted, 

offloaded to the storage pad. Shipments of silicon dioxide (SiO2), diboron trioxide 

(B2O3), and DUF4 are also received via tractor trailers and are surveyed and stored at 

separate locations. 

• The DUF6 autoclave building receives cylinders from the storage pad. They are placed 

via crane into steam autoclaves which vaporize the DUF6. The gaseous material is then 

piped into the process building. Empty uranium hexafluoride (UF6) cylinders are moved 

to a cold box for removing cylinder tails. 

• The DUF6 gas is piped to the DUF4 process building for conversion into DUF4 and AHF. 

Hydrogen gas is injected into a reaction vessel and interacts exothermically with the 

DUF6 to produce DUF4 powder and AHF gas. The DUF4 is collected in storage hoppers 

for transfer to the fluoride extraction process. The AHF gas is condensed and 

transferred to a storage tank. 

• The AHF staging containment building is used to store liquid AHF before shipment. The 

AHF is stored in 8,000-pound (lb) tanks, which are surrounded by dikes capable of 

holding the entire contents of a ruptured tank. The building is totally enclosed and 

contains leak detection and water spray systems to mitigate the airborne release of 

spills. 

• A DUF4 container staging building is located just outside the DUF4 process building. The 

building is used to store and organize additional inventory of DUF4 before deconversion. 

The building will also be used to receive external shipments of DUF4 from offsite 

sources. 
1-3 

• The fluoride extraction process (FEP) building is used to further deconvert the DUF4. 

The SiO2 or B2O3 is combined with the DUF4 and then heated and mixed in a calciner. 

The products consist of SiF4 or BF3 gas, respectively, and DUO powder. The DUO is 

collected and packaged in a container approved by the U.S. Department of 

Transportation (DOT) for shipment to a licensed low-level radioactive waste disposal 

site. The SiF4 or BF3 gas is condensed in a cold trap and transferred to a gas storage 

cylinder. 

• A number of storage, shipping, and environmental protection activities are conducted in 

ancillary buildings onsite. The LA, Section, 1.1.2.1 (IIFP, 2012a) describes these 

activities. Secondary facilities for the IIFP include data processing facilities, emergency 

response facilities, electrical distribution systems, security fencing and portals, a pump 

house, an air generation plant, a cooling tower, a boiler system, a training facility, a 

maintenance facility, storage facilities, and waste accountability facilities. 

The first chapter of the LA (IIFP, 2012a) and ISA Summary (IIFP, 2012b) provide additional 

descriptions, drawings, and process details for each of the areas described above. The IIFP 

facility is primarily a chemical facility which processes depleted uranium, among other 

operations. The chemical reactions involve the conversion of DUF6 to DUO. This process can 

be summarized as DUF6 reacting with hydrogen gas to produce DUF4 and AHF. The DUF4 is 

then chemically reacted with SiO2 or B2O3 to produce gaseous SiF4 or BF3, respectively, and 

DUO powder. These processes have a radiological component because of the natural 

radioactivity of uranium and daughter products. The facility does not process special nuclear 

material. 

The various fluoride and DUO streams result in residual offgases of hydrogen fluoride, SiF4, 

BF3, and other trace gas contaminants. The facility uses particulate filters and liquid scrubbers 

to minimize the release of effluents to the environment. These offgases are processed through 

a three-stage potassium hydroxide (KOH) scrubbing system. The system is designed to 

remove 99.9 percent of fluoride-bearing components before being vented to the plant stack. 

The effluents released from the stack are continuously monitored. Once the KOH scrubbing 

liquid has absorbed significant quantities of fluorides, the KOH is reacted with a lime slurry to 

regenerate the solution. The fluorides in the solution produce calcium fluoride which is 

collected, dried, and packaged for sale or disposal. The regenerated KOH is reused in the 

scrubbing system. 

The proposed facility produces both solid and liquid wastes. Solid waste consists of industrial, 

mixed (radioactive and chemical), and hazardous waste. These wastes are collected, volume 

reduced if feasible, and shipped to a licensed disposal facility. Liquid wastes containing 

uranium are transferred to the decontamination building to recover the uranium, if possible. 

Other process liquids, including sanitary sewage, are treated, sampled and used for 

landscaping—or are evaporated. The primary byproduct from the facility is DUO, which is 

shipped offsite to a licensed disposal facility. The LA, Section, 1.1.7 (IIFP, 2012a) provides a 

summary list of the raw materials, byproducts, wastes, and finished products involved in the 

process. 

The NRC staff reviewed the general description of the proposed facility and process using 

Section 1.1 of NUREG-1520, Revision 0 (NRC, 2002). 
1-4 

IIFP adequately described the general facility description according to Section 1.1 of the 

Standard Review Plan. IIFP adequately described (1) the facility and its processes so that the 

NRC staff has an overall understanding of the relationships of the facility features and 

(2) the function of each feature. IIFP cross-referenced its general description with the more 

detailed descriptions elsewhere in the application. The NRC staff also reviewed comparable 

sections in both the ISA Summary and the Emergency Management Plan and, based on that 

review, the NRC staff determined that the information is consistent with the information provided 

in the LA. Therefore, the NRC staff finds that IIFP has complied with the requirements of 

10 CFR 40.31, "Application for a specific license," and 10 CFR 70.65(b)(1-2), as applicable to 

this section. 

1.2 Institutional Information

The purpose of this review is to establish whether the LA includes adequate information 

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

1.2.1 Regulatory Requirements 

The regulations in 10 CFR 40.32(b) require the applicant to provide sufficient information to 

demonstrate that it is technically and financially qualified to operate the facility safely. 

Consistent with this requirement, the following institutional information is necessary: (1) the 

identity of the applicant; (2) the identity of the licensed material—including name, chemical and 

physical form—and maximum amount that will be possessed; and (3) the purpose for which the 

licensed material will be used. 

1.2.2 Regulatory Guidance and Acceptance Criteria 

Section 1.2.4 of NUREG-1520 (NRC, 2002), and the regulations in 10 CFR 40.32 contain the 

acceptance criteria applicable to NRC’s review of the institutional information section of the 

application. 

1.2.3 Staff Review and Analysis 

1.2.3.1 Corporate Identity 

In Section 1.2.1 of the LA (IIFP, 2012a), the applicant provided information on the corporate 

organization. IIFP is a wholly owned subsidiary of International Isotopes Inc. (INIS). INIS was 

originally organized as a Texas corporation in 1995, but has since moved operations to Idaho 

Falls, ID. In 2001, the company obtained a license under 10 CFR Part 30, “Rules of General 

Applicability to Domestic Licensing of Byproduct Material,” to manufacture high quality 

radiochemicals for medical, industrial, and research applications (Docket 030-35486 and 

License 11-27680-01). In 2005, INIS received a license under 10 CFR Part 40, “Domestic 

Licensing of Source Material,” to produce fluoride compounds from DUF4 (Docket 040-09058 

and License SUB-1587). In December 2009, INIS submitted the application for the proposed 

IIFP facility in New Mexico. Its wholly owned subsidiaries are International Isotopes Idaho, Inc.; 

IIFP; and International Isotopes Transportation Services, Inc., all of which are Idaho 

corporations. 

The IIFP’s principal office is located at 4137 Commerce Circle, Idaho Falls, ID 83401. The 

facility is planned to be built in Lea County, NM, 24 kilometers (km) [15 miles (mi)] west of 

Hobbs, NM, and approximately 17 mi (27 km) west of the Texas state border. The 
1-5 

LA, Section, 1.2.1 (IIFP, 2012a) provide additional information on the IIFP corporate identity. 

This description meets the guidance in Section 1.2.4.3(1) of NUREG-1520 (NRC, 2002), and is, 

therefore, acceptable. 

1.2.3.2 Financial Qualifications 

Section 1.2.2 of the LA (IIFP, 2012a) states that the estimated project cost to build and 

operate the facility is $75–90 million dollars in 2009 dollars. IIFP plans to obtain these 

funds from a number of capital investors. In addition, a surety bond is planned for 

financial assurance for decommissioning. The LA, Chapter 10 (IIFP, 2012a), provides 

additional information on the Decommissioning Funding Plan (DFP), which is reviewed 

separately in Chapter 10 of this safety evaluation report (SER). The information 

provided by the applicant meets the guidance in Section 1.2.4.3(2) of NUREG-1520 

(NRC, 2002), and is, therefore, acceptable. 

1.2.3.3 Type, Quantity, and Form of Licensed Material 

Source material makes up the vast majority of licensed material at the IIFP. Table 1-4 of the LA 

(IIFP, 2012a) lists the types, quantities, and forms of the licensed material. The maximum 

uranium onsite at any one time is limited to 1,650,000 lbs (750,000 kg) of depleted uranium in 

any chemical form. The source material will be in the chemical form of UF6, UF4, uranium 

oxides, and other trace uranium compounds. All of the process material will consist of DUF6. 

IIFP only uses commercial enrichment plant suppliers and does not accept cylinders that 

contain technetium or transuranics, contaminants from facilities related to the U.S. Department 

of Energy (DOE). 

During shipment from the enrichment facilities and storage onsite, the DUF6 is in a solid 

crystalline form. In the autoclaves, a percentage of the DUF6 may become liquefied during 

gasification, but it is in the gaseous state during processing. The DUF4, both received from 

other facilities and produced onsite is a powder during shipment, storage, and processing. The 

waste products, depleted uranium dioxide (DUO2), and depleted triuranium octoxide (DU3O8), 

are powders which are collected in containers for offsite disposal. Precautions, including 

ventilation and personal protective equipment, are in place to protect against small quantities of 

depleted uranyl fluoride (DUO2F2) formed by the reaction of DUF6 with moisture in the air. 

Waste liquids may contain this chemical before uranium recovery. 

The applicant identified the maximum quantity and chemical and physical forms of the licensed 

material present at the facility. The information provided by the applicant is consistent with 

Section 1.2.4.3(3) of NUREG-1520 (NRC, 2002), and the regulatory requirements and is, 

therefore, acceptable. 

1.2.3.4 Authorized Uses 

The application is for the issuance of a license under 10 CFR Part 40. As indicated above, IIFP 

does not store or process special nuclear material. 

If granted, the proposed source material license would authorize the deconversion of DUF6 into 

oxides. This process would involve the recovery of the fluoride products for commercial resale 

and the disposal of the oxides as waste for licensed disposal. The deconversion services would 

be marketed to enrichment facilities as a means for them to dispose of depleted uranium. 
1-6 

The applicant has requested a 40-year license term. The maximum extent allowed for license 

terms are not prescribed by regulation, but are determined by Commission policy. Previously, 

the Commission approved, in the NRC staff requirements memorandum to SECY-06-0186, 

“Increasing Licensing Terms for Certain Fuel Cycle Facilities,” the issuance of up to a 40-year 

license for facilities which implement an ISA, consistent with the requirements in 

10 CFR Part 70, “Domestic Licensing of Special Nuclear Materials,” Subpart H, “Additional 

Requirements for Certain Licensees Authorized To Possess a Critical Mass of Special Nuclear 

Material.” The ISA Summary serves as one of the primary safety bases for approval of the 

application, and it must be updated annually. The management measures and change 

processes required for the ISA mitigate the aging phenomena, thereby allowing longer license 

terms of up to 40 years. The DFP is also required to be updated every 3 years or as stipulated 

in the license. Additional information on the decommissioning review is available in Chapter 10. 

IIFP is implementing the 10 CFR Part 70, Subpart H, requirements until similar requirements are 

included in 10 CFR Part 40, Subpart H, which is being developed via rulemaking at the time of 

this review. The NRC staff has concluded that the ISA requirements for IIFP are consistent with 

the Commission’s approval for up to a 40-year license term. 

As stated above, the applicant provided a summary—nontechnical, narrative description—for 

each activity or process in which it proposed to acquire, deliver, receive, possess, produce, use, 

process, transfer, or store licensed material. The authorized uses of licensed material proposed 

for the facility are described and are consistent with the Atomic Energy Act of 1954, et seq. The 

description is also consistent with more detailed process descriptions submitted as part of the 

ISA Summary (IIFP, 2012b), as reviewed in Chapter 3 of this SER. The information provided by 

the applicant meets the guidance in Section 1.2.4.3(4) of NUREG-1520 (NRC, 2002), and is, 

therefore, acceptable. 

1.2.3.5 Special Exemptions or Special Authorizations 

1.2.3.5.1 Use of ICRP 68 and 72 DAC and ALl Values 

In Chapter 1, Section 1.5, of the LA (IIFP, 2012a), IIFP requested an exemption from the 

derived air concentration (DAC) and annual limit on intake (ALI) values provided in 

10 CFR Part 20, Appendix B. In their place, the exemption requested authorization to use the 

International Committee on Radiological Protection (ICRP) Publication 68, entitled “Dose 

Coefficients for Intake of Radionuclides by Workers” (ICRP, 1994c) for determining dose to 

workers and ICRP Publication 72, entitled “Age-dependent Doses to the Members of the Public 

from Intake of Radionuclides Part 5, Compilation of Ingestion and Inhalation Coefficients” 

(ICRP, 1996) for determining dose to the public due to effluents. 

The purpose of this request is to allow for the use of updated radiation dose conversion factors 

for DAC and ALI values calculated using the internal dosimetry models described in ICRP 

Publications 68 and 72. The current NRC regulations governing airborne radioactivity are 

derived from the inhalation dose conversion factors (DCF) in ICRP Publications 26 (ICRP, 1977) 

and 30 (ICRP, 1982), as disseminated by the U.S. Environmental Protection Agency in the 

Federal Guidance Report 11. Subsequent work by the ICRP refined the dosimetry models used 

to derive ICRP Publications 26 and 30. ICRP Publications 68 and 72 contain improved dose 

conversion factors based on updated scientific data, but the derived documents for 10 CFR 20 

are still based on the outdated ICRP 26 and 30 data. This exemption request proposes to 

replace the use of ICRP 26 and 30 data with the updated ICRP 68 and 72 data as applicable for 

internal dose calculations, determining the allowed limits for intakes, and calculating derived air 

concentration and effluent concentration limits. 
1-7 

The Commission has determined that requests to use ICRP 68 and 72 must be considered on a 

case-specific basis as exemptions. On April 21, 1999 (SRM-SECY-99-077), the Commission 

approved the NRC staff’s granting of exemptions for the use of ICRP 68 and 72 on a case-bycase basis based on the precedent set by the Commission’s decision in the OSRAM, Inc.’s, 

exemption request (SECY-99-077). In addition, the majority of fuel fabrication facilities and 

several source material facilities have been granted this exemption. 

In a rulemaking to 10 CFR 20, finalized May 21, 1991 (56 FR 23391), the NRC revised the 

regulation of radioactive materials so that it would be based on dose instead of intake. A 

consequence of the dose-based rule is that compliance would not necessarily be constrained by 

use of a specific set of parameters to calculate dose. Title 10 CFR Part 20, in fact, allows 

certain adjustments to be made to the model parameters if specific information is available, 

such as adjustments when the particle size of the airborne radioactive material is known, rather 

than using a default particle size. However, because of the way Part 20 was written, NRC 

licensees are not permitted to use the revised and updated internal dosimetry models, unless 

permission is specifically granted in the license. 

In the same year that the Commission approved the final 10 CFR Part 20 rule based on 

ICRP 26 and 30, ICRP published a major revision of its radiation protection recommendations in 

ICRP 60. In the several years following the revision of ICRP 60, ICRP published a series of 

reports in which it described the components of an extensive updated and revised internal 

dosimetry model. These reports include ICRP Publications 60 (ICRP, 1991), 68 (ICRP, 1994a), 

66 (ICRP, 1994b), 71, (1995a), 72 (ICRP, 1996) and 78 (ICRP, 1997). 

Although the dose per unit intake calculated using the new models (ICRP 68 and 72) does not 

differ by more than a factor of about two from the values in 10 CFR Part 20 for most 

radionuclides, the differences are substantial for some, particularly for the isotopes of thorium 

(Th), uranium (U), and some of the transuranic nuclides. Dose calculations using the DAC and 

ALI values from ICRP 26 and 30 (i.e., 10 CFR Part 20, Appendix B) are several times higher 

than real life; therefore, 10 CFR Part 20 requires significantly more protective measures for Th 

and U than would be warranted based on the updated models in ICRP 68 and 72. This is IIFP’s 

primary concern, and they have requested to be allowed to use DAC and ALI values based on 

the coefficients from ICRP 68 and 72. Through NRC staff reviews of IIFP’s Radiation Safety 

Protection program, the NRC staff concludes that the program is sufficiently sophisticated by 

training and expertise to utilize the ICRP models in a manner equivalent to those listed in 

10 CFR 20, Subpart C and D, i.e., IIFP’s program is capable of properly calculating the total 

effective dose equivalent (TEDE) for its occupational workers and the public using ICRPs 68 

and 72. 

IIFP’s use of ICRP 68 and 72 in place of existing values in Appendix B to Part 20 gives its 

workers and the public equivalent radiological protection. Thus, the exemption is authorized by 

law and will not result in undue hazard to life or property. 

The NRC staff has developed an Environmental Impact Statement (EIS) (NRC, 2011) separate 

from this SER which includes this exemption request. The exemption is described in the 

Environmental Report (ER) (IIFP, 2011a), which states that ICRP 68 and 72 are used for 

calculating dose estimates. 

The NRC staff finds that IIFP’s request for an exemption is acceptable under 10 CFR 20.2301 

and 10 CFR 40.14, which state that the Commission may—upon application by a licensee or 
1-8 

upon its own initiative—grant an exemption from the requirements of the regulations if the NRC 

determines the exemption is authorized by law and would not result in undue hazard to life or 

property. This determination is based on the NRC staff’s safety evaluation, as described above, 

and the EIS (NRC, 2011) conducted for this application. 

Consistent with the 10 CFR 2301 and 10 CFR 40.14, IIFP is granted an exemption from DAC 

and ALI values in 10 CFR 20.1003 and 10 CFR Part 20, Appendix B. This exemption 

authorizes IIFP to use the ALIs and DACs based upon dose coefficients published in the ICRP 

Publication 68, entitled “Dose Coefficients for Intake of Radionuclides by Workers,” (ICRP, 

1994a) for determining dose to workers. The exemption also authorizes IIFP to use ICRP 

Publication 72, entitled “Age Dependent Doses to the Members of the Public from Intake of 

Radionuclides Part 5, Compilation of Ingestion and Inhalation Coefficients,” (ICRP, 1996) for 

determining dose to the public due to effluents. 

This exemption does not change, in any way, the NRC dose limits which the applicant must 

maintain. Rather, the exemption changes the methodology IIFP uses to calculate the doses, 

and assures the calculations are based on updated and internationally recognized standards 

published in the ICRP 68 and 72, as described above. 

1.2.3.5.2 Memorandum of Understanding between IIFP and New Mexico Environmental 

Department 

On October 22, 2009, IIFP finalized a memorandum of agreement (MOA) with the State of New 

Mexico Environment Department (NMED) to limit the maximum quantities of depleted uranium 

onsite (IIFP, 2009). The limits established in the MOA (4,850,169 lbs [2,200,000 kg] of uranium) 

are significantly higher than the possession limits requested in the LA (1,653,500 lbs 

[750,000 kg] of uranium of any form) (IIFP, 2012a). The uranium possession limit requested in 

the LA represents approximately 60 full cylinders of DUF6. The NRC staff finds this possession 

limit to be reasonable for the types of operations presented in the LA (IIFP, 2012a), which 

include DUF6 storage before processing, uranium within the process, and DUO awaiting 

dispatch. The IIFP/NMED agreement is not binding on the NRC and represents an independent 

agreement between IIFP and NMED. 

1.2.3.5.3 Financial Assurance Exemption 

In Section 1.6 of the LA (IIFP, 2012a), the applicant requested an exemption from the 

requirements of 10 CFR 40.36(d) to provide an alternate timeframe for implementing the 

financial instruments for decommissioning. 

The NRC requirements for financial assurance are described in 10 CFR 40.36(d) and include a 

requirement that a licensee certify that financial assurance has been provided in the amount of 

the cost estimate for decommissioning at the time of licensing. IIFP’s cost estimate for 

decommissioning is based on a 40-year operating life and includes the estimated 

decommissioning costs of the facility and site, and the expected costs associated with the 

disposition of the depleted uranium onsite. The proposed exemption would allow IIFP to submit 

its financial instrument for decommissioning to the NRC for review 6 months before the startup 

of operations and fully fund the instrument at least 21 days before operations. Because of the 

time required for construction, the alternate schedule proposed in the exemption request for 

finalizing the financial assurance funding instrument differs significantly from the schedule 

required by 10 CFR 40.36(d) (i.e., full funding at the time the license is issued). 
1-9 

IIFP has requested the exemption on the basis that no significant quantity of licensed material 

will be onsite before the financial assurance funding instrument is approved by the NRC or 

funded by IIFP (IIFP, 2012a). Therefore, the cost of maintaining the financial assurance funding 

instrument for an extended period before operations, when licensed material is not onsite, would 

represent a significant financial burden to the applicant. Submittal of the financial assurance 

funding instrument 6 months before operations would allow sufficient time for the NRC to verify 

the adequacy of the instrument. Fully funding the instrument 21 days before operations would 

ensure that the funds for decommissioning are in place before operations. Thus, the requested 

exemption would allow IIFP to satisfy the applicable decommissioning funding assurance 

requirements without imposing an unnecessary financial burden. 

Under the provisions of 10 CFR 40.14, “Specific Exemptions,” the Commission may, upon 

application by any interested person or upon its own initiative, grant exemptions from the 

requirements of 10 CFR Parts 40, respectively, when the exemptions are authorized by law, will 

not endanger life or property or the common defense and security, and are otherwise in the 

interest of the public. 

The NRC staff evaluated the exemption request. The proposed exemption is authorized by law 

because the Atomic Energy Act of 1954, as amended, contains no provisions prohibiting an 

applicant from providing decommissioning funding commensurate with the quantity of licensed 

material onsite. The NRC staff also determined that, because the funding will be reviewed by 

the NRC and will be provided before receipt of licensed material, the approach will not endanger 

life or property or the common defense and security. Reducing the time and cost associated 

with maintaining the financial assurance before operations has no negative impact on 

decommissioning. Therefore, the NRC staff has determined that the proposed approach will be 

in the public’s interest by reducing unnecessary regulatory costs. 

The exemption granted from 10 CFR 40.36(d) financial assurance requirements meets the 

criteria for categorical exclusion under 10 CFR 51.22(c)(25), “Criterion for Categorical 

Exclusion; Identification of Licensing and Regulatory Actions Eligible for Categorical Exclusion 

or Otherwise Not Requiring Environmental Review.” The granting of this exemption from 

regulatory requirements meets the following criteria for categorical exclusion described in 

10 CFR 51.22(c)(25): 

(i) "There is no significant hazards consideration; 

(ii) There is no significant change in the types or significant increase in the amounts of 

any effluents that may be released offsite; 

(iii) There is no significant increase in individual or cumulative public or occupational 

radiation exposure; 

(iv) There is no significant construction impact; 

(v) There is no significant increase in the potential for or consequences from radiological 

accidents; and 

(vi) The requirements from which an exemption is sought involve: 

.... 

(H) Surety, insurance or indemnity requirements; ...” 

Therefore, no Environmental Assessment or EIS is required for this action. 

The NRC staff grants the requested exemption. The license will include a license condition that 

will address IIFP’s commitments to submit financial assurance documents for review and 

approval. Chapter 10 of this SER discusses this license condition further. 
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1.2.3.5.4 License Conditions for Construction and Operational Readiness 

The NRC will implement the following license conditions into the IIFP license to allow NRC staff 

to conduct routine inspections during the construction of the IIFP facility and evaluate the 

Operational Readiness Review (ORR). The inspectors will verify that IIFP is constructing the 

facility consistent with internal written procedures, commitments in the application, and relevant 

regulations. The second license condition will require IIFP to conduct an ORR. IIFP will be 

required to summarize the ORR in a report or series of reports and provide the report(s) to the 

NRC for review at least 60 days prior to operations. The applicant reviewed each of the 

following license conditions and agrees to their requirements. 

Construction License Condition 

During construction of the IIFP facility, the licensee shall afford to the NRC at all 

reasonable times, opportunity to inspect the premises and facilities and to verify that the 

facility has been constructed in accordance with the requirements of the license and 

onsite implementing procedures. 

Operational Readiness Review License Condition 

In addition, prior to commencing operations of the deconversion facility, IIFP shall 

conduct an operational readiness and management measures verification review and 

document the results in a report(s) to the NRC. The review must verify and document 

that IROFS and their management measures have been implemented to ensure 

compliance with the performance requirements of the integrated safety analysis. IIFP 

shall provide the NRC with a copy of the ORR report(s). The report(s) may address the 

entire facility as a whole or be submitted in parts. Each part must address major 

subsections of the facility with clearly defined boundaries, e.g., the uranium tetrafluoride 

building. The report(s) must be provided to the NRC no later than 60 days in advance of 

the start of commercial operations. (i.e., excluding testing) for that section of the facility. 

IIFP shall provide NRC opportunity to verify, through review and inspection, the 

report’s(s’) findings. 

1.2.3.6 Security of Classified Information 

All processes, materials, and information at the IIFP are unclassified. Therefore, 

10 CFR Part 95 is not applicable as stated in the acceptance criteria in NUREG-1520, 

Section 1.2.4.3 (NRC, 2002). Security-related information will be controlled consistent with 

internal written procedures. 

The NRC staff reviewed the institutional information for the proposed IIFP according to 

Section 1.2 of NUREG-1520 (NRC, 2002). On the basis of the review, the NRC staff has 

determined that the applicant has adequately described and documented the corporate 

structure and financial information, and is in compliance with those parts of 10 CFR 40.31 and 

70.65 related to other institutional information. In addition, the applicant has adequately 

described the types, forms, quantities, and proposed authorized uses of licensed materials to be 

permitted at this facility: 

 
1-11 

Table 1-1 Types, Forms, Quantities, and Proposed Authorized Uses 

Of Licensed Materials 

Material Form Quantity Authorized Use(s) 

Uranium (depleted) 

and daughter 

products 

Physical: solid, liquid, 

and gas Chemical: 

UF6, UF4, UO2F2, 

oxides, and other 

compounds 

750,000 Kilograms as 

uranium 

IIFP will be authorized 

to acquire, deliver, 

receive, possess, 

produce, use, 

transfer, and/or store 

source material 

consistent with their 

license. 

The applicant’s proposed activities are consistent with the Atomic Energy Act of 1954, as 

amended. The applicant has provided all institutional information necessary to understand the 

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

handled in connection with the requested license. 

1.3 Site Description

The purpose of the NRC’s review of the applicant's site description is to evaluate whether the 

application adequately describes the geographic, demographic, meteorological, hydrologic, 

geologic, and seismologic characteristics of the site and the surrounding area. The site 

description summarizes the information that the applicant used in preparing the LA, ER, 

Emergency Plan, and ISA Summary. 

1.3.1 Regulatory Requirements 

The regulations in 10 CFR 40.32, and 10 CFR 70.65(b)(1) require each application to include a 

general description of the site, with emphasis on those factors that could affect safety (i.e., 

nearby facilities, meteorology, and seismology). In addition, 10 CFR 70.61(f) requires each 

licensee to establish a controlled area, as defined in 10 CFR 20.1003, “Definitions,” and to 

retain authority to exclude or remove personnel and property from the area. 

1.3.2 Regulatory Guidance and Acceptance Criteria 

The acceptance criteria applicable to NRC's review of the site description section of the LA and 

ISA Summary (IIFP, 2012a, b) are contained in 10 CFR 40.32, 10 CFR 70.65(b)(1), and 

Section 1.3.4.3 of NUREG-1520, Revision 0 (NRC, 2002). Chapter 1 of NUREG-1520, 

Revision 0 (NRC, 2002), applies to the IIFP facility in its entirety. 

1.3.3 Staff Review and Analysis 

1.3.3.1

1.3.3.2 Site Geography 

The site location of the proposed facility is described in Section 1.7.1.1 of the LA (IIFP, 2012a) 

and Section 1.1.1 of the ISA Summary (IIFP, 2012b). These sections indicate that the proposed 

site is in Lea County, New Mexico, approximately 19 km (12 mi) west of Hobbs, New Mexico 

and 26 km (16 mi) west of the Texas border. The IIFP facility lies approximately 1.6 km (1.0 mi) 

from U.S. Highway 62/180 and along the east side of New Mexico Highway 483. 
1-12 

The proposed site has an average elevation of approximately 1,158 m (3,800 ft) higher than the 

Gulf of Mexico and a relatively flat topology. The IIFP site is currently an open, vacant land. 

The only agricultural activity in the site vicinity is domestic livestock ranching. 

1.3.3.2.1 Nearby Highways and Railroads 

The applicant discussed the highways located near the proposed IIFP facilities in the ISA 

Summary, Section 1.2.4.1 (IIFP, 2012b). According to the applicant, the IIFP site is located 

north of U.S. Highway 62/180 and east of New Mexico Highway 483. New Mexico Highway 483 

is the nearest transportation route to the IIFP safety structure, more than 0.94 km (0.58 mi) 

away. The nearest railroad is more than 8 km (5 mi) away from the IIFP facility. 

The applicant stated that the IIFP structures will be designed to withstand an overpressure of 

6.9 kilopascals (kPa) (1 pound per square inch [psi]). With this design basis, the applicant 

estimated that the safe separation distance for an explosion not to affect an IIFP safety 

significant structure is approximately 0.4 km (0.24 mi), assuming a postulated explosion of 

4,536 kg (10,000 lbs) of propane. The applicant also calculated the safe separation distance of 

0.5 km (0.3 mi) from an explosion at a transportation route using the maximum probable 

hazardous solid cargo for a single highway truck suggested in Regulatory Guide 1.91, 

Revision 1, “Evaluations of Explosions Postulated to Occur on Transportation Routes Near 

Nuclear Power Plants,” issued February 1978 (NRC, 1978), assuming this solid is subject to a 

vapor effective explosion (IIFP, 2012a). In addition, the applicant estimated the safe separation 

distance of 0.94 km (0.58 mi) from a rail transportation accident using the maximum explosive 

cargo in a single railroad box car and assuming a vapor effective explosion for the cargo. As 

discussed earlier, the nearest highway to any IIFP safety-significant structure is more than 

0.94 km (0.58 mi) away, and the nearest railroad is more than 8 km (5 mi) away from the IIFP 

facility. Because these distances exceed the corresponding safe separation distance the 

applicant estimated, IIFP concluded that potential explosion hazards from nearby transportation 

routes will not pose safety concerns to the IIFP site. 

The NRC staff reviewed the information the applicant provided regarding the potential hazards 

caused by explosions at nearby highways and railroads. The NRC staff finds that the applicant 

used the approach recommended in Regulatory Guide 1.91, Revision 1 (NRC, 1978), to 

estimate the safe separation distance from the effects of explosions of hazardous materials that 

may be carried on nearby transportation routes. Regulatory Guide 1.91, Revision 1 

(NRC, 1978), provides an acceptable approach and was used for licensing nuclear-related 

facilities in the past. Therefore, the NRC staff finds that the applicant’s assessment of explosion 

effects from nearby highways and railroads using the methodology in Regulatory Guide 1.91. 

Revision 1 (NRC, 1978), is acceptable. In addition, the NRC staff accepts the applicant’s 

conclusion that nearby transportation route explosion accidents do not require further 

consideration because the distance between the nearest highways and railroads and the IIFP 

facility site is greater than the corresponding safe separation distance. Based on this review, 

the NRC staff finds that the explosion accidents from nearby transportation routes are not a 

safety concern to the IIFP facility. Therefore, the applicant meets the regulatory requirements in 

10 CFR 70.65(b)(1). 

1.3.3.2.2 Nearby Industrial Facilities 

The applicant discussed the potential hazards associated with the nearby industrial facilities to 

the proposed IIFP facilities in the ISA Summary, Section 1.2.4 (IIFP, 2012b). Three gas-fueled, 

electric-generating power plants and a gas-processing facility are located near the proposed 
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IIFP site: (1) Xcel Energy Cunningham Station has four units 1.6 km (1.0 mi) west of the site, 

(2) Xcel Energy Maddox Station is 3.5 km (2.2 mi) east-southeast of the site, (3) Colorado 

Energy Hobbs Generating Station is 3.1 km (1.9 mi) east-northeast of the site, and (4) DCP 

Midstream Linam Ranch Plant is 5.8 km (3.16 mi) southeast of the site (IIFP, 2012b). 

The Xcel Energy Cunningham Station consists of two steam-electric generating units and two 

natural gas combustion turbine units. The Xcel Energy Maddox Station consists of a natural 

gas-fired, steam-electric generating unit and two natural gas combustion turbine units. The 

Colorado Energy Hobbs Generating Station uses a combined-cycle technology, including both a 

gas combustion turbine and a steam-driven turbine. DCP Midstream Linam Ranch Plant is a 

natural gas processing facility. 

The applicant pointed out that the potential hazards to the proposed IIFP facility from these 

industrial facilities are facility explosions caused by the natural gas at the facilities themselves 

and an aircraft crash into the facilities, facility fires, and power outages (IIFP, 2012b). The 

applicant considered the fire and power outage events resulting from the nearby industrial 

facilities in its ISA, and suitable measures and controls are identified to alleviate the potential 

impact (IIFP, 2012b). The review of the fire protection measures are addressed further in 

Chapter 7, “Fire Safety,” and the review of the electrical systems are addressed in Appendix D, 

“Electrical Systems” of this SER. 

The applicant stated that the explosions caused by an aircraft crash into a nearby industrial 

facility would be highly unlikely events. The applicant made the assessment based on the 

analysis of a potential aircraft crash into the IIFP facility. Even if an aircraft were to crash into a 

nearby facility, the induced explosion would not be a close enough distance to damage the IIFP 

facility. Based on its analysis of a natural gas pipeline explosion at the Cunningham Power 

Station (closest to the IIFP facility compared to other nearby facilities), the applicant determined 

that the natural gas explosion would not affect the operation’s safety of the IIFP facility because 

of the distance between these facilities and the proposed IIFP facility. Consequently, the 

applicant concluded that no industrial facilities are close enough to the proposed facility site to 

pose a safety concern regarding explosions to the IIFP facility operations. 

The NRC staff reviewed the information regarding the nearby industrial facilities and finds that 

the applicant provided sufficient information to support evaluation of potential hazards 

associated with the nearby industrial facilities. The NRC staff finds that the applicant’s 

conclusion that explosion-related hazards do not pose a safety concern to the proposed facility 

is acceptable because these facilities are located at a substantial distance from the IIFP facility, 

avoiding any appreciable damage, according to Regulatory Guide 1.91, Revision 1 

(NRC, 1978). The NRC staff also finds that the applicant adequately considered the nearby 

facility-related fire and power outage accident events in its ISA. Based on this review, the NRC 

staff concludes that the applicant has met the regulatory requirements in 10 CFR 70.65(b)(1). 

1.3.3.2.3 Nearby Gas Pipelines 

The applicant discussed the nearby gas pipeline hazards in the ISA Summary, Section 1.2.4.2 

(IIFP, 2012b). The applicant stated that, based on the easement records of Lea County, one 

petroleum gas pipeline and 12 natural gas pipelines are located within 1.6 km (1 mi) of the 

proposed IIFP site. The pipeline diameter and gas pressure data are known for all but two 

pipelines. Among these pipelines, the largest pipeline diameter is 30.5 cm (12 in), and the 

largest gas pressure is 10 megapascals (MPa) (1,500 psi). IIFP verified that these values are 

conservative and bound the diameter and pressures of the unknown piblines (IIFP, 2012b). 
1-14 

The applicant determined the design basis for external explosion-induced overpressure caused 

by dynamic effects to be 6.9 kPa (1 psi) for the IIFP process building. The applicant’s approach 

to assessing the potential hazards of a gas pipeline explosion was to determine the annual 

probability of a nearby gas pipeline explosion (following rupture of the pipelines and subsequent 

detonation) that could generate an overpressure greater than the design-basis, blast-induced 

overpressure for the IIFP process building using the approach suggested in Regulatory 

Guide 1.91, Revision 1 (NRC, 1978). The approach in Regulatory Guide 1.91, Revision 1 

(NRC, 1978), requires the following information: (1) explosion rate of gas pipelines in question 

in explosions per mile, (2) number of gas pipelines, and (3) exposure distance in miles. In 

conducting the analysis, the applicant estimated the explosion rates per year per pipeline mile 

for natural gas pipelines and petroleum gas pipelines using the gas pipeline safety data from the 

DOT Web site (“Data & Statistics” at http://phmsa.dot.gov/pipeline/library/data-stats). 

The applicant estimated the annual probabilities for explosions of the natural and petroleum gas 

pipelines to be less than 3×10−6

. Therefore, the applicant concluded that gas pipeline explosion 

accidents are highly unlikely near the proposed IIFP facility. 

The NRC staff reviewed the discussions the applicant provided on gas pipeline hazards near 

the proposed IIFP facility and finds that the applicant used reliable DOT data to estimate the gas 

pipeline explosion rates. The NRC staff finds that the applicant’s assumptions used to 

determine the average release rate, produce a conservative result. The NRC staff also finds 

that use of the Area Locations of Hazardous Atmospheres (ALOHA) computer code to calculate 

blast radius of a pipeline explosion is acceptable because the ALOHA computer code, which the 

U.S. Environmental Protection Agency endorses, is commonly used by the industry. In addition, 

the NRC staff finds that the applicant estimated the exposure distance of a gas pipeline 

consistent with that defined in Regulatory Guide 1.91, Revision 1 (NRC, 1978). Furthermore, 

the NRC staff finds that the applicant’s approach of estimating annual weighted exposure 

distance for each pipeline using the annual frequency of the atmospheric conditions in the area 

is acceptable because the annual frequency of the atmospheric conditions was determined 

based on local specific data (IIFP, 2012b). Therefore, the NRC staff concludes that the 

applicant acceptably determined the annual probability of a gas pipeline explosion that could 

result in an overpressure greater than the design basis for a process building and that nearby 

gas pipelines are not a safety concern to the proposed IIFP facility. Based on this review, the 

NRC staff concludes that the applicant has met the regulatory requirements in 

10 CFR 70.65(b)(1). 

1.3.3.2.4 Air Transportation 

The applicant discussed aircraft crash hazards to the proposed IIFP facility in the ISA Summary, 

Section 1.2.4.3 (IIFP, 2012b). Seventeen airports are located within 161 km (100 mi), but more 

than 8 km (5 mi) away from, the IIFP site. Among these airports, Lea County Regional Airport, 

Hobbs Industrial Airpark, and Lea County Zip Franklin Memorial Airport are within 32 km (20 mi) 

of the IIFP site. The Lea County Regional Airport is located approximately 13 km (8 mi) 

east-southeast, the Hobbs Industrial Airpark is 13.7 km (8.5 mi) east-northeast, and the Lea 

County Zip Franklin Memorial Airport is 27 km (17 mi) north-northwest of the proposed IIFP site. 

There are four military routes within a 48 km (30 mi) radius of the proposed IIFP site, with the 

closest approach approximately 27 km (17 mi) southwest of the IIFP site (IIFP, 2012b). The 

number of military operations at the Lea County Regional Airport is 561 annually. In addition, a 
1-15 

special-use airspace for two military operations areas (MOAs) are located north of the IIFP site. 

The closest edge of the MOA is approximately 9.3 km (5.8 mi) from the IIFP facility. 

In assessing the probability of aircraft crash hazards, the applicant (IIFP, 2012b) used the first 

proximity acceptance criterion provided in NUREG-0800, “Standard Review Plan for the Review 

of Safety Analysis Reports for Nuclear Power Plants: LWR Edition,” Section 3.5.1.6, “Aircraft 

Hazards” (NRC, 2010). NUREG-0800 provides three proximity acceptance criteria. If all three 

criteria are met, the probability of aircraft accidents with potential radiological consequences is 

considered to be less than about 1×10−7 per year. The first proximity criterion states that, for the 

aircraft crash hazards related to the airport operations to be insignificant to the site, the annual 

number of operations for the airports with a distance (1) between 8 and 16 km (5 and 10 mi) 

from the site of interest should be less than 500×D2

 and (2) more than 16 km (10 mi) from the 

site should be less than 1,000×D2

 (NRC, 2010). Note that D is the distance between the airport 

of interest and the IIFP site. In assessing the probability of hazards associated with airport 

operations near the IIFP facilities, the applicant used the first proximity acceptance criterion. By 

comparing the annual number of operations with the first proximity acceptance criterion, the 

applicant determined that the annual number of operations for all of the 17 airports located 

within 161 km (100 mi) of the IIFP site is less than the maximum aircraft activities stipulated in 

the first proximity acceptance criterion (IIFP, 2012b). Therefore, the applicant concluded that, 

consistent with NUREG-0800 (NRC, 2010), the probability of aircraft accidents with potential 

radiological consequences is considered to be less than 1×10−7

 per year. 

So that military-related aircraft activities do not affect a site of interest, the second proximity 

acceptance criterion states that the facility of interest should be at least 8 km (5 mi) from the 

nearest edge of military training routes, including low-level training routes. Facilities outside this 

range may still be affected if there are greater than 1,000 flights per year or where activities 

(such as practice bombing) may create an unusual stress situation (NRC, 2010). As the 

applicant indicated (IIFP, 2012b), the nearest military route is approximately 27 km (17 mi) 

southwest, and the closest edge of the MOA is approximately 9.3 km (5.8 mi) from the IIFP site. 

Therefore, the applicant determined that the activities associated with the military training route 

and the MOA are not a safety concern to the IIFP site. In addition, the applicant indicated that 

the number of military operations (561 annually) at the Lea County Regional Airport, which is 

13 km (8 mi) east-southeast of the IIFP site, is less than 1,000 flights per year as specified in 

the second proximity acceptance criterion. Consequently, the applicant concluded that military 

operations are not a safety concern to the IIFP facility. 

The third proximity acceptance criterion in NUREG-0800, Section 3.5.1.6 (NRC, 2010), states 

that the facility of interest should be at least 3 km (2 mi) beyond the nearest edge of a Federal 

airway, holding pattern, or approach pattern. The applicant indicated that four high-level 

airways en route are within 56 km (35 mi) of the proposed IIFP facility, with the closest airway 

(Q20) passing at 16.7 km (10.4 mi) from the IIFP facility (IIFP, 2012b). The applicant further 

indicated that three low-level airways en route pass through the nearby air space, with the 

closest airway (V68) passing at 5.1 km (3.2 mi) of the IIFP facility. Based on the information 

regarding the airway distances from the IIFP facility, the applicant determined that all of the 

Federal airways meet the third proximity acceptance criterion. For further confirmation, the 

applicant conducted additional analysis to calculate the annual probability of an aircraft on the 

V68 airway crashing onto the proposed IIFP facility area using the methods from both 

NUREG-0800, Section 3.5.1.6 (NRC, 2010), and DOE-STD-3014-2006, “Accident Analysis for 

Aircraft Crash into Hazardous Facilities” (DOE, 2006). The annual aircraft crash probability is 

2.7×10−8

 using the NUREG-0800 method and 3.3×10−7 using the DOE method. Based on these 

probabilities, the applicant determined that an aircraft crash from Federal airways onto the IIFP 
1-16 

facility is highly unlikely. Therefore, the applicant concluded that the nearby Federal airways are 

not a safety concern to the IIFP facility. 

For the holding and approach patterns, the applicant evaluated air traffic at the Lea County 

Regional Airport, Hobbs Industrial Airpark, and Lea County Zip Franklin Memorial Airport, which 

are within 32 km (20 mi) of the IIFP site (IIFP, 2012b). The applicant stated that Lea County 

Regional Airport has seven instrument flight rule and two visual flight rule procedures for holding 

and approach patterns, and all holding and approach patterns are more than 3 km (2 mi) from 

the IIFP site. The closest landing and takeoff would be 12.1 km (7.5 mi) from the site. The 

applicant indicated that the Hobbs Industrial Airpark has no air carrier, general aviation, or 

military operations. The only operations at the Airpark are from 32 airpark-based aircraft. This 

Airpark does not have instrument flight rule procedures or specific holding patterns. The 

applicant estimated that an aircraft could come within 6 km (4 mi) of the IIFP site during a visual 

landing approach. For the Lea County Zip Franklin Memorial Airport, the applicant evaluated 

the airport’s instrument flight rule and visual flight rule procedures and determined that holding 

and approach patterns would be more than 30 km (18 mi) from the IIFP site. Based on the 

evaluations discussed in this paragraph, the applicant concluded that the holding and approach 

patterns for the nearby airports are more than 3 km (2 mi) away from the IIFP site and satisfy 

the third proximity acceptance criterion in NUREG-800 (NRC, 2010),. Therefore, the applicant 

concluded that aircraft crash hazards do not pose any safety concern to the IIFP facility. 

The NRC staff reviewed the information the applicant presented on aircraft crash hazards to the 

IIFP facility. The NRC staff finds that the applicant’s use of the proximity acceptance criteria 

specified in NUREG-0800, Section 3.5.1.6 (NRC, 2010), to assess the aircraft crash hazards to 

the IIFP facility is acceptable because these criteria are an NRC-accepted approach to 

determining potential aircraft crash hazards to nuclear power plants in the United States. The 

NRC staff verified the applicant’s evaluation of airport operations on the three airports nearest to 

the IIFP site—Lea County Regional Airport, Hobbs Industrial Airpark, and Lea County Zip 

Franklin Memorial Airport—using the airport operations information from the Federal Aviation 

Administration’s, “Air Traffic Activity System (ATADS): Airport Operations,” at 

http://aspm.faa.gov/opsnet/sys/Airport.asp and from Airnav.com’s, “Airport Information,” at 

http://www.airnav.com/airports/. 

The NRC staff concludes that the applicant’s acceptably demonstrated that the operations of the 

airports near the IIFP facility site do not present a safety concern to the facility by showing that 

the annual number of operations at the nearby airports is less than the maximum operation 

number, as specified in the first proximity acceptance criterion in NUREG-0800 (NRC, 2010). 

The NRC staff also finds that the applicant provided sufficient information to demonstrate that 

the nearest edge of military training routes, including low-level training routes, meets the second 

proximity acceptance criterion. The aircraft holding and approach patterns at various airports 

are of sufficient distance from the IIFP site to satisfy the third proximity criterion. Based on this 

review, the NRC staff concludes that the aircraft crash hazards are not a safety concern to the 

proposed IIFP facility. The NRC staff finds that the applicant has met the regulatory 

requirements in 10 CFR 70.65(b)(1). 
1-17 

1.3.3.2.5 Nearby Bodies of Water 

The applicant indicated that there are two playas on the site, but no significant bodies of water 

or rivers are within 80 km (50 mi) of the site (IIFP, 2012a, b). The applicant stated that the 

nearest river to the proposed site is Pecos River approximately 80 km (50 mi) south to 

southeast from and at an elevation approximately 213 m (700 ft) below the proposed site. The 

nearest dams are Brantley Dam, approximately 98 km (61 mi), and Avalon Dam, approximately 

106 km (66 mi) from the proposed site. Both dams are at an elevation more than 168 m (550 ft) 

below the site. Because of this information, the NRC staff determined—based on the large 

distances to the nearest bodies of water—as described in the LA, that such bodies of water do 

not present a credible disruptive event for the proposed IIFP facility. The NRC staff finds the 

applicant’s conclusions acceptable that there are no intermediate– or high–consequence events 

related to nearby bodies of water. The NRC staff finds that the applicant has met the regulatory 

requirements in 10 CFR 70.65(b)(1). 

1.3.3.3 Demographics 

There is no residential population or recreational areas within the 5-mile radius of the site. The 

nearest residential center and the closest town to the site is Hobbs, located in Lea County, 

19 km (12 mi) east of the site on U.S. Highway 62/180. Hobbs is the largest city in Lea County 

with a population of 28,657 in the 2000 census. The tri-county area surrounding the site 

includes Lea County, New Mexico; Gaines County, Texas; and Andrews County, Texas. Lea 

County has the largest population with 67,727 in the 2010 census, and the three counties had a 

combined population of approximately 100,000 in 2010 compared to 83,000 in the 2000 census. 

Individuals nearest the facility include IIFP employees onsite, a local resident 2.74 km (1.7 mi) 

northwest of the site, employees who work at the four industrial facilities within 5 miles of IIFP, 

and occasional ranchers or oil field workers. 

The industrial facilities located within 5 miles of the IIFP consist of three power plants and one 

natural gas recovery facility. The power stations include Cunningham Station on the facilities 

west boundary, Colorado Energy Station 3.1 km (1.9 mi) northeast, and Xcel Energy Maddox 

Station 3.5 km (2.2 mi) east. DCP Midstream Linam Ranch Plant is a natural gas recovery 

facility which operates 5.8 km (3.6 mi) southeast of the plant. IIFP projects a shift workforce of 

approximately 140 employees spread over three 8-hour shifts. 

The nearest local school facilities, daycare, nursing homes and hospitals are located in Hobbs, 

NM. The educational institutions include three colleges, a high school and an alternative high 

school, three middle schools, and twelve elementary schools—as well as two private schools. 

The Lea Regional Medical Center is the nearest hospital at 31.4 km (19.5 mi) from the site and 

has a capacity of 221-licensed beds. There are no school facilities or hospitals located within 

8 km (5 mi) of the facility. 

The applicant provided a summary of demographic information based on the most recent 

census data that showed the population distribution as a function of distance from the proposed 

facility. The applicant=s descriptions are consistent with the more detailed information in the ISA 

Summary (IIFP, 2012b), the ER (IIFP, 2011a), and the EP (IIFP, 2012c). The NRC staff finds 

that the applicant has met the regulatory requirements in 10 CFR 70.65(b)(1). 
1-18 

1.3.3.4 Meteorology 

The LA (IIFP, 2012a) and Section 1.3 of the ISA Summary (IIFP, 2012b) provide a 

meteorological description of the site and its surrounding area. The information includes several 

tables on prevailing winds and precipitation. The descriptions also address severe weather, 

thunderstorms, lightning, wind, and snow. The section provides the bounding meteorological 

conditions for the local area of the proposed facility, and the ISA Summary contains the 

evaluation of the potential high and intermediate accident sequences. 

1.3.3.4.1 Tornado Hazard and Tornado-Generated Missiles 

The applicant discussed tornadoes at the proposed IIFP site in Section 1.7.3.3 of the LA 

(IIFP, 2012a); Section 1.3.2.3 of the ISA Summary (IIFP, 2012b; and Section 3.6.1.6 of the ER 

(IIFP, 2011a). Ninety-two tornadoes were recorded in Lea County, NM, where the IIFP facility is 

located, during 1950–2010 (IIFP, 2012b). Among these recorded tornadoes, eight were F2 

(Fujita Scale) tornadoes and one was an F3 tornado (IIFP, 2012a). 

Using the methodology in NUREG/CR-4461, Revision 2, “Tornado Climatology of the 

Contiguous United States,” issued February 2007 (NRC, 2006), the applicant estimated that the 

tornado strike probabilities for the 1-, 2-, and 4-degree boxes that contain the IIFP site are 

greater than 10−5

 per year (highly unlikely) (IIFP, 2012b); therefore, a potential tornado hazard 

to the IIFP facility needs to be considered. The applicant indicated that the design-basis wind 

speed of 217 kilometers per hour (km/h) (135 miles per hour [mph]) is proposed for buildings 

and structures containing licensed material or chemicals or processes that may affect licensed 

material is greater than the estimated tornado wind speed corresponding to an annual 

probability of 10−5

 per year. Therefore, the potential hazard from tornado to the IIFP facility 

does not have to be considered further. Section 1.3.3.3.2 of this SER evaluates and discusses 

the applicant’s design-basis wind speed. 

Using Figure 8-1 of NUREG/CR-4461, Revision 2 (NRC, 2006), the NRC staff verified the 

applicant’s conclusion and determined that the estimated tornado wind speed is less than the 

design-basis wind speed of 217 km/h (135 mph). NUREG/CR-4461, Revision 2 (NRC, 2006), 

provides recommended tornado design-basis wind speed maps for the continental United 

States. Therefore, the NRC staff concludes that the applicant used appropriate meteorological 

data and methodology to assess the probability of a tornado strike. The NRC staff further 

concludes that the applicant appropriately determined that no special consideration of tornado 

hazard is necessary in the design basis for the IIFP facility. Based on this review, the NRC staff 

concludes that the applicant meets the requirements in 10 CFR 70.65(b)(1). 

1.3.3.4.2 High Winds 

The applicant discussed extreme winds at the proposed IIFP site in Section 1.7.3.3 of the LA 

(IIFP, 2012a) and Section 1.3.2.3 of the ISA Summary (IIFP, 2012b). The applicant estimated 

the extreme wind hazard curve for the IIFP site (IIFP, 2011b) using the gust wind speed data 

established in Table 3-2 of DOE-1020-2002, “Natural Phenomena Hazards Design and 

Evaluation Criteria for Department of Energy Facilities,” issued in 2002 (DOE, 2002), for both 

Sandia National Laboratories and Los Alamos National Laboratory in New Mexico. Based on 

this hazard curve, the applicant selected a straight gust wind speed of 217 km/h (135 mph) as 

the design-basis wind speed for its IIFP process building. This gust wind speed corresponds to 

an annual probability of 10−4

. 
1-19 

The NRC staff reviewed the extreme wind information the applicant provided and finds that the 

applicant adequately characterized the extreme wind hazard at the IIFP site. The NRC staff 

finds the applicant’s use of the gust wind speed data established in Table 3-2 of 

DOE-1020-2002 (DOE, 2002) for DOE’s facilities to be acceptable because this standard is 

widely used for wind design of the structures, systems, and components of the DOE facilities to 

safely withstand the high wind hazard. The NRC staff also finds that the applicant’s use of the 

straight gust wind speed of 217 km/h (135 mph) for an annual probability of 10−4

 as a high-wind 

design basis is greater than that recommended in Revision 1 of the Coats and Murray (Coats 

and Murray, 1985) report entitled, “Natural Phenomena Hazards Modeling Project: Extreme 

Wind/Tornado Hazard Models for Department of Energy Sites.” This provides additional 

conservatism in the wind protection design. Based on this review, the NRC staff finds that the 

applicant meets the requirements in 10 CFR 70.65(b)(1). 

1.3.3.4.3 Extreme Precipitation 

The applicant discussed extreme precipitation at the proposed IIFP site in Section 1.7.3.3 of the 

LA and Section 1.3.2.2 and 1.3.2.6 of the ISA Summary (IIFP, 2012a, b) and Section 3.6.1.3 of 

the ER (IIFP, 2011a). The IIFP site is located in a semiarid climate with an annual average 

rainfall of 31.37–40.46 centimeters (cm) (12.35–15.93 inches [in]) (IIFP, 2012b). Frequent rain 

showers and intense thunderstorms occur during summer and account for more than half the 

annual precipitation (IIFP, 2012b). The estimated average rainfall for a 1-hour event in Hobbs, 

NM, is 8.51–8.66 cm (3.35–3.41 in); the estimated average rain fall for a 24-hour event is 

16.43–17.96 cm (6.47–7.07 in) (IIFP, 2011a). 

Using the National Oceanic and Atmospheric Administration’s (NOAA) precipitation data 

(Bonnin, et al., 2011), the applicant calculated the 1-hour, 24-hour, and 48-hour all-season 

precipitations for a return period of 100,000 years using a linear least-square regression 

technique. The estimated 1-hour, 24-hour, and 48-hour all-season precipitations are 18.3 cm 

(7.2 in), 36.6 cm (14.4 in), and 43.2 cm (17.0 in), respectively. 

The NRC staff reviewed the applicant’s information regarding extreme precipitation. The NRC 

staff finds that the applicant’s estimation of extreme precipitations for a return period of 

100,000 years is acceptable because the site-specific precipitation data the applicant used are 

from a reliable Federal Government agency, NOAA, and were peer reviewed. In addition, the 

applicant used the linear least square regression technique to estimate extreme precipitations 

with an annual probability of 10−5

, which is an acceptable method to estimate behavior that is 

outside the data range. Based on this review, the NRC staff finds that the applicant provided 

sufficient information on extreme precipitation to support conduct of the ISA and meets the 

requirements in 10 CFR 70.65(b)(1) 

1.3.3.4.4 Flood 

The applicant discussed potential flooding at the proposed IIFP site in Section 1.7.3.3 of the LA, 

Section 1.3.2.6 of the ISA Summary, and Section 3.4.11 of the ER (IIFP, 2012a,b and 2011a). 

According to the applicant, the IIFP site grade is above the 100-year flood elevation 

 

(IIFP, 2011a). The applicant indicated that the Federal Emergency Management Agency 

(FEMA) classifies the IIFP site to be in FEMA Flood Zone D (IIFP, 2012b). By FEMA definition, 

for the areas located in Flood Zone D, area flooding is possible with undetermined flood hazard 

(FEMA, 2011). 
1-20 

No significant bodies of water or rivers are within 80 km (50 mi) of the site. Therefore, the 

applicant indicated that any flooding at the site would be caused by short-term precipitations 

resulting in flash flooding. Sixty-eight flood events were recorded in Lea County, NM, between 

1950–2010. Twenty-nine of the 68 events were reported in Hobbs, NM (in FEMA Flood 

Zone D) (IIFP, 2012a). Hobbs, NM, is approximately 18.3 km (11.4 mi) east of the IIFP site and 

is at an elevation of 38–52 meters (m) [125–170 feet (ft)] lower than the IIFP site. 

In general, parts of the area west and north of the IIFP site are at a slightly higher elevation than 

the IIFP site, and the topography gradually slopes down to the east and south directions for 

several miles. Therefore, the applicant stated that the IIFP site could receive drainage water 

from the northwest direction with a slope of approximately 0.21 percent and would discharge to 

the northeast, southeast, and southwest directions with a slope ranging from 0.35 to 

0.46 percent (IIFP, 2012b). 

The applicant conducted a preliminary flood screening analysis for the IIFP site using the 

guidelines in DOE-1023-95, “Natural Phenomena Hazards Assessment Criteria” (DOE, 1995) 

and concluded that flooding is not a design-basis event (IIFP, 2012b). In the screening 

analysis, the applicant assessed the flooding potential from the Pecos River and failure of the 

Brantley and Avalon Dams. The Pecos River is approximately 80 km (50 mi) away from and at 

an elevation of 213 m (700 ft) below the IIFP site. Brantley Dam is approximately 98 km (61 mi) 

away from and 168 m (550 ft) below the IIFP site. Avalon Dam is approximately 106 km (66 mi) 

away from and 192 m (630 ft) below the IIFP site. Because of the substantial distances of these 

features and the large differences in their elevation from the IIFP site, the applicant concluded 

that the Pecos River, Brantley Dam, and Avalon Dam are not flood hazards. The only plausible 

flooding hazard to the IIFP site is from storm water runoff during rain events. The applicant 

further excluded hurricanes and tsunamis from being potential flood hazards to the IIFP site 

because the site is approximately 805 km (500 mi) away and 1,158 m (3,800 ft) higher than the 

Gulf of Mexico. 

To estimate the potential effects of rainfall-induced storm water runoff, the applicant estimated 

the 1-hour, 24-hour, and 48-hour all-season precipitation corresponding to a 100,000-year 

return period by extrapolating the NOAA precipitation data (Bonnin, et al, 2011) using a linear 

least-square procedure. Considering the 1-hour, 24-hour, and 48-hour all-season precipitation 

for a 100,000-year return period, the applicant estimated the maximum flood (standing water) 

level to be 12.2 cm (4.8 in) (IIFP, 2012a). This estimate was determined using the general site 

contour information with a gentle slope of 2.1 percent northwest of the site and assuming that 

the probable maximum area with rainfall water that might affect the process building is 65,154 

square kilometers (km2

) (1.61 acres). The applicant indicated that the building and structures 

containing items relied on for safety (IROFS) and the plant roadway levels will be constructed at 

least 15.2 cm (6 in) above grade so that the potential flood water (at an estimated level of 

12.2 cm [4.8 in]) will not prevent the IROFS from performing the intended safety functions. 

The NRC staff reviewed the information on potential flood hazard at the IIFP site and 

determined that the applicant acceptably screened the potential sources for flooding hazards 

because the applicant conducted the assessment using reasonable distance and elevation 

information for the features analyzed and the IIFP site. The NRC staff also determined that the 

applicant’s precipitation estimates for rainfall with an annual probability of 10−5

 are acceptable 

because the applicant used the rainfall data from a reliable source (NOAA) and extrapolated the 

precipitation data via a commonly used method. In addition, the NRC staff determined that the 

applicant’s maximum flood level is reasonable because the applicant used site-specific 

topographic information. Consequently, the NRC staff finds that the applicant’s flood hazard 
1-21 

assessment considering only that the site precipitation is reasonable, and flooding is not a 

hazard to the IIFP site. Based on this review the NRC staff finds that the applicant meets the 

regulatory requirements in 10 CFR 70.65(b)(1) 

1.3.3.4.5 Snow 

The applicant discussed snow in Hobbs, NM, in Section 1.7.3.3 of the LA (IIFP, 2012a), 

Section 1.3.2.7 of the ISA Summary (IIFP, 2012b), and Section 3.6.1.3 of the ER (IIFP, 2011a). 

The mean annual snowfall is 13.0 cm (5.1 in) recorded at the Hobbs weather station. The 

maximum recorded snow accumulation for Hobbs, NM, is 31.0 cm (12.2 in), and a 100-year, 

2-day snowfall is 30.7 cm (12.1 in) (IIFP, 2011a). 

The applicant stated that it will consider the snow loads in the design of the IIFP facility. The 

applicant proposed to use the sum of the snow load associated with a 100-year snowfall and the 

load from the estimated 48-hour all-season precipitation as the design-basis extreme 

environmental ground snow load. This snow load has an annual probability of 10−5

 for the IIFP 

site. The applicant stated that it will use the guidance specified in American Society of Civil 

Engineers (ASCE) 7-05, “Minimum Design Loads for Buildings and Other Structures,” to 

calculate roof snow load using the design-basis ground snow load for the IIFP building 

structures (IIFP, 2011a). 

The NRC staff reviewed the applicant’s snow hazard discussion for the IIFP facility and 

determined that the snow information is adequate for the applicant to assess potential snow 

hazard in its ISA. Based on this review, the NRC staff finds that the applicant meets the 

requirements in 10 CFR 70.65(b)(1). 

1.3.3.4.6 Lightning 

The applicant discussed lightning phenomena in Section 1.7.3.3 of the LA (IIFP, 2012a); 

Section 1.3.2.5 of the ISA Summary (IIFP, 2012b); and Section 3.6.1.6 of the ER (IIFP, 2011a). 

The applicant indicated that, from 1950–2004, two lightning events were reported to have 

caused loss of life, injury, and property damage in Lea County, NM. One occurred in Hobbs 

with minor property damage, and the other occurred in Lovington with two deaths. 

The applicant evaluated the potential for a lightning strike to be credible at the proposed site. 

The applicant accounted for the lightning hazard by providing lightning protection to the IIFP 

facility, as indicated in Chapter 7 of the ISA Summary (IIFP, 2012b). In Table 7-1 of the ISA 

Summary, the applicant listed the National Fire Protection Association (NFPA) 780, “Standard 

for the Installation of Lightning Protection Systems,” for lightning protection (NFPA, 2004). 

Based on the inclusion of NFPA 780 for lightning protection of the IIFP structures, the NRC staff 

finds that the applicant adequately considered lightning strike hazard through lightning 

protection design. Based on this review, the NRC staff finds that the applicant meets the 

requirements in 10 CFR 70.65(b)(1). 

1.3.3.4.7 Meteorology Conclusions 

As discussed in Sections 1.3.3.3.1 through 1.3.3.3.6 of this SER, the applicant provided 

appropriate meteorological data, including a summary of design-basis values for accident 

analysis of maximum snow loads and probable maximum precipitation, as presented in the ISA 

Summary (IIFP, 2012b). The applicant also provided appropriate design-basis information for 
1-22 

lightning, high winds, tornadoes, hurricanes, and extreme precipitation. The applicant’s 

descriptions are consistent with the more detailed information in the ISA Summary (IIFP, 2012b) 

and the ER (2011a). Based on this review, the NRC staff finds that the applicant meets the 

requirements in 10 CFR 70.65(b)(1). 

1.3.3.5 Hydrology 

Site surface water and groundwater hydrology is discussed in Sections 1.7.3.4 of the LA 

(IIFP, 2012a); Section 1.4 of the ISA Summary (IIFP, 2012b); and 3.4.11 of the 

ER (IIFP, 2011a). 

The applicant stated that there is a small stream running through the property. This stream is 

predominantly dry during the year (IIFP, 2012b). There are also two playas on the site, but no 

significant bodies of water or rivers are within 80 km (50 mi) of the site (IIFP, 2012a, b). The 

applicant stated that surface water is lost through evaporation, resulting in high salinity 

conditions in the waters in and soils associated with the playas. Because IIFP is located in a 

semi-arid region, evaporation and transpiration dominate resulting minimal surface water 

occurrence or groundwater recharge. 

The IIFP site sits on the Ogallala (High Plains) Aquifer underground reservoir system. This 

aquifer provides water supply to the region, including the proposed IIFP site. The water table of 

this aquifer generally varies from 15 to 90 m (50 to 200 ft) below ground. Specifically, for the 

IIFP site, the Ogallala Aquifer is at a depth of 9 to 47 m (30 to 150 ft). The applicant stated that, 

although the Ogallala Aquifer contains hard water, the quality of the water is good. The 

applicant further indicated that withdrawals from the aquifer exceed recharge to it. Therefore, 

the Ogallala Aquifer is being depleted. 

The applicant stated the IIFP site is above the elevation of the 100-year and the 500-year flood 

plains. The IIFP site storm system will be designed to accommodate a 100-year return period 

precipitation event. 

The NRC staff reviewed the information provided in the LA (IIFP, 2012a), ISA Summary 

(IIFP, 2012b), and the ER (IIFP, 2011a). Based on the review of the information concerning site 

hydrology, the information provided by the applicant meets the regulatory acceptance criteria in 

Sections 1.3.4.3(4) and (5) of NUREG-1520 (NRC, 2002) and is, therefore, acceptable because: 

(1) the information is accurate and is from reliable sources; and (2) the applicant provides 

design-basis information for hydrological conditions applicable to the site. In addition, the NRC 

staff reviewed the applicant’s hydrological data in the ISA Summary (IIFP, 2012b) and 

ER (IIFP, 2011a) and finds that they provide sufficient information to assess site flooding 

hazards and to assess the ground and surface water impacts. Based on this review, the NRC 

staff finds that the applicant meets the requirements in 10 CFR 70.65(b)(1). 

1.3.3.6 Site Geology 

The applicant summarized the geology of the IIFP site in Section 1.7.4.1 of the LA 

(IIFP, 2012a); Section 1.5.1 of the ISA Summary, (IIFP, 2012b); and Section 3.3 of the 

ER (IIFP,2011a). 
1-23 

1.3.3.6.1 Physiography 

As the applicant documented in the aforementioned references, the IIFP site is located within 

the Pecos Valley section of the Great Plains Physiographic Province of North America. The site 

is near the southwestern margin of the Llano Estacado; a flat, semiarid table and that covers 

97,000 km2

 (37,500 square miles [mi2]) of western Texas and eastern New Mexico. The 

tableland rises in elevation from 900 m (3,000 ft) in the southeast to 1,500 m (5,000 ft) in the 

northwest. Erosion of the tableland forms a prominent topographic escarpment along its 

western and northwestern edge called the Mescalero Ridge, which lies just to the west of the 

IIFP site. The Llano Estacado is composed of porous late Tertiary age fluvial gravels and eolian 

sands of the Ogallala Formation, which are covered by late Pleistocene and Holocene 

dark-brown to reddish brown sands, sandy loams, and clay loams (Trimble, 1980). The 

Ogallala Caprock, an erosion-resistant caliche unit, holds up the tableland. Based on this 

review, the NRC staff finds that the applicant meets the requirements in 10 CFR 70.65(b)(1). 

1.3.3.6.2 Basin Geology 

The proposed IIFP site is located within the Central Basin Platform of the Permian Basin, 

between the Midland and Delaware sub-basins. The Permian Basin is a downward flexure of a 

large thickness of originally flat-lying bedded, sedimentary rock. Downward flexure and overall 

development of the Permian Basin is a relic of geological activity that is estimated to have taken 

place between 65 and 350 million years ago. The top of the Permian section is approximately 

434 m (1,425 ft) below ground surface. The Permian Basin strata are overlain by sedimentary 

rocks of the Triassic age Dockum Group. The upper formation of the Dockum Group is the 

Chinle Formation, locally overlain by the Tertiary Ogallala, Gatuña, or Antlers Formations or by 

Quaternary alluvium. Based on this review, the NRC staff finds that the applicant meets the 

requirements in 10 CFR 70.65(b)(1). 

1.3.3.6.3 Soils 

As discussed in the ER (IIFP, 2011a), soils for this part of Lea County generally comprise 

shallow to deep gravelly and loamy horizons to deep sandy soils formed from eolian and fluvial 

processes in the recent geologic past. Because of the arid environment, zones of soft and hard 

caliche are common. Loamy and sandy soils have low shrinking and swelling potential, which 

limits the impact that soil swelling and shrinkage may have on building foundations. The NRC 

staff determined that the applicant has made a conservative assumption based on this 

description that the soil profiles at the site will include several meters of loose eolian sands 

above dense to very dense and partially cemented sands and silts. 

The applicant commits to implement a site-specific geotechnical and geophysical investigation 

and analysis plan (soil characterization plan) as described in Section 3.2.4.3 of the LA (IIFP, 

2012a) to obtain accurate values for the soil liquefaction, settlement, bearing capacity and static 

and dynamic soil properties based on soil measurements at the IIFP site. The NRC staff 

reviewed the soil characterization plan and the commitments to implement the plan consistent 

with the appropriate ASTM standards listed in Section 3.1.5.3 of the LA (IIFP, 2012a). The NRC 

determined the commitments to implement the site characterization plan are adequate to obtain 

accurate data for the detailed design. The NRC staff concludes that commitments to construct 

the facility consistent with the data obtained using the soil characterization plan committed to by 

IIFP in the LA (IIFP, 2012a) meet the baseline design criteria for natural phenomena hazards 

described in 10 CFR 70.64(a)(2) and the content requirement in 70.65(b)(1), and are, therefore, 

acceptable. 
1-24 

The applicant’s geotechnical and geophysical investigation and analysis plan, as described in 

the LA, Section 1.7.4 (IIFP, 2012a), requires the applicant to acquire additional detailed 

information on the site soil class, seismic site response, liquefaction potential, soil settlement 

potential, and allowable bearing capacity for the IIFP site (see ISA Summary, Section 1.5.4 

[IIFP, 2012b]). The field activities include dilatometer soundings, seismic cone penetration 

tests, crosshole seismic tests, standard penetration tests, auger borings, and soil sampling. 

The applicant commits to implement the geotechnical and geophysical investigation analysis 

plan in accordance with the appropriate American Society for Testing and Materials (ASTM) 

standards listed in the LA, Section 3.1.5 (IIFP, 2012a). The NRC staff reviewed the plan and 

determined its implementation will provide accurate data. The applicant’s commitments to 

refine the soil calculations using site-specific data further supports the NRC staff’s finding that 

IIFP meets the baseline design criteria in 10 CFR 70.64 and the content requirement in 

70.65(b)(1). 

The NRC staff reviewed the site geology description for the IIFP site in accordance with 

Section 1.3 of NUREG-1520, Revision 0 (NRC, 2002), and finds that the applicant adequately 

described and summarized the geological information needed to support evaluations of the 

potential natural hazards that could impact the safe operation of the proposed facility. The 

information on which the applicant relied is consistent with the NRC staff’s understanding of the 

site’s geologic conditions based on information provided by the applicant, as well as information 

available from the geologic literature. The review verified that the site description is consistent 

with the information used as a basis for the LA, ER, and ISA Summary. Based on this review, 

the NRC staff finds that the applicant has met the requirements in 10 CFR 70.65(b)(1). 

1.3.3.6.4 Seismic Hazard 

The applicant described seismic hazards of the IIFP site in Section 1.7.4.2 of the LA 

(IIFP, 2012a); Section 1.5.2 of the ISA Summary (IIFP, 2012b); and Section 3.3.1 of the 

ER (IIFP, 2011a). 

The NRC staff reviewed the following areas concerning the seismic hazard, which are 

applicable to the safety analysis and design of the proposed IIFP facility: 

• seismic source characterization 

• seismic hazard calculation 

• site-specific hazard spectra 

• surface faulting 
1-25 

1.3.3.6.5 Seismic Source Characterization 

Geological and Tectonic Settings 

As noted in the ER (IIFP, 2011a), the area around Hobbs, NM, is seismically quiet and 

structurally stable. The Laramide Orogeny (late Cretaceous to Early Tertiary time) uplifted the 

region to its present elevation, and no substantial tectonic activity has occurred since this early 

Tertiary deformation. The Permian Basin has subsided slightly since the Laramide Orogeny. 

However, this subsidence is believed to have resulted from dissolution of the Permian evaporite 

deposits by groundwater or possibly compaction from the extraction of oil and gas deposits. No 

active faults have been identified at the site, and the only evidence of past active faulting is 

related to the development of the Permian Basin and the Laramide Orogeny. The nearest 

evidence of Quaternary faulting (fault displacements within approximately the last 2 million 

years) is 161 km (100 mi) west of the site in the Basin and Range tectonic province. 

Historic Seismicity 

Earthquakes in the region of the proposed IIFP site include isolated and small clusters of 

low-to-moderate magnitude events toward the Rio Grande Valley of New Mexico and in Texas, 

southeast of the proposed site. According to the applicant, no earthquakes in the site region are 

known to be correlated to specific faults, and many of the small magnitude events are thought to 

be related to water injections and withdrawal used for secondary recovery operations in oil fields 

in the Central Basin Platform area. 

The largest earthquake known to occur within 322 km (200 mi) of the site was the 

August 16, 1931, earthquake near Valentine, TX. This earthquake had an estimated body wave 

magnitude (mb) of 5.6 to 6.4 and produced a maximum Modified Mercalli (MM) intensity of VIII 

(Dumas, 1980). This earthquake occurred approximately 260 km (160 mi) from the proposed 

IIFP site location. Within 80 km (50 mi) of the site, the largest earthquake was a local 

magnitude (ML ) 4.7 event in 1992, which occurred approximately 50 km (30 mi) southeast of 

Hobbs, NM. No injuries and only minor damage were reported from this event (The New York 

Times, 1992). 

The ER (IIFP, 2011a) provides a catalog of the largest earthquakes to occur in New Mexico 

since 1869. The catalog was compiled by Yarger (2009) at the New Mexico Institute of Mining 

and Technology and includes 30 events with estimated magnitudes between 4.5 and 5.5 (all 

magnitudes in the catalog are reported as duration magnitudes, which closely approximate ML). 

These earthquakes have maximum estimated MM intensity values of VI to VIII. MM values of 

VIII can result in considerable damage to poorly constructed structures, but only slight damage 

to specially designed structures. 

The NRC staff reviewed the site geology description for the IIFP site in accordance with 

Section 1.3 of NUREG-1520, Revision 0 (NRC, 2002), and finds that the applicant adequately 

described and summarized the seismic source information needed to support the seismic 

hazards evaluation. The information relied upon by the applicant is consistent with the NRC 

staff’s understanding of the potential seismic sources in eastern New Mexico and western 

Texas, based on information from the geologic literature. The review verified that the 

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

environmental review, and ISA Summary. Based on this review, the NRC staff finds that the 

applicant meets the requirements in 10 CFR 70.65(b)(1). 
1-26 

1.3.3.6.6 Seismic Hazard Calculation 

Table 1-8 of the LA (IIFP, 2012a) and Table 1-3 of the ISA Summary (IIFP, 2012b) provide the 

seismic hazard criteria for the proposed IIFP facility. The criteria specify peak ground 

accelerations for 2,500-year return periods based on the U.S. Geologic Survey (USGS) 2002, 

National Seismic Hazard Maps (Frankel, et al., 2002). Although not specified in the LA or ISA 

Summary, the applicant indicated in the RAI response to RAI#SS-5-2, Revision C, page 33 

(IIFP, 2011b), that the seismic design of all structures will be based on the ground motions from 

the USGS National Seismic Hazard Map for the 2-percent probability of exceedence in 

50 years, which approximately equates to a 2,500-year return period. The applicant also noted 

that the specific design methodology used to ensure a sufficient design margin necessary to 

achieve the beyond-design-basis performance objectives will depend on the consequence 

classification. Buildings that contain material that have the potential for high radiological 

consequences to the public will be designed as PC-3 structures, as described in 

DOE-1020-2002 (DOE, 2002). According to the DOE standard, PC-3 structures designed to the 

2,500-year return period earthquake using nuclear grade design codes and standards will retain 

their safety functions, even if subjected to earthquakes loads from the 10,000-year return period 

earthquake. 

 

For the proposed IIFP site, the USGS hazard maps show that the 2,500-year return period peak 

ground acceleration is 0.12 times the acceleration of gravity (g). The 2,500-year return period 

spectral accelerations for 1.0 second and 0.2 second corner frequencies are 0.041 g and 

0.216 g, respectively. The USGS earthquake hazard maps were developed assuming relatively 

firm bedrock conditions (Site Class B) with average shear-wave velocity (Vs30) in the first 30 m 

(100 ft) of subsoil of 760 m per second (m/s) (2,500 ft per second (ft/s)). Vs30 serves as a 

measure of soil rigidity. 

As noted in the NRC staff’s review of the Louisiana Energy Services, Inc.’s, National Enrichment 

Facility (NRC, 2005), the USGS results for the eastern New Mexico region may be conservative 

relative to other site-specific probabilistic seismic hazard assessments for southeastern New 

Mexico. For example, the USGS 2,500-year return period peak ground acceleration of 0.12 g 

for the proposed IIFP site is comparable to the 10,000-year return period ground motions of 

0.15 g for the National Enrichment Facility, based on the site-specific seismic hazard 

assessment (LES, 2005) and the calculated 10,000-year return period peak ground acceleration 

of 0.15 g for the Waste Isolation Pilot Plant (DOE, 2003). Conservatism in the USGS model 

arises because the USGS hazard calculation used larger maximum magnitudes and seismicity 

rates than are generally supported by the historical or geological record. For example, the 

USGS seismic source model includes regional earthquakes with magnitudes of 6.0. These are 

large in contrast to the historical earthquake catalogs for New Mexico and west Texas, such as 

the one compiled by the applicant, which show that the largest recorded earthquakes have 

magnitudes of 5.5 or less. 

The NRC staff reviewed the seismic hazard calculation for the IIFP site in accordance with 

Section 1.3 of NUREG-1520, Revision 0 (NRC, 2002). The NRC staff finds that the applicant 

adequately described and summarized the geological information needed to support evaluations 

of the seismic hazards. The information relied on by the applicant is consistent with the NRC 

staff’s understanding of the seismic hazard potential in eastern New Mexico and western Texas 

based on information provided by the applicant, as well as information available from the 

geologic literature. The review verified that the seismic hazard calculation is consistent with the 
1-27 

information used as a basis for the LA, ER, and ISA Summary. Based on this review, the NRC 

staff finds that the applicant has met the regulatory requirements in 10 CFR 70.65(b)(1). 

1.3.3.6.7 Development of Site-Specific Spectra 

As described above, the applicant used published soil characterization data from the region of 

the proposed facility to determine the geotechnical properties for the site response calculation at 

the time of this LA review. The applicant’s commitment to use a methodology consistent with 

ASCE/SEI 7-05 and DOE-STD-1020-2002, as described in the following paragraphs, provides 

an adequate basis for regulatory compliance. 

The NRC staff requested through RAIs (IIFP, 2011b) that the applicant commit to use 

conservative values for the soil-specific spectra and commit to a geotechnical and geophysical 

investigation and analysis plan for how the applicant would verify and refine the soil data. IIFP 

provided commitments to implement the soil characterization plan in the LA, Section 1.7.4.1 

(IIFP, 2012a), and provided additional information on the sources for the soil data in LA, 

Section 1.7.4.2 (IIFP, 2012a). Thus, the NRC staff has confidence that the soil properties being 

used for the site-specific spectra are conservative and will be further refined during the detailed 

design. 

In response to NRC staff’s RAIs, the applicant provided a preliminary description of the site soil 

conditions (IIFP, 2011b). Based on its preliminary visits to the proposed IIFP facility site and 

discussions with Pettigrew Associates (a licensed engineering firm) located in Hobbs, NM, the 

applicant stated that the soil conditions at the site consist of approximately a 0.6 m- (2-ft) thick 

layer of soft caliche over hard caliche rock. Based on this soil profile, the conceptual design site 

soil class is either class B or class C. Because the USGS hazard information uses a site soil 

class of B, this class was used by IFFP. Therefore, no modification to the site hazard values will 

be needed if the final site geotechnical investigations confirm site soil class B conditions. If the 

geotechnical investigations to which the applicant committed reveal soil site class C conditions, 

then small amplifications of the hazard spectra are anticipated. In the RAI response, the 

applicant provided an analysis of the amplification of the design spectra, following the 

methodology in ASCE/SEI 7-05, (ASCE, 2006). This analysis indicated spectral amplification 

factors of between 1.2 and 1.7. In the LA, Section 3.2.4.3 (IIFP, 2012a), the applicant 

committed to follow the applicable portions of DOE-STD-1020-2002 for a PC-3 facility in order to 

develop the amplification factors that will be used to develop the seismic inputs for the seismic 

design. This approach to site response is consistent with the applicant’s approach to determine 

seismic design criteria for IROFS, including the process buildings, following the guidance in 

DOE-STD-1020-2002 (DOE, 1995). 

The NRC staff reviewed the seismic hazard calculation for the IIFP site, in accordance with 

NUREG-1520, Revision 0, Section 1.3 (NRC, 2002). The NRC staff finds that the applicant 

adequately described and summarized the site soil information, based on its preliminary 

assessment, to support evaluations of the seismic site response. The applicant has committed 

to keep the site soil data updated and incorporate new information into the detailed design as 

appropriate. The applicant also committed to developing a site response analysis consistent 

with their approach to determine the seismic design criteria per the guidance in 

DOE-STD-1020-2002 (DOE, 2002). The information on which the applicant relied is consistent 

with the NRC staff’s understanding of the seismic hazard potential in eastern New Mexico and 

western Texas based on information the applicant provided, as well as information available 

from the geologic literature. The review verified that the seismic site response calculation is 

consistent with the information used as a basis for the LA (IIFP, 2012a), ISA Summary 
1-28 

(IIFP, 2012b), and ER (IIFP, 2011a). Based on this review of commitments made by the 

applicant to implement a written, site-specific plan to verify the soil characteristics; implement 

the soil characterization plan consistent with industry standards; and use the measurements to 

design and build the facility in accordance with accepted industry standards, the NRC staff finds 

that the applicant meets the requirements in 10 CFR 70.65(b)(1). 

1.3.3.6.8 Surface Faulting 

There is no geologic, geophysical, or seismological evidence of active surface faulting in the 

vicinity of the proposed facility site. As stated in the ISA Summary (IIFP, 2012b), the nearest 

recent faulting is located more than 161 km (100 mi) west of the site. Therefore, the applicant 

did not identify surface faulting as a credible disruptive event for the proposed IIFP facility. 

Based on this review, the NRC staff finds that the applicant meets the requirements in 

10 CFR 70.65(b)(1). 

1.3.3.6.9 Slope Stability 

The applicant discussed the site topography in Section 1.1.1 of the LA (IIFP, 2012a) and 

Section 1.1.1 of the ISA Summary (IIFP, 2012b). The applicant stated that the IIFP site is 

relatively flat with slight undulations in elevation (IIFP, 2012a). In general, the topography of the 

IIFP site gradually slopes down to the east and south directions for several miles (IIFP, 2012b). 

Based on the applicant’s description of the IIFP site topography, the NRC staff finds that slope 

stability should not be a safety concern at the IIFP facility. Therefore, the NRC staff finds that 

the applicant meets the requirements in 10 CFR 70.65(b)(1). 

1.3.3.6.10 Liquefaction and Soil Settlement 

At the time of this LA review, the soil liquefaction, settlement, and bearing capacity for the 

proposed IIFP site were not currently available. The applicant developed a geotechnical and 

geophysical investigation and analysis plan and commits to acquiring necessary information to 

determine the liquefaction potential, soil settlement potential, and allowable bearing capacity of 

the soil for the IIFP site (see ISA Summary, Section 1.5.4 [IIFP, 2012b]). The field activities are 

to include dilatometer soundings, seismic cone penetration tests, crosshole seismic tests, 

standard penetration tests, auger borings, and soil sampling. The applicant indicated that it will 

conduct the geotechnical and geophysical investigation in accordance with appropriate ASTM 

standards listed in the LA, Section 3.1.5 (IIFP, 2012a). 

The NRC staff reviewed the information the applicant provided regarding its plan to verify soil 

liquefaction, settlement, and bearing capacity for the proposed IIFP site. The NRC staff finds 

that the activities the applicant proposed for the geotechnical and geophysical investigation and 

analysis are reasonable and will enable the applicant to collect relevant data to verify the site 

liquefaction potential, soil settlement potential, and allowable bearing capacity of the soil for the 

IIFP site during implementation of the geotechnical and geophysical investigation plan. The 

NRC staff also finds that the ASTM standards (IIFP, 2012a) the applicant plans to use for 

conducting the geotechnical and geophysical investigation are accepted standards in the 

industry and appropriate for determining site class, seismic site response, soil liquefaction, 

settlement, and bearing capacity. Therefore, the NRC staff finds that the geotechnical and 

geophysical investigation and analysis plan will provide accurate data regarding the soil 

liquefaction, settlement, and bearing capacity for the IIFP site. The NRC staff finds that the 
1-29 

applicant’s commitments to incorporate the actual site soil characterization into the facility 

design, consistent with industry standards, provides adequate assurance of safety. 

Implementation of this plan will be monitored and verified by the NRC’s construction inspection 

staff. Consequently, the NRC staff finds that the applicant meets the requirements in 

10 CFR 70.65(b)(1) . 

1.3.3.6.11 Site Geology Conclusions 

As discussed in Sections 1.3.3.5.1 through 1.3.3.5.10 of this SER, the NRC staff finds that the 

applicant provides adequate information on soil, faulting, and seismicity of the proposed site as 

presented in the LA (IIFP, 2012a) and ISA Summary (IIFP, 2012b). In addition, the NRC staff 

finds that the site soil characteristics are based on conservative values for their region in 

southeastern New Mexico. The applicant also provided appropriate seismic design basis 

information. The applicant provided sufficient information on the site geotechnical and 

geophysical investigation program that will be conducted to confirm the design of the IIFP and a 

commitment to update the design, as needed, once the measured site soil characterization 

becomes available. The applicant’s descriptions are consistent with the more detailed 

information in the ISA Summary (IIFP, 2012b) and the ER (IIFP, 2011a). The information the 

applicant provided is consistent with the guidance in Sections 1.3.4.3 of NUREG-1520, 

Revision 0 (NRC, 2002), and is, therefore, acceptable. 

1.3.4 Site Description Conclusions 

The NRC staff has reviewed the site description for IIFP according to Section 1.3 of the 

Standard Review Plan. The applicant has adequately described and summarized general 

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

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

and commercial and manufacturing facilities; (2) population information on the basis of the most 

current available census data to show population distribution as a function of distance from this 

proposed facility; (3) meteorology, hydrology, and geology for the site; and (4) applicable design 

basis events. The reviewer verified that the site description is consistent with the information 

used as a basis for the ER, Emergency Management Plan, and ISA Summary. 

The NRC staff finds that the site description, as discussed above, is consistent with the 

acceptance criteria in NUREG-1520 Chapter 1 (NRC, 2002) and meets the regulatory 

requirements in 10 CFR 40.32 and 10 CFR 70.65. 

1.4 References

(ASCE, 2006) American Society of Civil Engineers, “Minimum Design Loads for Buildings and 

Other Structures,” ASCE/SEI 7–05, Reston, VA: American Society of Civil Engineers, 2006. 

(Bonnin, et al., 2011) Bonnin, M., D. Martin, B. Lin, T. Parzybok, M. Yekta, and D. Riley, 

“Precipitation Frequency Atlas of the United States: NOAA Atlas 14, Volume 1: Semiarid 

Southwest (Arizona, Southeast California, Nevada, New Mexico, Utah),” Version 5.0, Silver 

Spring, MD: U.S. Department of Commerce, National Oceanic and Atmospheric Administration, 

National Weather Service, 2011. 

(Coats and Murray, 1985) Coats, D.W., and R.C. Murray, “Natural Phenomena Hazards 

Modeling Project: Extreme Wind/Tornado Hazard Models for Department of Energy Sites,” 

Revision 1, CA: Lawrence Livermore National Laboratories, 1985. 
1-30 

(DOE, 2006) U.S. Department of Energy, “Accident Analysis for Aircraft Crash into Hazardous 

Facilities,” DOE STD 3014 2006, Reaffirmed May 2006. 

(DOE, 2003) U.S. Department of Energy, “Waste Isolation Pilot Plant Contact Handled (CH) 

Waste Safety Analysis Report,” DOE/WIPPS95S2065, Rev. 7, 2003. 

(DOE, 2002) U.S. Department of Energy, “Natural Phenomena Hazards Design and Evaluation 

Criteria for Department of Energy Facilities,” DOE STD 1020 2002, 2002. 

(DOE, 1995) U.S. Department of Energy, “Natural Phenomena Hazards Assessment Criteria,” 

DOE 1023 95, 1995, Reaffirmed with Errata April 2002. 

(Dumas, et al., 1980) Dumas, B.B., H.J. Dorman, and G.V. Latham, “A Reevaluation of the 

August 16, 1931 Texas Earthquake,” Bulletin of the Seismological Society of America, Vol. 70, 

pp. 1,171–1,180, 1980. 

(FEMA, 2011) Federal Emergency Management Agency, “Definitions of FEMA Flood Zone 

Designations,” http://www.msc.fema.gov, 2011. 

(Frankel, et al., 2002) Frankel, A.D., M.D. Petersen, C.S. Mueller, K.M. Haller, R.l.L. Wheeler, 

E.V. Leyendecker, R.L. Wesson, S.C. Harmsen, C.H. Cramer, D.M. Perkins, and K.S. 

Rukstales, “Documentation for the 2002 Update of the National Seismic Hazard Maps,” U.S. 

Geological Survey Open File Report 02 420, Reston, VA: U.S. Geological Survey, 2002. 

(ICRP, 1977) International Commission on Radiological Protection, “Recommendations of the 

Internaltional Commission on Radiological Protection,” Publication 26, 1977. 

(ICRP, 1982) International Commission on Radiological Protection, “Limits for Intakes of 

Radionuclides by Workers,” Publication 30, 1982. 

(ICRP, 1991) International Commission on Radiological Protection, “1990 Recommendations of 

the International Commission on Radiological Protection,” Publication 60, 1991. 

(ICRP, 1994a) International Commission on Radiological Protection, “Dose Coefficients for 

Intakes by Workers,” Publication 68, 1994. 

(ICRP, 1994b) International Commission on Radiological Protection, “Human Respiratory Tract 

Model for Radiological Protection,” Publication 66, 1994. 

(ICRP, 1994c) International Commission on Radiological Protection, “Dose Coefficients for 

Intakes by Workers,” Publication 68, 1994. 

(ICRP, 1995a) International Commission on Radiological Protection, “Age dependent Doses to 

Members of the Public from Intake of Radionuclides Part 4 Inhalation Dose Coefficients,” 

Publication 71, 1995. 

(ICRP, 1996) International Commission on Radiological Protection, “ICRP Publication 72: Age 

dependent Doses to the Members of the Public from Intake of Radionuclides Part 5, 

Compilation of Ingestion and Inhalation Coefficients,” Publication 72, 1996. 
1-31 

(ICRP, 1997) International Commission on Radiological Protection, “Individual Monitoring for 

Internal Exposure of Workers (preface and glossary missing),” Publication 78, 1997. 

(IIFP, 2012a) International Isotopes Fluorine Products, Inc., “Fluorine Extraction Process and 

Depleted Uranium Deconversion Plant (FEP/DUP) License Application, Revision B,” May 2012, 

Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML12123A245. 

(IIFP, 2012b) International Isotopes Fluorine Products, Inc., “ISA Summary Rev. B for IIFP,” 

May 2012, Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML12123A245. 

(IIFP, 2012c) International Isotope Fluorine Products, Inc., “Emergency Plan Rev B of IIFP 

License Application,” May 2012, Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML12123A245. 

(IIFP, 2011a) International Isotopes Fluorine Products, Inc., “Fluorine Extraction Process and 

Depleted Uranium Deconversion Plant (FEP/DUP) Environmental Report,” Rev. A, 2011, 

Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML100120758. 

(IIFP, 2011b) International Isotopes Fluorine Products, Inc., “Official Response to Seismic and 

Structural RAIs,” Rev. C, September 2011, Agencywide Documents Access and Management 

System (ADAMS) Accession No. ML11263A259. 

(IIFP, 2009) International Isotopes Fluorine Products, Inc., “Memorandum of Agreement 

Between the International Isotopes, Inc., and the New Mexico Environment Department.” 

October 2009. Agencywide Documents Access and Management System (ADAMS) Accession 

No. ML093080290 

(LES, 2005) Louisiana Energy Services, “Integrated Safety Analysis Summary,” Rev. 4, 

Albuquerque, NM, 2005. 

(NFPA, 2004) National Fire Protection Association, “Standard for the Installation of Lightning 

Protection Systems,” NFPA 780 2004, Quincy, MA: National Fire Protection Association, 2004. 

(NRC, 2011) U.S. Nuclear Regulatory Commission, “Environmental Impact Statement for the 

Proposed Fluorine Extraction Process and Depleted Uranium Deconversion Plant in Lea 

County, New Mexico – Draft Report for Comment,” NUREG–2113, 2002 (ADAMS) Accession 

No. ML093080290 

(Note: Final NUREG 2113 will be published after this SER is finalized in fall 2012) 

(NRC, 2010) U.S. Nuclear Regulatory Commission, “Standard Review Plan for the Review of 

Safety Analysis Reports for Nuclear Power Plants: LWR Edition,” Chapter 3, “Design of 

Structures, Components, Equipment, and Systems,” Section 3.5.1.6, Rev. 4, “Aircraft Hazards,” 

NUREG 0800, 2010. 

(NRC, 2006) U.S. Nuclear Regulatory Commission, “Tornado Climatology of the Contiguous 

United States,” NUREG/CR 4461, Rev. 2, 2006. 
1-32 

(NRC, 2005) U.S. Nuclear Regulatory Commission, “Safety Evaluation Report for the National 

Enrichment Facility in Lea County, New Mexico,” NUREG 1827, 2005. 

(NRC, 2002) U.S. Nuclear Regulatory Commission, “Standard Review Plan for the Review of a 

License Application for a Fuel Cycle Facility,” NUREG 1520, Revision 0, 2002. 

(NRC, 1978) U.S. Nuclear Regulatory Commission, “Evaluations of Explosions Postulated To 

Occur on Transportation Routes Near Nuclear Power Plants,” Regulatory Guide 1.91, Rev. 1, 

February 1978. 

(The New York Times, 1992) “Light Earthquake Sends Tremor through Texas and New 

Mexico,” The New York Times, http://www.nytimes.com/1992/01/03/us/light earthquake sends 

tremor through texas and new mexico.html, January 3, 1992. 

(Trimble, 1980) Trimble, D.E., “The Geologic Story of the Great Plains,” Geological Survey 

Bulletin 1493, Washington, DC: U.S. Geological Survey, 1980. 

(Yarger, 2009) Yarger, F., “Seismic Probability in Lea County, NM: A Brief Analysis.” Socorro, 

NM: New Mexico Institute of Mining and Technology, New Mexico Center for Energy Policy, 

2009. 
2-1 

2.0 ORGANIZATION AND ADMINISTRATION 

The purpose of the U.S. Nuclear Regulatory Commission’s (NRC’s) review of the International 

Isotopes Fluorine Products, Inc.’s (IIFP or the applicant), organization and administration is to 

ensure that the proposed management hierarchy and policies will provide reasonable assurance 

that the licensee plans, implements, and controls site activities in a manner that ensures the 

safety of workers and the public and protects the environment. The review also ensures that 

the applicant has identified and provided adequate qualification descriptions for key 

management positions. 

2.1 Regulatory Requirements

The requirements in Title 10 of the Code of Federal Regulations 10 CFR 40.32, “General 

Requirements for Issuance of Specific Licenses,” pertain to the applicant’s training, experience 

and its proposed equipment, facilities and procedures for the effective implementation of health, 

safety, and environment functions. Effectively implementing these functions will better ensure 

adequate health and safety for workers and the public and protection of the environment, and 

not be inimical to the common defense and security. 

2.2 Regulatory Guidance and Acceptance Criteria

Chapter 2 of NUREG-1520, Revision 0, “Standard Review Plan for the Review of a License 

Application for a Fuel Cycle Facility” (NRC, 2002), contains the guidance applicable to NRC’s 

review of the organization and administration section of the license application (LA) 

(IIFP, 2012a). Although this guidance was written to address fuel cycle facilities regulated 

under 10 CFR Part 70, the requirements of 10 CFR Part 40, “Domestic Licensing of Source 

Material,” and 10 CFR Part 70 are sufficiently similar that the majority of the criteria are 

applicable to the IIFP project. Section 2.4.3, “Regulatory Acceptance Criteria,” of NUREG-1520, 

Revision 0 (NRC, 2002), lists acceptance criteria for new and existing fuel cycle facilities. 

Section 2.4.3 of NUREG-1520, Revision 0 (NRC, 2002), contains the acceptance criteria used 

for this review. 

2.3 Staff Review and Analysis

Chapter 2 of the applicant’s LA (IIFP, 2012a) describes the facility’s organization and 

administration policies. These policies are designed to maintain a safe work place, protect the 

public, and comply with applicable regulations. The safety and quality assurance (QA) 

organizations are run independently to ensure that safety remains a priority. The International 

Isotopes Inc.’s (INIS), parent corporation has a quality management system which seeks to 

comply with the International Organization for Standardization’s international standard for quality 

management, issued in 2000, ISO 9001, “Quality Management System Standard” (ISO, 2000). 

In addition, IIFP has a dedicated QA Program, defined in Appendix A to the LA (IIFP, 2012a). 

The purpose of these programs is to ensure that IIFP operations, products, and services comply 

with the regulatory requirements and are safe and reliable. 

This chapter provides an overview of the facilities management structure and describes the 

responsibilities and qualifications required for key management functions. Figures 2-1 and 2-2 

of the LA (IIFP, 2012a) provide flow charts to illustrate the chain of command both during the 

design and build (DB) phase of construction and the startup and operations phase. For both 

phases, the top tier of authority consists of the INIS Board of Directors, followed by the INIS 
2-2 

President, the IIFP President, and the IIFP Chief Operations Officer (COO). The LA describes 

the roles and qualifications of these and subordinate managers, which are summarized below. 

2.3.1 Organization 

As described in Chapter 2 of the LA (IIFP, 2012a), the INIS President and Chief Executive 

Officer (CEO) establish corporate policy and provide direction for all corporate activities. Under 

the INIS President, the IIFP President establishes policy, directs overall operations, and ensures 

that QA and safety procedures are disseminated throughout the entire facility. The IIFP COO 

serves under the IIFP President and oversees the Environmental Safety and Health (ESH) 

Manager and a QA Manager. During construction activities, the COO will oversee the facility 

design, engineering, and contractors. The COO will assign an Engineering Manager to oversee 

the day-to-day responsibility for design and construction. Once activities transition to 

operations, the COO will delegate authority for day-to-day operations to a Plant Manager. 

During the design and construction phase of operations, the INIS corporate QA and ESH 

management will assist to establish the IIFP QA and ESH Programs, under the direction of the 

IIFP COO. The IIFP QA and ESH Managers will be hired before construction to ensure that a 

period of overlap exists to facilitate a smooth transition from corporate QA and ESH 

management to IIFP QA and ESH management. 

During transition from construction to operations, the top level of the organizational structure will 

continue to be led by the IIFP President at the top, followed by the COO, followed by the 

independent positions of an Administration Manager, QA Manager, ESH Manager, Plant 

Manager, Engineering Manager and Training Manager (see LA Figure 2-2 [IIFP, 2012a]). 

Additional supervisors will serve under these managers to oversee operations. 

The NRC staff concludes that the IIFP administration has a clearly defined management 

structure, which stretches from the IIFP President down to the supervisor level. The IIFP 

President has ultimate responsibility for the facility’s safety, security, and the environment. The 

applicant described the management structure for both the DB and operations phases. The 

organizations have clear, unambiguous management structure and line communication, as 

illustrated in the LA Figures 2-1 and 2-2 (IIFP, 2012a). The description of the organizational 

groups responsible for facility oversight and management meet NRC requirements in 

10 CFR 40.32(b) and comply with the guidance in Sections 2.4.3 of NUREG-1520, Revision 0 

(NRC, 2002). Therefore, the NRC staff finds them acceptable. 

2.3.2 Pre-operational Testing and Initial Startup 

In Section 2.1.3 of the LA (IIFP, 2012a), the applicant described the plans for transition from the 

DB phase to the operations phase. The applicant will update the management structure for 

operations to include an Administration Manager, QA Manager, ESH Manager, Plant Manager, 

Engineering Manager, and Training Manager (see the LA Figure 2-2 [IIFP, 2012a]). Additional 

subordinate supervisors will oversee operations. 

The applicant will ensure that these key management positions are filled in advance of startup 

to facilitate a smooth transition from construction to operations. Before operations, the COO 

and the Plant Manager will develop a transition plan for subordinate departments. The plan will 

address the turnover of physical systems, design information, records, and as-built drawings 

from the design build team (see the LA Figure 2-1 [IIFP, 2012a]) to the operations team (see the 

LA Figure 2-2 (IIFP, 2012a). 
2-3 

Before startup, the facility systems and equipment will undergo an acceptance review and 

testing in accordance with the QA Program (IIFP, 2012b). The testing will include verification of 

items relied on for safety (IROFS). Once the systems have successfully passed the QA review, 

the transition plan will be implemented to place the system into operations under the direction of 

the Plant Manager and the “Plant Operation Organization,” as described in Section 2.1.4 of the 

LA (IIFP, 2012a). The acceptance review will cover the system design, function, records, and 

as-built drawings. 

The NRC staff’s review determined that the applicant committed to develop specific plans to 

transition systems from the construction phase to startup and operations. The transition plan 

will be reviewed consistent with the QA Program (IIFP, 2012b). The description of the transition 

plan provides adequate management and organizational controls to ensure that operations will 

meet NRC regulatory requirements. The information the applicant provided meets the guidance 

in Section 2.4.3 of NUREG-1520, Revision 0 (NRC, 2002), and is, therefore, acceptable. 

2.3.3 Organizational Responsibilities and Qualifications 

Section 2.2 of the LA (IIFP, 2012a) provides information concerning the minimum qualifications, 

functions, and responsibilities for key staff positions. This information is summarized below. 

Corporate Employees

The President/CEO is responsible to the INIS Board of Directors and oversees all corporate 

activities. The President/CEO ensures that ESH and QA Programs are established and 

operated consistently throughout all INIS corporate operations. The President/CEO must have, 

as a minimum, a bachelor’s degree in a scientific field or business and 8 years of experience in 

management in a related field (chemical, radiological, or nuclear facilities). 

The corporate ESH and QA Managers support the development of the ESH and QA Programs 

at the IIFP facility, in addition to ensuring consistent programs throughout the parent and 

subsidiary corporate organizations. These corporate managers provide advice, oversight, and 

consultation to assist the IIFP QA and ESH Managers in implementing regulatory compliance. 

These individuals must have a minimum of a bachelor’s degree in engineering or a scientific 

field and at least 5 years of related experience. 

IIFP Facility Employees

The IIFP President reports directly to the INIS President and is ultimately responsible for all 

operations, safety, and security of the IIFP facility. Although corporate ESH and QA managers 

may provide support to the IIFP managers, the IIFP President is responsible for ensuring that 

the IIFP ESH and QA Programs implement, maintain, and distribute sound policies and 

procedures throughout the facility. The IIFP President must have, as a minimum, a bachelor’s 

degree in a scientific field and at least 5 years of management experience in a related field. 

The IIFP President appoints the COO to manage the overall safety and operations of the facility. 

This includes oversight of design, construction, and startup. Once operations begin, the COO 

will oversee hiring of key management, oversight of operations, and proper implementation of 

the ESH and QA Programs, among others. The COO has plant shutdown authority and grants 
2-4 

this authority to appropriate line managers. The COO must review and approve restart 

activities. The COO must have, as a minimum, a bachelor’s degree in a scientific field and 

5 years of relevant experience. 

The Engineering Manager oversees all construction and engineering activities during the DB 

phase of operations and works under the direction of the COO. The Engineering Manager 

oversees the engineering projects, reviews design changes, implements the engineering 

policies and procedures, and tests the operability of systems. During startup and operations, 

this individual will oversee all engineering-related activities and configuration management. The 

Engineering Manager must have, as a minimum, a bachelor’s degree in an engineering field 

and a minimum of 5 years of relevant experience. 

A Plant Engineering and Maintenance Manager will work under the Engineering Manager to 

oversee maintenance support for the facility. This individual will be responsible for supervising 

day-to-day engineering services and corrective maintenance. The individual must have, as a 

minimum, a bachelor’s degree in engineering or science and at least 5 years of experience in a 

related field. Additional support staff, including the Maintenance Supervisors and Configuration 

Management Lead, will work under the direction of the Engineering Manager. 

The INIS President and Board of Directors will select a DB contractor to design and build the 

IIFP facility. The DB contractor will report to the COO during construction. The DB contractor, 

under the direction of the IIFP managers, will be responsible for implementing all Federal, State, 

and local regulatory requirements. The DB contractor will have experience in work at related 

facilities (chemical, radiological or nuclear facilities). 

The QA Manager is independent of other organization structures and reports directly to the 

COO. The QA Manager is responsible for establishing and maintaining the IIFP QA Program. 

The individual provides support to line managers and staff to ensure proper implementation of 

the QA Program throughout the facility. This individual has the authority to report issues directly 

to the IIFP President and to shut down operations, if needed. The QA Manager must have, as a 

minimum, a bachelor’s degree in engineering or science, with at least 5 years of relevant 

experience in the implementation of a QA Program at a related facility. 

The ESH Manager is responsible for overseeing plant safety, compliance with regulatory 

requirements, and environmental protection. This is accomplished through oversight of five 

separate organizations, described below. The ESH Manager reports directly to the COO. This 

individual has the authority to report issues directly to the IIFP President and to shut down 

operations, if needed. The ESH Manager oversees the Configuration Management Program to 

ensure that facility changes are reviewed by the safety and licensing organization and 

incorporated into facility documentation. The ESH Manager must have, as a minimum, a 

bachelor’s degree in engineering, science, or a related field and 5 years of relevant experience 

in a related field. 

The ESH Manager oversees five subordinate organizations, including the Fire Protection, 

Radiation Protection, Emergency Preparedness/Security, Industrial Hygiene/Safety and 

Environmental Programs. The ESH Manager is responsible for the effective coordination of 

these subordinate organizations: 

• The Fire Protection Manager is responsible for developing and maintaining fire 

protection plans and procedures and ensuring their implementation in day-to-day 

operations. 
2-5 

• The Radiation Protection Manager ensures development and implementation of the 

Radiation Protection Program, including training staff and radiation monitoring. 

• The Emergency Preparedness/Security Lead develops the Emergency Plan and 

Security Plan, oversees implementation of emergency procedures, plans exercises, and 

assesses the effectiveness of those plans. 

• The Industrial Hygiene/Safety Lead ensures industrial safety through, training, audits, 

monitoring work place conditions, inspections, and compliance with the Occupational 

Safety and Health Administration’s regulations. 

 

• The Environmental Program Lead oversees effluent monitoring, regulatory compliance, 

an audit program, and recordkeeping. 

Each of these submanagers is required to have, as a minimum, a bachelor’s degree in an 

engineering or scientific field with between 2 and 5 years of applicable experience, as defined in 

Section 2.2 of the LA. 

The Plant Manager will report directly to the COO and oversee the daily operations. These 

responsibilities will include day-to-day compliance with regulatory requirements and 

implementation of internal procedures. This individual will have the authority to shut down 

operations and must approve the restart of the facility. The Plant Manager must have, as a 

minimum, a bachelor’s degree and 5 years of experience in a related field. Several types of 

subordinate managers will serve under the Plant Manager: 

• The Facility Safety Engineer will oversee implementation of the integrated safety 

analysis (ISA), including IROFS. 

• The Shift Superintendents will coordinate operations across the facility and direct facility 

operations on the backshift (e.g., evenings, weekends, holidays, etc.). 

• The Area Shift Supervisors will oversee individual crews and implementation of 

procedures. 

A Training Manager will report to the COO during the design and construction phase of the 

facility to assist the engineering department in developing operation procedures and a training 

program for new employees. Once operations begin, the Training Manager will oversee the 

training program as well as the document control system. The Training Manager must have, as 

a minimum, a bachelor’s degree in science or engineering and a minimum of 5 years of relevant 

experience. 

The Facility Safety Review Committee (FSRC) provides oversight and review of proposed 

changes to procedures and facility systems, including items in the ISA. The FSRC also 

oversees the audit programs and receives input from the As Low As Reasonably Achievable 

(ALARA) Safety Committee in matters related to radiation protection. The Committee reports 

directly to the COO, and the COO appoints the chairperson. The members of the FSRC are 

drawn from the safety and operations groups throughout the facility and have as a minimum a 

bachelor’s degree in engineering or science and at least 3 years of applicable experience at a 

chemical or nuclear facility. Seven years of experience may be substituted for the bachelor’s 

degree. 
2-6 

Based on its review, the NRC staff finds that the applicant has provided clear, unambiguous 

management controls and lines of communication and authority within the organization for 

managing the design, construction, and operation of the facility. The qualifications and 

responsibilities of key individuals involved in health and safety oversight are defined. The 

individuals have authority to shut down operations if conditions warrant. The proposed 

organization summarized in the LA, Chapter 2, provides an acceptable management system for 

ensuring that the design, construction, and operation of the facility will meet NRC regulatory 

requirements. The information the applicant provided meets the applicable acceptance criteria 

in Section 2.4.3 of NUREG-1520, Revision 0 (NRC, 2002), and is, therefore, acceptable. 

2.3.4 Procedures 

Managers are responsible for developing and implementing written procedures for their 

respective areas of responsibility. Procedures are used to define management measures and 

highlight operations which may be impacted by IROFS. Individuals are required to be trained in 

accordance with the written procedures. Non-routine tasks will be conducted under temporary 

orders or radiation work permits. The commitment to use procedures, as described throughout 

the LA and, in particular, Section 2.3.4, “Procedures” (IIFP, 2012a), meets the applicable 

acceptance criteria in Section 2.4.3 of NUREG-1520, Revision 0 (NRC, 2002), and is, therefore, 

acceptable. 

2.3.5 Incidents and Items of Concern 

All individuals working on the IIFP site have the responsibility and right to report to a supervisor 

conditions that are a safety or health concern. The supervisor is responsible for evaluating the 

report and suspending work on the task, if appropriate, or addressing the concern. Supervisors 

are encouraged to seek assistance from the ESH organization. The NRC staff has the right and 

responsibility to raise unresolved concerns to higher levels of management, up through the ESH 

Manager. As described in Section 2.3.7 “Employee Concerns,” of the LA (IIFP, 2012a), 

additional methods of reporting potential unsafe conditions include the following: 

• unusual event report to immediate supervisor 

• direct contact with the ESH organization 

• notification of facility supervisors and management 

• notification of a member of a safety committee 

• contacting IIFP facility managers 

The ESH organization will conduct audits and inspections to self-identify potential safety 

concerns. Walk downs and inspections will be conducted frequently based on the safety 

concerns. The ESH organization will conduct a formal audit focused on safety and regulatory 

compliance quarterly. 

Incidents or accidents will be formally investigated in accordance with the QA Program. 

Management will organize a qualified team of personnel from the plant to evaluate the root 
2-7 

causes and recommend corrective actions. The corrective actions will be implemented, 

followed through to resolution, and documented. 

Based on its review, the NRC staff finds that the applicant provided several methods for IIFP 

employees to report safety concerns to management. In addition, the applicant will conduct 

regular reviews of the facility, including audits, to self-identify potential problems. Items 

identified and accidents will be followed up with corrective actions. The information the 

applicant provided meets the applicable acceptance criteria in Section 2.4.3 of NUREG-1520, 

Revision 2 (NRC, 2002), and is, therefore, acceptable. 

2.4 Evaluation Findings

The NRC staff has reviewed the organization and administration for the IIFP facility according to 

Chapter 2 of NUREG-1520, Revision 0 (NRC, 2002). The application provides (1) clear 

responsibilities and associated resources for the design, construction, operations, and 

modifications of the facility and (2) a plan for managing the project. The application adequately 

describes the management structure during both construction and operation of the facility. The 

flow charts promote a basic understanding of the chain of command. The qualifications 

required for key managers are also addressed. The NRC staff has reviewed these plans and 

commitments and concludes that they provide reasonable assurance that an acceptable 

organization, detailed administrative policies, and sufficient qualified resources have been 

established or are committed to satisfy the applicant’s commitments for the design, construction, 

operations, and modifications of the facility. Therefore, the NRC staff concludes that the LA 

demonstrates compliance with the requirements in 10 CFR 40.32(b) and (c) and provides a 

reasonable assurance of safety. 

2.5 References

(IIFP, 2012a) International Isotopes Fluorine Products, Inc., “Fluorine Extraction Process and 

Depleted Uranium Deconversion Plant (FEP/DUP) License Application, Revision B,” May 2012, 

Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML12123A245. 

(IIFP, 2012b) International Isotopes Fluorine Products, Inc., “Appendix A QAPD Rev B of IIFP 

License Application,” May 2012, Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML12123A245. 

(ISO, 2000) International Organization for Standardization, ISO 9001 2000, “Quality 

Management System Standard,” 2000. 

(NRC, 2002) U.S. Nuclear Regulatory Commission, “Standard Review Plan for the Review of a 

License Application for a Fuel Cycle Facility,” NUREG 1520, Rev. 0, 2002. 

 
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3-1 

3.0 INTEGRATED SAFETY ANALYSIS 

AND INTEGRATED SAFETY ANALYSIS SUMMARY 

International Isotopes Fluorine Products, Inc. (IIFP or the applicant) is required to meet the 

requirements of Title 10 of the Code of Federal Regulations (10 CFR) Part 40, “Domestic 

Licensing of Source Material,” which the U.S. Nuclear Regulatory Commission (NRC) is revising 

to require an integrated safety analysis (ISA) and ISA Summary for large facilities that are or 

would be licensed to possess significant quantities of uranium hexafluoride or uranium 

tetrafluoride. Since the NRC expected to finalize the revisions to 10 CFR Part 40 after IIFP 

submitted its application, the Commission directed the NRC staff, in the staff requirements 

memorandum to SECY-07-0146, “Regulatory Options for Licensing New Uranium Conversion 

and Depleted Uranium Deconversion Facilities,” dated October 24, 2007 (NRC, 2007), to 

impose the performance requirements of Subpart H, “Additional Requirements for Certain 

Licensees Authorized To Possess a Critical Mass of Special Nuclear Material,” of 

10 CFR Part 70, “Domestic Licensing of Special Nuclear Materials,” as part of the licensing 

basis for the application review. In anticipation of a final revised 10 CFR Part 40, the applicant 

conducted an ISA in accordance with the regulations in 10 CFR Part 70, Subpart H, and 

submitted the ISA Summary along with its license application (LA) to the NRC. 

The NRC’s review of the applicant’s ISA and ISA Summary determined whether IIFP addressed 

appropriate hazards through independent analysis or qualitative evaluations of accepted 

engineering practices. The review also determined whether the applicant designated 

acceptable items relied on for safety (IROFS), management measures, and likelihoods and 

consequences for higher risk accident sequences and, whether, with IROFS, the applicant met 

the requirements of 10 CFR 70.61,” Performance Requirements,” which the NRC is 

incorporating into 10 CFR Part 40. 

In particular, the review described in this chapter considered information provided by the 

applicant that is related to the following: 

• commitments regarding the applicant’s safety program, including the ISA, in accordance 

with the requirements of 10 CFR 70.62, “Safety Program and Integrated Safety 

Analysis,” which the NRC is incorporating into 10 CFR Part 40 

• the ISA Summary submitted in accordance with the requirements of 

10 CFR 70.62(c)(3)(ii) and 10 CFR 70.65, “Additional Content of Applications,” which the 

NRC is incorporating into 10 CFR Part 40 

3.1 Regulatory Requirements

The following regulatory requirements of 10 CFR Part 70, which the NRC is incorporating into 

10 CFR Part 40, apply to the ISA and ISA Summary content: 

• 10 CFR 70.62(a), which requires the establishment and maintenance of a safety 

program, including performance of an ISA, that demonstrates compliance with the 

performance requirements of 10 CFR 70.61 

• 10 CFR 70.62(c), which specifies requirements for conducting an ISA, including a 

demonstration that credible high-consequence and intermediate-consequence events 

meet the safety performance requirements of 10 CFR 70.61 
3-2 

• 10 CFR 70.64, “Requirements for New Facilities or New Processes at Existing 

Facilities,” which outlines requirements for baseline design criteria (BDC) and facility and 

system design and facility layout 

• 10 CFR 70.65(b), which describes the contents of an ISA Summary including 

identification of IROFS and management measures. 

The regulations in 10 CFR 70.62 require an applicant to establish and maintain a safety 

program that demonstrates compliance with the performance requirements of 10 CFR 70.61. 

The safety program is required to contain (1) process safety information, (2) an ISA, and 

(3) management measures. The applicant must conduct and maintain the ISA, which must 

identify the following, in accordance with 10 CFR 70.62(c): 

• radiological hazards related to possessing or processing licensed material at the facility 

• chemical hazards of licensed material and hazardous chemicals produced from licensed 

material 

• facility hazards that could affect the safety of licensed materials and thus present an 

increased radiological risk 

• potential accident sequences caused by process deviations or other events internal to 

the facility and credible external events, including natural phenomena 

• consequences and likelihood of occurrence of each potential accident sequence 

identified and the methods used to determine the consequences and likelihoods 

• each IROFS identified in accordance with 10 CFR 70.61(e); the characteristics of its 

preventative, mitigative, or other safety function; and the assumptions and conditions 

under which the item is relied upon to support compliance with the performance 

requirements of 10 CFR 70.61 

As required by 10 CFR 70.61, the ISA must evaluate compliance with performance 

requirements. The requirements in 10 CFR 70.61(b) specify that the risk of each credible, 

high-consequence event must be limited such that the likelihood of occurrence is highly unlikely. 

In addition, the requirements in 10 CFR 70.61(c) specify that the risk of each credible, 

intermediate-consequence event must be limited such that the likelihood of occurrence is 

unlikely. 

The LA must describe the safety program developed under 10 CFR 70.65(a). In addition, the 

applicant is required to submit to the NRC an ISA Summary. As outlined in 10 CFR 70.65(b), 

the ISA Summary is required to contain the following elements: 

• a general description of the site with emphasis on those factors that could affect safety 

• a general description of the facility with emphasis on those areas that could affect safety, 

including an identification of the controlled area boundaries 

• a description of each process analyzed in the ISA in sufficient detail to understand the 
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theory of operation, and, for each process, the hazards that were identified in the ISA in 

accordance with 10 CFR 70.62(c)(1)(i)–(iii) and a general description of the types of 

accident sequences 

• information that demonstrates the licensee’s compliance with the performance 

requirements of 10 CFR 70.61, including a description of the management measures, 

and, if applicable, the requirements of 10 CFR 70.64, “Requirements for New Facilities 

or New Processes at Existing Facilities” 

• a description of the team, qualifications, and the methods used to perform the ISA 

• a list briefly describing each IROFS identified under 10 CFR 70.61(e) in sufficient detail 

to understand their functions in relation to the performance requirements of 

10 CFR 70.61 

• a description of the proposed quantitative standards used to assess the consequences 

to an individual from acute chemical exposure to licensed material or chemicals 

produced from licensed materials which are onsite, or expected to be onsite, as 

described in 10 CFR 70.61(b)(4) and 10 CFR 70.61(c)(4) 

• a descriptive list that identifies all IROFS that are the sole item preventing or mitigating 

an accident sequence that exceeds the performance requirements of 10 CFR 70.61 

• a description of the definitions of unlikely, highly unlikely, and credible, as used in the 

evaluations in the ISA 

3.2 Regulatory Guidance and Acceptance Criteria

Chapter 3 of NUREG-1520, “Standard Review Plan for the Review of a License Application for a 

Fuel Cycle Facility,” (NRC, 2002), contains the guidance applicable to the NRC’s review of the 

applicant’s ISA and ISA Summary (IIFP, 2012b). These sections are applicable in their entirety, 

except for the sections dealing with criticality safety and NUREG-1520,Section 3.4.3.2(5)(b)(i-ix) 

(NRC, 2002) regarding process hazard analysis methods. Because the applicant used the 

What-if method described in NUREG-1513, “Integrated Safety Analysis Guidance Document,” 

(NRC, 2001), IIFP does not have to address the conditions in Section 3.4.3.2(5)(b)(i–ix). 

Sections 3.4.3.1 and 3.4.3.2 of NUREG-1520 (NRC, 2002) contain the acceptance criteria 

applicable to this review. 

3.3 Staff Review and Analysis

This section contains the NRC staff’s programmatic review of the applicant’s proposed safety 

program, proposed ISA commitments, proposed ISA methodology, proposed BDC, and 

proposed defense-in-depth practices.
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3.3.1 Safety Program and Integrated Safety Analysis Commitments 

The NRC staff reviewed the applicant’s proposed safety program commitments identified in 

Section 3.1 of the LA (IIFP, 2012a) to determine whether the three safety program elements 

(process safety information, ISA, and management measures) demonstrate compliance with the 

requirements of 10 CFR 70.62(a)–(d). The NRC staff also reviewed the applicant’s proposed 

commitments to maintain records that demonstrate compliance with the requirements of 

10 CFR 70.62(b)-(d) and records of IROFS failures, in accordance with 10 CFR 70.62(a)(2) 

and 10 CFR 70.62(a)(3). 

3.3.1.1 Process Safety Information 

In Section 3.1.1 of the LA (IIFP, 2012a), the applicant committed to maintaining process 

information addressing the following: 

• the hazards of materials used or produced in the process, including information on 

chemical and physical properties 

• technology of the process, including block flow diagrams or simplified process flow 

diagrams; brief outline of process chemistry; range of operating parameters (e.g., 

temperature, pressure, flow, and concentration); and evaluation of the health and safety 

consequences of potential process accidents 

• equipment used in the process, which includes general information on topics such as the 

materials of construction, piping and instrumentation diagrams, ventilation requirements, 

design codes and standards employed, material and energy balances, IROFS, electrical 

classification, and relief system design and design basis 

In Section 3.1.1 of the LA (IIFP, 2012a), the applicant commits to maintain process safety 

information up-to-date in accordance with the Configuration Management Program described in 

Chapter 11 of the LA (IIFP, 2012a). As discussed in Section 11.3.1 of this Safety Evaluation 

Report (SER), the applicant will use its configuration management system to control 

documentation and review design changes. 

IIFP will submit changes to the ISA Summary to the NRC in accordance with 10 CFR 70.72(d). 

Therefore, the NRC staff concludes that the above-mentioned program elements provide 

reasonable assurance of the following: 

• Consistent with Section 3.4.3.1(1)(a) of NUREG-1520 (NRC, 2002), the LA (IIFP, 2012a) 

contains commitments to compile and maintain an up-to-date database of process safety 

information and is, therefore, acceptable. 

• Consistent with Section 3.4.3.1(1)(b) of NUREG-1520 (NRC, 2002), the applicant will 

implement a safety program that includes procedures and criteria for changing the ISA, 

in accordance with 10 CFR 70.72, “Facility Changes and Change Process,” and is, 

therefore, acceptable. 

As described in Section 3.2.7 of the LA (IIFP, 2012a), the ISA (IIFP, 2012b) was performed and 

will be maintained by a team with expertise in engineering, process safety, safety analysis, and 
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facility process operations. The NRC staff concludes that, consistent with 10 CFR 70.62(c)(2) 

and Section 3.4.3.1(1)(c) of NUREG-1520 (NRC, 2002), the ISA element of the applicant’s 

safety program engaged personnel with appropriate experience and expertise in engineering 

and process operations and is, therefore, acceptable. 

3.3.1.2 Integrated Safety Analysis Commitments 

In Section 3.1.2 of the LA (IIFP, 2012a), the applicant stated that IIFP has conducted an ISA for 

each process. These analyses identified the radiological hazards, chemical hazards, potential 

accident sequences, consequences and likelihood of each accident sequence, and IROFS, 

including the assumptions and conditions under which they support compliance with the 

performance requirements of 10 CFR 70.61. IIFP evaluated the entire facility as part of a plant 

wide process hazards analysis with respect to radiological hazards. The NRC staff finds this 

acceptable since the ISA performed by IIFP addresses the hazards related to the safety of 

licensed materials and the purely chemical hazards of the separated fluoride compounds would 

be addressed under the Occupational Safety and Health Administration’s Process Safety 

Management Program. 

Section 3.3.2 of this SER describes the NRC staff’s evaluation of the applicant’s methods and 

criteria for implementing the ISA methodology. 

In Section 3.1.2 of the LA (IIFP, 2012a), the applicant has committed to implementing programs 

to maintain the ISA and supporting documentation so that it is accurate and up to date. The 

applicant will submit changes to the ISA Summary to the NRC in accordance with 

10 CFR 70.72(d)(1)–(3). Based on this commitment, the NRC staff has determined the ISA 

update process to be adequate and in accordance with the requirements of 10 CFR 70.62(a) for 

adequately maintaining the ISA Summary up to date. 

Section 3.1.2 of the LA states that personnel used to update and maintain the ISA and ISA 

Summary (IIFP, 2012b) are trained in the ISA methods. Section 11.3.3, “Training and 

Qualification,” of this SER describes the general training and qualification of personnel used to 

update or maintain the ISA. The ISA Summary (a nonpublic document) lists the ISA team 

members and their areas of expertise, qualifications, and experience. The NRC staff finds 

these commitments to be consistent with the requirements of 10 CFR 70.62(a) for adequately 

maintaining the ISA Summary up to date. 

In Section 3.1.2 of the LA (IIFP, 2012a), the applicant commits to evaluating proposed changes 

to the facility or its operations using the ISA methods described in the ISA Summary. New or 

additional IROFS and appropriate management measures are designated as required. The 

adequacy of existing IROFS and associated management measures are promptly evaluated by 

IIFP staff to determine whether they are impacted by changes to the facility or its processes. 

The NRC staff finds compliance with this commitment along with the requirements of 

10 CFR 70.72(d)(1)-(3) will result in IROFS being sufficiently available and reliable to ensure 

compliance with performance requirements of 10 CFR 70.61. 

In Section 3.1.2 of the LA (IIFP, 2012a), the applicant commits to addressing unacceptable 

performance deficiencies associated with IROFS. The NRC staff finds compliance with this 

commitment along with the requirements of 10 CFR 70.62 will result in IROFS being sufficiently 

available and reliable to ensure compliance with performance requirements of 10 CFR 70.61. 
3-6 

Section 3.1.1 of the LA (IIFP, 2012a) contains a commitment to maintain the ISA and its 

supporting documentation by the Configuration Management Program described in Chapter 11 

of the LA. The NRC staff finds this commitment to be consistent with Section 3.4.3.1(2)b of 

NUREG-1520 (NRC, 2002). 

Section 3.2.7 of the LA (IIFP, 2012a) contains a commitment to train personnel in the facility’s 

ISA methods and use suitably qualified personnel to update and maintain the ISA and ISA 

Summary. The NRC staff finds this commitment to be consistent with Section 3.4.3.1(2)c of 

NUREG-1520 (NRC, 2002). 

Section 3.1.2 of the LA (IIFP, 2012a) contains a commitment to evaluate proposed changes to 

the facility or its operations by means of the ISA methods and to designate new or additional 

IROFS and appropriate management measures. The NRC staff finds this commitment to be 

consistent with Section 3.4.3.1(2)d of NUREG-1520 (NRC, 2002). 

Section 3.1.2 of the LA (IIFP, 2012a) contains a commitment to address any IROFS’s unacceptable performance deficiencies that are identified through updates to the ISA. The NRC staff 

finds this commitment to be consistent with Section 3.4.3.1(2)e of NUREG-1520 

(NRC, 2002). 

Section 3.1.2 of the LA (IIFP, 2012a) contains a commitment to maintain written procedures 

onsite. The NRC staff finds this commitment to be consistent with Section 3.4.3.1(2)f of 

NUREG-1520 (NRC, 2002). 

Section 3.1.2 of the LA (IIFP, 2012a) contains a commitment to maintain IROFS so that they are 

available and reliable when needed. This commitment is consistent with Section 3.4.3.1(2)g of 

NUREG-1520 (NRC, 2002). 

Based on the assessment provided above, the NRC staff finds the applicant’s ISA commitments 

to be acceptable and consistent with the requirements of 10 CFR 70.62 and acceptance criteria 

contained in Sections 3.4.3(2)b, 3.4.3(2)c, 3.4.3(2)d, 3.4.3(2)e, 3.4.3(2)f, and 3.4.3(2)g of 

NUREG-1520 (NRC, 2002). 

3.3.1.3 Management Measures 

Section 3.1.3 of the LA (IIFP, 2012a) contains a commitment to utilize management measures 

to maintain the IROFS so that they are available and reliable to perform their safety functions 

when needed. Management measures are used to ensure the availability and reliability of 

IROFS on a continuous basis and for any planned changes made to the facility under 

10 CFR 70.72 or any unplanned changes to the facility that result from events outside the 

control of the applicant, such as a seismic event. Management measures ensure compliance 

with the performance requirements assumed in the ISA documentation. The management 

measures are applied to particular structures, systems, components, equipment, and activities 

of personnel and may be graded commensurate with the reduction of risk attributable to that 

IROFS. Chapter 10 of the LA (IIFP, 2012a) further details the management measures, which 

are evaluated in Chapter 11 of this SER. 

3.3.1.4 Baseline Design Criteria 

Section 3.2.4.3 of the LA (IIFP, 2012a) and Section 4.4 of the ISA Summary (IIFP, 2012b) 

address BDC. BDC apply to all structures, systems, and components relied on for safety, 
3-7 

including IROFS. BDC listed in 10 CFR 70.64 for new facilities or processes to which the 

applicant has committed include quality standards and records; natural phenomena hazards; 

fire protection; environmental and dynamic effects; chemical protection; emergency capability; 

utility services; inspection, testing, and maintenance; and instrumentation and controls. In 

addition, the applicant has committed to basing the IIFP facility and system designs on 

defense-in-depth practices. Preference is given to engineered controls over administrative 

controls. For engineered controls, preference is given to passive controls over active controls. 

IROFS are designed to perform their intended safety functions in the event of credible 

(1) natural external events, (2) fire and explosion conditions, (3) dynamic effects, such as the 

effects of missiles and discharging fluids, and (4) chemical releases that can affect radiological 

safety. Based on its review, the NRC staff finds that adequate implementation of the applicant’s 

commitments contained in Section 3.2.4.3 of the LA (IIFP, 2012a) related to BDC in the areas of 

quality standards and records; natural phenomena hazards; fire protection; environmental and 

dynamic effects; chemical protection; emergency capability; utility services; inspection, testing, 

and maintenance; instrumentation and control; and defense-in-depth, will ensure the needed 

availability and reliability of IROFS and are, therefore, consistent with the requirements of 

10 CFR 70.64. 

3.3.1.5 Safety Program and Integrated Safety Analysis Commitments Conclusion 

Based on the evaluations in Sections 3.3.1.1 through 3.3.1.3, above, the NRC staff concludes 

that the applicant meets the requirements of 10 CFR 70.62(a)(1)–(3) to establish and maintain a 

safety program that includes process safety information, ISA, management measures, and 

appropriate safety program records. The NRC staff also concludes that the applicant has an 

appropriate program for establishing and maintaining records of IROFS failures that will be 

retrievable for NRC inspection. IIFP has also appropriately addressed the BDC for new facilities 

and defense-in-depth requirements of 10 CFR 70.64. 

3.3.2 Integrated Safety Analysis Methodology 

Section 3.2.5 of the LA (IIFP, 2012a) and Section 5.2.1 of the ISA Summary (IIFP, 2012b) 

describe the ISA methodology used by IIFP. The IIFP ISA uses a hazard analysis method (i.e., 

the What-if method) to identify the hazards relevant to each node or the IIFP facility in general. 

IIFP’s ISA team reviewed the hazards identified for the “credible worst-case” consequences. 

The team first assigned the credible high- or intermediate-severity consequence accident 

scenarios accident description identifiers, accident descriptions, frequency, or probability and 

then performed a risk index determination. The risk index was used by IIFP to evaluate 

unmitigated risk as unacceptable or acceptable. 

For each accident scenario having an unacceptable unmitigated risk index, IIFP’s ISA team 

defined IROFS and determined the mitigated likelihood for each accident scenario. Using the 

unmitigated initiating event frequency and the failure probability of each IROFS, the team also 

determined the mitigated scenario likelihood and mitigated risk. The risk index method is 

regarded as a screening method for indicating the adequacy or inadequacy of the IROFS for 

any particular accident. A risk analysis approach is used to evaluate the credible accidents that 

potentially exceed the levels identified in 10 CFR 70.61. Figure 3-1, “Integrated Safety Analysis 

Process Flow Diagram,” of the LA (IIFP, 2012a) describes the ISA process steps. The following 

subsections correspond to the blocks in the flow diagram. 
3-8 

3.3.2.1 Define Nodes To Be Evaluated 

The first step of the ISA is to systematically break down the process system, subsystem, facility 

area, or operation being studied into well-defined nodes. The ISA nodes establish the study 

area boundaries in which the various process systems and supporting systems entering or 

exiting the node, or activities occurring in the area, can be defined to allow interactions to be 

studied. 

The applicant divided the plant site into four types of facilities as part of the process hazards 

analysis (PHA) effort: the depleted uranium tetrafluoride (DUF4) facility, the silicon tetrafluoride 

(SiF4) facility, the boron trifluoride (BF3) facility, and support facilities. Specific process 

operations within these facilities are separated logically into “nodes” for PHA evaluation. The 

PHA is broken down in this manner to reduce the complexity of the facility to a manageable 

level and to organize the PHA process and results in a consistent format. These nodes define 

process boundaries for the PHA and are unique process steps within the facility. Equipment 

located outside the process boundary is not evaluated in the node, although interaction between 

systems and potential initiating events from other systems is considered. Operations were 

treated in this manner so that the entire IIFP facility was evaluated in a logical process flow 

approach. This approach is also used to evaluate the hazards associated with each process or 

operation and to identify any new hazards resulting from modifications made to an existing 

process or operation. The applicant identified boundaries that define the point of process 

separation of a hazardous chemical, as well as segregation points where the release of a 

hazardous chemical would not adversely affect licensed materials. The NRC staff finds that the 

ISA Summary lists the defined nodes for the IIFP facility. 

3.3.2.2 Hazard Identification 

3.3.2.3 As indicated in the LA, Section 3.2.5.2 (IIFP, 2012a), the applicant used the “What-if” 

analysis method to identify the hazards for the IIFP process. The NRC staff finds this to be 

appropriate as this method is consistent with the guidance provided in NUREG-1520 

(NRC, 2002) and NUREG-1513 (NRC, 2001). 

According to Section 3.2.5.2 of the LA (IIFP, 2012a), the hazard identification process 

addresses materials with the following characteristics: 

• radioactive 

• flammable 

• explosive 

• toxic 

• reactive 

According to Section 3.2.5.2 of the LA (IIFP, 2012a), the hazards identification process results 

in identification of radiological or chemical characteristics that have the potential for causing 

harm to workers, the public, or the environment. The hazards of concern for the IIFP facility 

relate to either a release (loss of confinement) of uranium hexafluoride (UF6) or hydrogen 

fluoride (HF) or chemicals that may generate HF. In general, the loss of confinement of UF6
3-9 

would initially result in the moisture in the air reacting with the UF6 to form uranyl fluoride 

(UO2F2) and HF as byproducts. UO2F2 is a significant inhalation problem because of its 

dispersible and small particle size. HF can also be released as the byproduct of uranium 

tetrafluoride (UF4) or be generated by the exposure of SiF4 or BF3 to air. The HF, which is in a 

gaseous form, and UO2F2 could be transported through the IIFP facility and ultimately beyond 

the site boundary. Both HF and UO2F2 are toxic chemicals with the potential to cause harm to 

the workers or the public, as documented in Chapter 6 of the LA (IIFP, 2012a). 

For licensed material or hazardous chemicals produced from licensed materials, chemicals of 

concern are those that, in the event of release, have the potential to exceed consequences 

defined in 10 CFR 70.61(b)(4) or 10 CFR 70.61(c)(4). In the LA (IIFP, 2012a), Tables 3-1, 

“Consequence Severity Categories Based on 10 CFR 70.61,” and 3-2, “AEGL Thresholds from 

the EPA for Uranium Hexafluoride, Soluble Uranium, and Hydrogen Fluoride,” present the 

criteria for evaluating potential releases and characterizing their consequence as either “high” or 

“intermediate” for members of the public and facility workers. 

The NRC staff agrees with the applicant’s assessment that UF6, HF or chemicals that may 

generate HF are the hazards of concern for NRC- licensed activities at the IIFP facility. 

The applicant assessed worker exposures based on a duration of 10 minutes, which according 

to the applicant represents a sufficient amount of time to evacuate an area of hazardous 

material leak. Public exposures were estimated to last for 30 minutes. The NRC staff finds this 

to be consistent with self-protective criteria for UF6/HF plumes listed in NUREG-1140, “A 

Regulatory Analysis on Emergency Preparedness for Fuel Cycle and Other Radioactive 

Material Licensees,” issued in 1988 and reprinted in August 1991 (NRC, 1988). The applicant 

used the acute exposure guideline level values of 1, 2, and 3 as threshold concentration levels 

to establish a low-, intermediate-, or high-severity consequence as shown in Table 3-1 of the LA 

(IIFP, 2012a). 

The regulation in 10 CFR 70.61(b)(3) states, in part, for a high-consequence event, “An intake 

of 30 mg or greater of uranium in soluble form by any individual located outside the controlled 

area identified pursuant to Paragraph(f) of this section....” The UF6 concentration in air is not 

directly equivalent to soluble uranium intake. Therefore, IIFP used a high-consequence 

threshold intake value of 75 milligrams (mg). This value corresponds to the threshold for life 

endangerment or “acute chemical exposure,” as defined in 10 CFR 70.61(b)(4). It is noted that, 

several years ago, based on a Nuclear Energy Institute report addressing the health effects of 

soluble uranium, the NRC established a working group with industry representatives to address 

soluble uranium intake thresholds for life endangerment, long-lasting health effects, and mild 

transient health effects. The working group recommended that a regulatory guide be developed 

on this topic. The working group also recommended that this regulatory guide address 

inhalation and dermal exposures to HF. The NRC staff is currently considering developing such 

a regulatory guide that would recommend soluble uranium and HF ISA thresholds. If and when 

the NRC issues new or revised ISA thresholds for soluble uranium and HF, the NRC staff would 

anticipate that its licensees would adopt such thresholds at that time. However, until that time, 

the NRC staff finds the 75 mg intake value for soluble uranium and the qualitative HF thresholds 

provided in Table 3-1 of the LA (IIFP, 2012a) to be acceptable. The NRC staff finds the 

proposed rankings and all other thresholds, for offsite public, environment and workers, 

provided in Table 3-1 of , to be acceptable as they are consistent with the requirements in 

10 CFR 70.61. 
3-10 

The applicant used the What-if analysis method, which is summarized in Table 5-5, “FEP/DUP 

Plan What-If Checklist,” of the ISA Summary (IIFP, 2012b), to identify process hazards for the 

UF6, UF4, SiF4, and BF3 process systems at the IIFP facility. This PHA technique is used to 

identify and document items identified in the hazard analysis meetings. For identified single 

failure events (i.e., those accidents that result from the failure of a single control), the What-if 

method is the recommended approach. 

IIFP used this approach for the process system hazard identification. The results of the 

unmitigated What-if scenarios are used directly as input to the development of the risk index. In 

addition, the hazard identification highlights potentially hazardous process conditions. As 

appropriate, IIFP assessed hazards individually for the potential impact on the discrete 

components of the process systems, as well as on a facility-wide basis for credible hazards from 

fires (e.g., external to the process system) and external events (e.g., seismic, severe weather). 

The applicant’s What-if techniques are an acceptable approach because it is consistent with 

NUREG-1513 (NRC, 2001). In addition, the accident sequences that were identified involve 

noncomplex systems and generally occur as a result of simple failure events. Therefore, the 

NRC staff concludes that the PHA method used by the applicant is acceptable for the 

identification of potential radiological, chemical, and facility hazards; potential accident 

sequences caused by process deviations; or other events internal to the facility, as well as 

credible external events, including natural phenomena that could lead to a loss of hazardous 

material confinement. 

3.3.2.4 Identify Accident Scenarios 

Credible accident scenarios or sequences are identified by analyzing initiating events. Using 

approved methods, the IIFP ISA team identified potential accident scenarios, including possible 

worst-case consequences; the causes (events that can initiate the accident); and controls to 

prevent the event or mitigate the consequences. Some controls are incorporated into the 

design features or administrative programs to provide defense-in-depth, in addition to the 

IROFS. Consequences of interest include radiological material releases; radiation exposures; 

chemical and toxic exposures from licensed material; or hazardous chemicals produced from 

licensed material, fires, and explosions. Hazards are defined to be materials, equipment, or 

energy sources with the potential to cause injury or illness to humans or adversely impact the 

environment. 

When analyzing accident scenarios, the IIFP ISA team considered process deviations, human 

errors, internal facility events, and credible external events, including natural phenomena. The 

ISA team used the Division of Fuel Cycle Safety and Safeguards’s (FCSS’s) Interim Staff 

Guidance (ISG)-08, “Natural Phenomena Hazards,” issued in 2005 (NRC, 2005), as guidance 

when evaluating natural phenomena hazards as initiating events. The team evaluated common 

mode failures and systems interactions that require preventive actions or control measures to 

prevent or mitigate accident scenarios. The team listed scenarios considered not credible. In 

addition to normal conditions, the team considered abnormal conditions, including startup, 

shutdown, maintenance, and process upsets. 

For each accident scenario, the team identified enabling conditions and conditional events that 

affect the outcome of the accident scenario (e.g., conditions that affect the likelihood of the 

scenario or could mitigate the consequences to either workers or the public) where appropriate. 

An enabling condition does not directly cause the scenario, but must be present for the initiating 

event to proceed to the consequences described. Enabling conditions are expressed as 
3-11 

probabilities and can reflect such things as the mode of operation (e.g., percent of operational 

online availability). 

The team also identified conditional events that affect the probability of the undesired outcome. 

These include probabilistic consideration of individual or administrative actions that would not be 

considered IROFS but would affect the overall likelihood of the accident. For example, if a 

scenario involves personal injury hazards, at least one worker must be present in the affected 

area at the time of the event for the injury to occur. Thus, the presence of workers in the 

affected area is a conditional modifier for a consequence involving personal injury. Another 

example of a conditional event is the probability that a worker can successfully evacuate from 

an area given that a hazard is present. 

In considering accident scenarios at the IIFP facility, it is necessary to determine which 

scenarios are considered not credible and which are credible. During the PHA, IIFP’s ISA team 

considered each accident scenario as credible, unless the scenario could be determined to be 

not credible. Section 3.2.5.5 of the LA (IIFP, 2012a) describes the criteria IIFP used to 

determine whether an accident scenario is credible. The NRC staff finds the applicant’s process 

for identifying accident scenarios consistent with the requirements of 10 CFR 70.61 and ISG-08 

(NRC, 2005). 

3.3.2.5 Determine Consequence Severity Level 

Table 3-1 of the LA (IIFP, 2012a) presents the radiological and chemical consequences severity 

limits of 10 CFR 70.61 for each of the accident consequence categories. Table 3-2 of the LA 

(IIFP, 2012a) provides information on the chemical dose limits specific to the IIFP facility. 

For each credible accident scenario identified, IIFP’s ISA team assigned a severity ranking for 

the consequences using the consequence severity rankings provided in Table 3-1 of the LA 

(IIFP, 2012a). Assigning a severity ranking allowed each accident scenario to be categorized in 

terms of the performance requirements outlined in 10 CFR 70.61(b) and 10 CFR 70.61(c). The 

severity ranking system is listed below: 

• A severity ranking of 3 corresponds to high consequences. 

• A severity ranking of 2 corresponds to intermediate consequences. 

• A severity ranking of 1 corresponds to low consequences. 

When estimating the possible worst-case consequences of an accident scenario, IIFP’s ISA 

team members used experience, guidance from NUREG/CR-6410, “Nuclear Fuel Cycle Facility 

Accident Analysis Handbook,” issued March 1998 (NRC, 1998), and best judgment. 

The severity of consequences is determined, both quantitatively and qualitatively, through a 

multitude of methods. Quantitative methods include source term and dispersion modeling. 

Qualitative methods may use as their basis worst-case assumptions or a comparison to similar 

events for which bounding conservative calculations have been made, or both. The 

consequence of concern is the chemo-toxic exposure to UF6, UF4, HF, uranium dioxide (UO2), 

and UO2F2. The applicant evaluated the dose consequences for each of the accident 

descriptions were evaluated and compared them to the 10 CFR 70.61 criteria for high (severity 

level 3) and intermediate (severity level 2) consequences. For a severity level of 1 (low), there 

is “no safety consequence of concern,” and no further analysis is required. The NRC staff finds 
3-12 

the applicant’s process to determine consequences of potential accidents consistent with 

Section 3.4.3.1(2)a of NUREG-1520 (NRC, 2002). 

3.3.2.6 Determine Unmitigated Likelihood 

IIFP determined the likelihood of an accident scenario occurring for the unmitigated case 

(unmitigated likelihood). Unmitigated likelihood is the likelihood or frequency that the initiating 

event or cause of the accident sequence occurs despite any actual or potential preventive or 

mitigating features. Therefore, this likelihood/frequency estimate assumes that none of the 

available safeguards or IROFS is available to perform its intended safety function. Table 3-4, 

“Unmitigated Likelihood Categories,” of the LA (IIFP, 2012a) qualitatively demonstrates the 

likelihood of occurrence limits drawn from 10 CFR 70.61 for each of the three likelihood 

categories. 

The IIFP’s ISA team assigned a likelihood level for each accident scenario using the defined 

categories in Tables 3-5, “Event Likelihood Categories,” and 3-6, “Determination of Likelihood 

Category,” of the LA (IIFP, 2012a). When assigning a likelihood category, the team relied on 

process knowledge, accident scenario information, operating history, and manufacturer/product 

information to determine which category of likelihood was appropriate. For accident scenarios 

in which multiple initiating events were identified, the team estimated the likelihood for the most 

credible initiating event. This ensured that the accident scenario was screened using the most 

conservative estimate of risk. The NRC staff finds the applicant’s process to determine 

unmitigated likelihoods of potential accidents consistent with Section 3.4.3.1(2)a of 

NUREG-1520 (NRC, 2002). 

3.3.2.7 Determine Unmitigated Risk 

The NRC staff finds that the scope of the ISA Summary includes the credible accident scenarios 

identified for the IIFP facility. These scenarios have the capability of producing conditions that 

fail to meet the performance requirements of 10 CFR 70.61(b), 10 CFR 70.61(c), or 

10 CFR 70.61(d). For each credible accident scenario, IIFP’s ISA team used the severity 

category ranking and unmitigated likelihood level to assign an unmitigated risk level. (The 

unmitigated risk level is determined from the product of the severity category and the 

unmitigated likelihood category.) The ISA team used the risk matrix in Table 3-7, “Unmitigated 

Risk Assignment Matrix,” of the LA (IIFP, 2012a) to determine the unmitigated risk. The 

unmitigated risk associated with each accident scenario indicates the relative importance of the 

associated controls. Accident scenarios in which the consequences and likelihoods yielded an 

unacceptable risk index required further evaluation to determine the IROFS and mitigated risk, 

as described in Section 3.2.5.8 of the LA (IIFP, 2012a). If the unmitigated risk was less than or 

equal to 4, the unmitigated risk was acceptable and no further action was required. The 

applicant updated the What-if table to reflect the conclusion of no further action. The NRC staff 

finds the applicant’s process to determine unmitigated risks of potential accidents consistent 

with Section 3.4.3.1(2)a of NUREG-1520 (NRC, 2002). 

3.3.2.8 Items Relied on for Safety and Risk Development 

For each accident scenario having an unacceptable unmitigated risk index (greater than 4), the 

ISA team defined the IROFS and determined the mitigated likelihood. The team used the 

unmitigated initiating event frequency and the failure probability of each IROFS to determine the 

mitigated likelihood. 
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The risk analysis presents an accident evaluation, including a detailed discussion concerning 

the selection of initiating events, IROFS, and the evaluation of the accident sequences. The risk 

analysis provides sufficient background and operational information to understand and examine 

accident scenarios that result in undesired outcomes for each initiating event. Each risk 

analysis provides details concerning an accident scenario’s quantification, including (1) the 

method used, (2) the initiating-event frequency determination, (3) the IROFS credited to prevent 

or mitigate the initiating event or events being analyzed, (4) the failure probabilities for the 

credited IROFS, and (5) the overall likelihood estimates. The risk analyses are controlled 

documents, and the Configuration Management Program described in the LA, Chapter 10, 

ensures that they remain up to date. The ISA Summary presents the results from each risk 

analysis. 

The mitigated likelihood of the accident scenario occurring with the preventive or mitigating 

IROFS in place must meet the provisions of 10 CFR 70.61, which require that unacceptable 

consequences be limited. The values of the index numbers for an accident scenario, depending 

on the number of events involved, are added to obtain a total likelihood index, T. Accident 

scenarios are then assigned to one of the three likelihood categories of the risk matrix, 

depending on the value of the likelihood index, in accordance with Table 3-5 of the LA 

(IIFP, 2012a). 

Safeguards are design features or administrative programs that provide defense-in-depth, but 

are not IROFS and are not credited with preventing or mitigating accident scenarios. 

IIFP committed to identify IROFS as a part of the ISA process and include the identification of 

the IROFS in the ISA Summary prepared and maintained for the IIFP facility. The IROFS are 

defined in such a way as to delineate their boundaries, describe the characteristics of the 

preventive/mitigating function, and identify the applicable assumptions and conditions. 

The NRC staff finds the applicant’s process to identify IROFS and assign reliability indices to 

these IROFS is consistent with Section 3.4.3.1(2)a of NUREG-1520 (NRC, 2002). 

3.3.3 Integrated Safety Analysis Integration and Team 

The ISA is intended to provide assurance that the potential failures, hazards, accident 

descriptions, scenarios, and IROFS have been investigated in an integrated fashion so as to 

adequately consider common-mode and common-cause situations. Included in this integrated 

review is the identification of IROFS functions that may simultaneously be beneficial and 

harmful with respect to different hazards, as well as interactions that the previously completed 

risk analysis might not have considered. This review is intended to ensure that the designation 

of one IROFS does not negate the preventive or mitigation function of another IROFS. The 

applicant’s ISA team performed an integrated review during the process hazard review and an 

overall integration review after the nodes were completed. The following items warranted 

special consideration during the integration process evaluation: 

• common-mode failures and common-cause situations 

• support system failures (e.g., loss of electrical power or water), the failure of which could 

have a simultaneous effect on multiple systems 

• divergent impacts of IROFS so as to ensure that the negative impacts of an IROFS, if 

any, do not outweigh the positive impacts (i.e., to ensure that the application of an 
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IROFS for one safety function does not degrade the defense-in-depth of an unrelated 

safety function) 

• other safety and mitigating factors that do not achieve the status of IROFS that could 

impact system performance 

• identification of scenarios, events, or event descriptions with multiple impacts, such as 

impacts on chemical, fire, or radiation safety (e.g., a flood which could cause both a loss 

of confinement and active safeguards) 

• potential interactions between processes, systems, areas, and buildings; any 

interdependence of systems or potential transfer of energy or materials 

• major hazards or events that tend to be common-cause situations leading to interactions 

between processes, systems, buildings, and the like 

The NRC staff finds the applicant’s process for integrating the ISA consistent with 

Section 3.4.3.1(2)a of NUREG-1520 (NRC, 2002). 

The applicant’s team with expertise in engineering, process safety, safety analysis, and facility 

process operations performed and will maintain the ISA. Specifically, the team that performed 

the ISA included individuals with experience in nuclear facility safety, radiological safety, PHA, 

safety analysis and risk assessment, fire safety, chemical process safety, operations and 

maintenance, and ISA methods. The NRC staff finds the applicant’s ISA team used for 

performing and maintaining the ISA to be consistent with Section 3.4.3.2(5) of NUREG-1520 

(NRC, 2002). 

3.4 Integrated Safety Analysis Summary Review Conclusion

The NRC staff finds that the applicant’s maintenance of process safety information is consistent 

with the requirements of 10 CFR 70.62(b) and the guidance of NUREG-1520 (NRC, 2002). The 

NRC staff further finds that the applicant’s commitment to conduct and maintain an ISA is in 

accordance with the requirements of 10 CFR 70.62(c)(1) and the guidance in NUREG-1520 

(NRC, 2002). 

During construction of IIFP facility, the NRC staff will implement a construction inspection 

program to verify that IIFP complies with the LA. This inspection program will conclude with a 

readiness review prior to operations. The NRC staff have provided two license conditions in 

Section 1.2.3.5.3 of this SER that defines the scope of the construction inspection program and 

readiness review. 

The NRC staff finds the ISA methodology to be complete because its use of the appropriate 

accident identification methodology from NUREG-1513 (NRC, 2001). The NRC staff finds the 

consequence determinations to be acceptable and in accordance with the guidance in 

NUREG/CR-6410 (NRC, 1998). The NRC staff also evaluated the consequence determination 

methodology and determined that it provides reasonable assurance that determination of 

consequences that meet the performance requirements of 10 CFR 70.61(b) and 

10 CFR 70.61(c) will result from its use. The NRC staff considers the likelihoods to have been 

derived using acceptable methods and to comply with acceptable definitions of “not unlikely,” 

“unlikely,” and “highly unlikely,” in accordance with the requirements of 10 CFR 70.61 and the 
3-15 

guidance in NUREG-1520 (NRC, 2002). The NRC staff finds that these descriptions (see 

Section 3.3.2.4 of this SER) conform to the guidance provided in NUREG-1520 (NRC, 2002) 

and meet the requirements of 10 CFR 70.65(b)(5). 

The NRC staff finds that the likelihood evaluation, including the definition of “not credible,” 

conforms to the guidance provided in NUREG-1520 (NRC, 2002) and is acceptable (see 

Section 3.3.2.3 of this SER). The NRC staff also finds that the determination of chemical 

consequences (see Section 3.3.2.4 of this SER) conforms to the guidance provided in 

NUREG-1520 (NRC, 2002) and is acceptable. 

On July 6–8, 2010, the NRC staff conducted a site visit in Oak Ridge, TN, with IIFP and its 

contractor. The purpose of the visit was to allow the NRC staff to conduct a horizontal and 

vertical slice review of the applicant’s onsite ISA documents and to discuss draft requests for 

additional information (RAIs) regarding the safety review. 

After the NRC staff reviewed the ISA Summary submitted with the LA and identified RAI 

questions and comments, the NRC staff conducted a horizontal and vertical slice review of the 

ISA and PHA. Following guidance in NUREG-1520 (NRC, 2002), for the horizontal review, the 

NRC staff reviewed the process used to identify and evaluate credible accidents. For the 

vertical slice review, the NRC staff examined the applicant’s analysis of certain selected 

accidents and the adequacy of the IROFS. The reviewers confirmed that the applicant’s 

methodology to determine the consequences of the accident sequences identified in the PHA is 

consistent with the information presented in the ISA Summary. The horizontal review revealed 

that, at the time of the site visit, the ISA continued to undergo updates, in part because of the 

continuing development of specific design information. The vertical review examined several 

accident sequences in depth, including the applicant’s What-if analysis, accident sequences, 

methods for assigning likelihood and consequence levels, and designation of IROFS. The 

chemical, fire, and ISA reviewers selected several accident sequences which the applicant 

described in detail. The NRC staff requested updates to the ISA Summary, which IIFP provided 

in its RAI reply (IIFP, 2011b). 

Many hazards and potential accidents can result in unintended exposure of persons to radiation, 

radioactive materials, or toxic chemicals incident to the processing of licensed materials. The 

NRC staff finds that the applicant has performed an ISA to identify and evaluate those hazards 

and potential accidents as required by the regulations. The NRC staff has reviewed the ISA 

Summary and other information, and finds that it provides reasonable assurance that the 

applicant has identified IROFS and established engineered and administrative controls to 

ensure compliance with the performance requirements of 10 CFR 70.61. Specifically, the NRC 

staff finds that the ISA results, as documented in the ISA Summary, provide reasonable 

assurance that the IROFS, the management measures, and the applicant’s programmatic 

commitments will, if properly implemented, make all credible intermediate-consequence 

accidents unlikely, and all credible high-consequence accidents highly unlikely. 

In accordance with 10 CFR 70.66(a), the NRC staff has determined that IIFP complied with the 

requirements of 10 CFR 70.21, “Filing”; 10 CFR 70.22, “Contents of Applications”; 

10 CFR 70.23, “Requirements for the Approval of Applications”; and 10 CFR 70.60, 

“Applicability,” through 10 CFR 70.65 relating to Subpart H. 

 
3-16 

3.5 References

(IIFP, 2012a) International Isotopes Fluorine Products, Inc., “Fluorine Extraction Process and 

Depleted Uranium Deconversion Plant (FEP/DUP) License Application, Revision B,” May 2012, 

Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML12123A245. 

(IIFP, 2012b) International Isotopes Fluorine Products, Inc., “ISA Summary Rev. B for IIFP,” 

May 2012, Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML12123A245. 

(NRC, 2007) U.S. Nuclear Regulatory Commission, Staff Requirements Memorandum to SECY 

07 0146, “Regulatory Options for Licensing New Uranium Conversion and Depleted Uranium 

Deconversion Facilities,” October 10, 2007. 

(NRC, 2005) U.S. Nuclear Regulatory Commission, “Natural Phenomena Hazards,” 

FCSS ISG-08, 2005. 

(NRC, 2002) U.S. Nuclear Regulatory Commission, “Standard Review Plan for the Review of a 

License Application for a Fuel Cycle Facility,” NUREG 1520, Rev. 0, March 2002. 

(NRC, 2001) U.S. Nuclear Regulatory Commission, “Integrated Safety Analysis Guidance 

Document,” NUREG-1513, May 2001. 

(NRC, 1998) U.S. Nuclear Regulatory Commission, “Nuclear Fuel Cycle Facility Accident 

Analysis Handbook,” NUREG/CR-6410, March 1998. 

(NRC, 1988) U.S. Nuclear Regulatory Commission, “A Regulatory Analysis on Emergency 

Preparedness for Fuel Cycle and Other Radioactive Material Licensees,” NUREG-1140, 1988 

(reprinted August 1991). 
4-1 

4.0 RADIATION PROTECTION 

The purpose of this review is to determine whether the International Isotopes Fluorine Products, 

Inc. (IIFP or the applicant), Radiation Protection (RP) Program is adequate to protect the 

radiological health and safety of workers and to comply with the associated regulatory 

requirements in the following parts of Title 10 of the Code of Federal Regulations (10 CFR): 

• 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 40, “Domestic Licensing of Source Material” 

• 10 CFR Part 70, “Domestic Licensing of Special Nuclear Material” 

Chapter 9 of this Safety Evaluation Report (SER) discusses public and environmental 

protection. 

4.1 Regulatory Requirements

The RP regulatory requirements are located in 10 CFR Part 20. The regulations in 

10 CFR 40.32, “General Requirements for Issuance of Specific Licenses,” apply to the 

organization, qualifications and written procedures for the RP staff. The regulations in 

10 CFR 19.12, “Instructions to Workers,” apply to the Radiation Safety Training Program. The 

regulations in 10 CFR 70.61, “Performance Requirements,” apply to restricting radiological risk 

associated with potential process accidents involving radioactive materials. 

4.2 Regulatory Guidance and Acceptance Criteria

Sections 4.4.1.3, 4.4.2.3, 4.4.3.3, 4.4.4.3, 4.4.5.3, 4.4.6.3, 4.4.7.3, and 4.4.8.3 of NUREG-1520, 

Revision 0, “Standard Review Plan for the Review of a License Application (LA) for a Fuel Cycle 

Facility” issued March 2002 (NRC, 2002), outlines the acceptance criteria for the U.S. Nuclear 

Regulatory Commission (NRC) staff’s review of the RP Program. 

4.3 Staff Review and Analysis

4.3.1 Radiation Protection Program Implementation 

The NRC staff reviewed the applicant’s RP Program implementation against the acceptance 

criteria in Section 4.4.1.3 of NUREG-1520, Revision 0 (NRC, 2002). The following discussion 

summarizes the NRC staff’s analysis as to whether the information provided by the applicant in 

the LA (IIFP, 2012a) meets the criteria. 

In Section 4.1 of the LA (IIFP, 2012a), the applicant stated that the RP Program will meet the 

requirements of 10 CFR Part 20, Subpart B, by committing that the program will be consistent 

with the guidance provided in Regulatory Guide 8.2, “Guide for Administrative Practice in 

Radiation Monitoring,” issued February 1973 (NRC, 1973). The applicant further committed to 

maintaining occupational exposures as low as reasonably achievable (ALARA) through 

exposure monitoring consistent with the regulations in 10 CFR Part 20, frequent interactions 
4-2 

between the Radiation Safety Committee and operations personnel, and annual RP Program 

assessments with senior management. 

The applicant will establish annual exposure goals for occupationally exposed personnel to 

ensure that personnel doses received will be below the limits specified in 10 CFR 20.1201. The 

RP Program content and implementation will be reviewed annually, at a minimum. In addition, 

the applicant will establish controls such that no member of the public is expected to receive a 

total effective dose equivalent (TEDE) in excess of 0.25 millisieverts per year (mSv/yr) 

(25 millirems per year [mrem/yr]). 

The applicant outlined the operations organizational structure in Section 2.1.4 of the LA (IIFP, 

2012a) and described the RP Program structure and defined the responsibilities of key program 

personnel in Section 4.1.1 of the LA (IIFP, 2012a). The Chief Operations Officer (COO) will 

have the overall responsibility of ensuring that facility operations are conducted in a manner that 

protects the employee, the environment, and the public from radiological, chemical, and 

industrial hazards; and that these operations will be carried out in accordance with all applicable 

regulations, licenses, and permits. The COO will perform his or her duties in accordance with 

written policies and procedures. The COO provides for safety and control of operations and 

protection of the environment by delegating and assigning responsibility to qualified plant and 

line supervisors. 

The Environmental, Safety and Health (ESH) Manager will report to the COO, and will interact 

with all IIFP managers on matters of EHS policies, regulatory requirements, plant safety and 

environmental compliance. The ESH Manager will be responsible for directing activities to 

ensure the facility complies with appropriate rules, regulations, and codes. This includes ESH 

activities associated with RP, chemical safety, environmental protection, industrial hygiene, 

industrial safety, emergency preparedness, security, regulatory affairs, and licensing. The ESH 

Manager will work with other managers and supervisors of the plant to ensure consistent 

interpretations of the requirements, perform independent reviews, and support facility and 

operations change control reviews. The ESH organization provides independent oversight of 

plant operations. The ESH manager will have the responsibility and authority to elevate any 

ESH or security-related issue to the IIFP President and Chief Executive Officer (CEO). 

The Radiation Protection Manager (RPM) will be administratively independent of operations and 

will report directly to the ESH Manager. The RPM will also have the authority to report to the 

COO any unresolved concerns related to ESH and RP. The RPM will be responsible for 

effectively implementing the RP Program and will ensure that the facility is staffed with suitably 

trained RP personnel and that sufficient resources are provided to implement an effective 

program. The RPM will approve restart of any operation that gets shut down by the RP function 

or as a result of RP concerns. 

The RP staff, including technicians and support personnel, will report to the RPM. Major 

responsibilities of the RPM and the RP staff will include, but not be limited to, the following: 

• Establish and maintain the RP Program, procedures, and training. 

• Conduct radiation and contamination monitoring and control programs. 

• Evaluate radiation exposures of employees, contractor personnel, and visitors and 

ensure the maintenance records and reporting of results. 
4-3 

• Establish and maintain the ALARA Program, including being a key member of the 

ALARA Safety Committee. 

• Evaluate the integrity and reliability of radiation detection instruments. 

• Provide support for integrated safety analyses (ISAs) and configuration control. 

Section 4.1.2 of the LA (IIFP, 2012a) addresses the independence of the RP Program. The RP 

Program will be independent of operations because both the RP and ESH Managers function 

independently of operations and have the authority to elevate concerns within the organization. 

As stated in Section 4.1.3 of the LA (IIFP, 2012a), the applicant commits that the ALARA 

Committee will review, at least annually, the content and implementation of the RP Program. 

The review will consider facility changes, new technologies, and other process enhancements 

that could improve overall program effectiveness. The NRC staff finds that the commitments in 

the LA (IIFP, 2012a) satisfactorily address the acceptance criteria in Section 4.4.1.3 of 

NUREG-1520, Revision 0 (NRC, 2002). 

4.3.2 As Low As Reasonably Achievable Program 

The NRC staff reviewed the applicant’s ALARA Program commitments against the acceptance 

criteria in NUREG-1520, Revision 0, Section 4.4.2.3 (NRC, 2002). The following discussion 

summarizes the NRC staff’s analysis as to whether the information provided by the applicant in 

the LA (IIFP, 2012a) meets the criteria. 

Section 4.2 of the LA (IIFP, 2012a) addresses the proposed ALARA Program, which will 

function as a subset of the RP Program. The objective of the program will be to make every 

reasonable effort to maintain facility exposures to radiation as far below the dose limits of 

10 CFR 20.1201 as is practical and to maintain radiation exposures to members of the public 

below the dose limits of 10 CFR 20.1101(d). The design and implementation of the ALARA 

Program will be consistent with guidance provided in Regulatory Guide 8.2 (NRC, 1973); 

Regulatory Guide 8.13, Revision 3, “Instruction Concerning Prenatal Radiation Exposure,” 

issued June 1999 (NRC, 1999a); Regulatory Guide 8.29, Revision 1, “Instruction Concerning 

Risks from Occupational Radiation Exposure,” issued February 1996 (NRC, 1996); and 

Regulatory Guide 8.37, “ALARA Levels for Effluents from Materials Facilities,” issued July 1993 

(NRC, 1993a). 

The ALARA Program will include the following features: 

• management commitment, demonstrated through a written policy statement, 

procedures, other directives, and periodic management reviews 

• formal program audits, conducted on at least an annual basis 

• well-supervised and defined RP capability, including appropriate supervisors and 

technicians, with all personnel onsite having the authority to stop work as needed to 

ensure that appropriate safety precautions are observed 

• appropriate training for the workforce, including training consistent with the requirements 

of 10 CFR 19.12 and incorporating appropriate portions of the guidance provided in 
4-4 

Regulatory Guides 8.13, Revision 3 (NRC, 1999a) and 8.29, Revision 1 (NRC, 1996) 

• appropriate authority vested in RP personnel including stop work authority 

• consideration of the need for plant modifications as warranted for reducing exposures 

and doses to personnel. 

The applicant will implement documented RP Program policies to ensure that the ALARA goal is 

met. Procedures will incorporate the ALARA philosophy into routine operations and ensure 

exposures will be maintained below the levels identified in 10 CFR 20.1101 levels (i.e., 

10 mrem/yr to a member of the public through airborne emissions). As stated previously, the 

applicant will also establish controls such that no member of the public will receive a TEDE in 

excess of 0.25 mSv/yr (25 mrem/yr). Radiological controlled areas (RCAs) will be established 

within the facility and identified through signs, ropes, gates, fences, or other visible means. 

Each zone will have specific entry requirements, survey requirements, and dosimetry 

requirements. The establishment of these areas will support the ALARA commitment to 

minimize the spread of contamination and reduce unnecessary exposure of personnel to 

radiation. 

The applicant will have an ALARA Safety Committee that supports the Facility Safety Review 

Committee. The ALARA Safety Committee will consist of key members of facility management, 

supervisory staff, and workforce and will meet periodically on a frequency established in the RP 

ALARA Program. The ALARA Safety Committee will use the guidance provided in Regulatory 

Guide 4.21, “Minimization of Contamination and Radioactive Waste Generation: Life-Cycle 

Planning,” issued June 2008 (NRC, 2008); Regulatory Guide 8.10, Revision 1, “Operating 

Philosophy for Maintaining Occupational Radiation Exposures as Low as Is Reasonably 

Achievable,” issued May 1977 (NRC, 1977); and Regulatory Guide 8.37 (NRC, 1993a) to 

formulate facility operating philosophy in reducing exposures. Membership of the ALARA 

Safety Committee includes the following individuals: 

• the COO 

• the RP Manager 

• selected department managers 

• the ESH Manager 

• selected supervisors and hourly personnel 

The ALARA Program is intended/expected to facilitate interaction between RP and operations 

personnel. The ALARA Safety Committee, comprising staff members responsible for RP and 

operations personnel, including hourly workers, will be utilized in achieving this goal. The scope 

of the ALARA Safety Committee’s activities will include at a minimum, annual review of the 

following: 

• site radiological operating performance, including trends in airborne concentrations, 

personnel exposures, and environmental monitoring results 

• operations and exposure records to determine where exposures may be reduced 
4-5 

• employee training and methods for utilizing information on the job to keep exposure 

ALARA 

• potential modifications of procedures and equipment when changes will reduce 

exposure at reasonable cost 

Specific goals of the ALARA Program will include maintaining occupational exposures, as well 

as environmental releases, as far below regulatory limits as is reasonably achievable. For 

environmental effluents, ALARA goals will be initially set to 20 percent of 10 CFR Part 20, 

Appendix B. The design and operation of the facility will also incorporate the ALARA concept. 

The size and number of areas with higher dose rates will be minimal. Written procedures will 

incorporate instruction such that time spent in higher dose rate areas will be controlled, and 

projects will be evaluated to ensure that workers receive minimum radiological exposure. Areas 

where personnel spend significant amounts of time will be designed to maintain the lowest dose 

rates reasonably achievable. 

Based on the review summarized above, the NRC staff finds that the commitments in the LA 

(IIFP, 2012a) satisfactorily address the acceptance criteria in Section 4.4.2.3 of NUREG-1520, 

Revision 0 (NRC, 2002). 

4.3.3 Organization and Personnel Qualifications 

The NRC staff reviewed the applicant’s organization and personnel qualifications against the 

acceptance criteria in NUREG-1520, Revision 0, Section 4.4.3.3 (NRC, 2002). The following 

discussion summarizes the NRC staff’s analysis as to whether the information provided by the 

applicant in the LA (IIFP, 2012a) meets the criteria. 

Several sections of the LA (IIFP, 2012a) discuss organization and personnel qualifications, 

although most of the information is repeated in Section 4.3. The applicant stated that the RP 

staff will be assigned responsibility for implementation of the RP Program functions. The facility 

will only employ suitably trained RP personnel. Staffing will be consistent with the guidance 

provided in Regulatory Guide 8.2 (NRC, 1973) and Regulatory Guide 8.10, Revision 1 

(NRC, 1977). 

The ESH Manager will be responsible for establishing and overseeing the RP, Industrial Safety 

and Industrial Hygiene, Environmental Protection, Fire Protection, and Emergency 

Preparedness/Security Programs to ensure compliance with applicable Federal, State, and local 

regulations and requirements. The ESH Manager will report to the COO, but also interacts with 

all managers on matters of ESH policies, regulatory requirements, plant safety, and 

environmental compliance. In addition, the ESH Manager has the authority and responsibility to 

elevate any ESH concerns to the corporate EHS manager and the IIFP President. The ESH 

Manager will have, as a minimum, a bachelor’s degree in engineering, science, or a related field 

and 5 years of experience of ESH activities at chemical, radiological, or nuclear facilities. 

The RPM will report directly to the ESH Manager and has the responsibility for establishing and 

implementing the RP Program. These duties will include the training of personnel in the use of 

equipment, control of radiation exposure of personnel, continuing evaluation and determination 

of the radiological status of the facility, and conducting the Radiological Environmental 

Monitoring Program. The RPM will have, as a minimum, a bachelor’s degree in engineering or 
4-6 

a scientific field and a minimum of 5 years of responsible experience that includes assignments 

involving responsibility for RP and the application and direction of RP programs. 

Staff health physicists will have, as a minimum, a bachelor’s degree in engineering or a 

scientific field and experience commensurate with health physics and RP duties. Staff radiation 

control technicians will have a high school diploma and experience commensurate with RP 

duties. 

The facility organization chart, reproduced as Figure 4-1 of the LA (IIFP, 2012a), establishes 

clear organizational relationships among the RP staff and the other facility line managers. In 

matters involving radiological protection, the RPM will have the responsibility and authority to 

elevate any radiation safety or environmental issue to the COO. The RPM will be skilled in the 

interpretation of RP data and regulations and familiar with the operation of the facility and RP 

concerns relevant to the facility. The RPM will be a resource for radiation safety management 

decisions. 

RP technicians, engineers, and supervisors will perform the functions of assisting and guiding 

workers in the radiological aspects of the job. These individuals will have the responsibility and 

authority to stop work or mitigate the effect of an activity, if it is suspected that the initiation or 

continued performance of a job, evaluation, or test will result in the violation of approved RP 

requirements. 

Based on the review as summarized above, the NCR staff finds that the commitments in the LA 

(IIFP, 2012a) satisfactorily address the acceptance criteria in Section 4.4.3.3 of NUREG-1520, 

Revision 0 (NRC, 2002). 

4.3.4 Written Procedures 

The NRC staff reviewed the applicant’s written procedure commitments against the acceptance 

criteria in NUREG-1520, Revision 0, Section 4.4.4.3 (NRC, 2002). The following discussion 

summarizes the NRC staff’s analysis as to whether the information provided by the applicant in 

the LA (IIFP, 2012a) meets the criteria. 

Sections 2.3.4, 4.4, 4.4.1, and 11.4 of the LA (IIFP, 2012a) discuss written procedure 

commitments. The applicant stated that it will use approved written procedures to conduct 

operations at the facility involving licensed materials. The applicant will develop or modify 

procedures through a formal process incorporating change controls, as described in 

Section 11.1 of the LA. The procedure process will utilize nine basic elements to accomplish 

procedure development, review, approval, and control. The LA, Section 11.4.2 (IIFP, 2012a), 

identifies these elements as (1) identification, (2) development, (3) verification, (4) review and 

comment resolution, (5) approval, (6) validation, (7) issuance, (8) change control, and 

(9) periodic review. Changes to work procedures, including safety requirements, will be 

reviewed with production personnel by their immediate supervisor or delegate. 

The RP staff will prepare, review, and approve RP procedures to carry out activities related to 

the RP Program. Approved written procedures will control RP activities to ensure that the 

activities are implemented in a safe, effective, and consistent manner. RP procedures will be 

reviewed and revised, as necessary, to incorporate facility or operational changes or changes to 

the ISA. The RP staff will prepare draft procedures that affected personnel will review to ensure 

that the procedures are appropriate and reasonable to implement. The RPM (or designee) will 

review and approve final RP procedures, as well as proposed revisions to RP procedures. 
4-7 

The RWP system will administer non-routine activities, particularly those performed by non-IIFP 

employees and not covered by approved written procedures. The RWP will describe the work 

to be performed and define the authorized activities. The RWP will specify the necessary 

radiation safety controls, as appropriate, to include personnel monitoring devices, attendance of 

RP staff, protective clothing, respiratory protective equipment, special air sampling, and 

additional precautionary measures to be taken. The RWP will also describe the radiological 

conditions in the immediate work area covered by the RWP. The RPM or designee must 

approve the RWP. The designee must meet the qualification requirements of the RPM. RWPs 

will have a predetermined period of validity with a specified expiration or termination time. 

Standing RWPs may be issued for routinely performed activities, such as tours of the plant. 

Before commencing work that requires an RWP, employees performing the job will have to 

review the RWP and document their review. An RP technician will monitor the work as 

required. RWPs will be available to workers for review at any time and will include expiration 

dates. An RP technician or the RPM (or designee) reviews the status of issued RWPs on a 

periodic basis. An RWP will be closed when the applicable work activity for which it is written is 

complete and terminated. Copies of RWPs and any associated records will be kept for the life 

of the facility. 

Based on the review summarized above, the NRC staff finds that the commitments in the LA 

(IIFP, 2012a) satisfactorily address the acceptance criteria in Section 4.4.4.3 of NUREG-1520, 

Revision 0 (NRC, 2002). 

4.3.5 Training 

The NRC staff reviewed the applicant’s training commitments against the acceptance criteria in 

NUREG-1520, Revision 0, Section 4.4.5.3 (NRC, 2002). The following discussion summarizes 

the NRC staff’s analysis as to whether the information provided by the applicant in the LA (IIFP, 

2012a) meets the criteria. 

The applicant discussed its training commitments in Section 4.5 of the LA (IIFP, 2012a), as well 

as in Sections 11.3.7.1, 11.3.7.2, and 11.3.8. The applicant stated that RP training will comply 

with the requirements in 10 CFR 19.12 and 10 CFR 20.2110 and will take into consideration a 

worker’s normally assigned work activities. Basic RP training will cover the following topics: 

• radiation safety principles, policies, and procedures 

• radiation hazards and health risks 

• correct handling of radioactive materials 

• location of and adherence to RP procedures 

• minimization of exposures to radiation and radioactive materials 

• contamination control 

• access and egress controls 

• monitoring for internal and external exposures 
4-8 

• ALARA and exposure limits 

• exposure monitoring methods and instrumentation 

• personal and area dosimetry 

• donning and doffing of personal protective equipment 

• emergency response. 

Additional topics can be included if applicable to the individual’s job responsibilities. 

The applicant will establish training programs for the various types of job functions (e.g., 

production, maintenance, RP technician, and contractor personnel) commensurate with RP 

responsibilities associated with each such position. Trained personnel will escort visitors to the 

controlled access area (CAA). 

Individuals attending training sessions must pass an initial examination covering the training 

contents to ensure the understanding and effectiveness of the training. The periodicity of 

refresher training will depend on the worker’s responsibilities; however, the basic refresher 

training will occur annually (not to exceed 15 months) and will include an exam. Training 

requirements will be documented and tracked for employees. The applicant will manage and 

store training records in accordance with 10 CFR 20.2110. Audits and assessments of 

production and maintenance personnel responsible for following the requirements related to the 

topics in the training they have received will also be used to evaluate the effectiveness of the 

training programs. 

Since contractor employees will perform diverse tasks in the CAA, training for these employees 

will be designed to address the type of work they perform. In addition to applicable radiation 

safety topics, training contents may include RWPs; special bioassay sampling; and special 

precautions for welding, cutting, and grinding in the CAA. The training manager will assign 

instructors to conduct these training programs who have the necessary knowledge to address 

chemical safety and RP. The applicant will maintain records of the training programs. 

Individuals requiring unescorted access to the CAA will receive annual continuing education 

training. Production personnel will be further instructed in the specific safety requirements of 

their work assignments by qualified personnel during on-the-job training. Employees must 

demonstrate an understanding of work assignment requirements based on observations by 

qualified personnel before working without direct supervision. Changes to work procedures, 

including safety requirements, will be reviewed with production personnel by their immediate 

supervisor or delegate. 

The RPM will review the contents of the Radiation Safety Training Program used by IIFP 

biannually. The review will address changes which have occurred in policies, procedures, and 

requirements, as well as changes to the ISA. 

Periodically, the training program will be systematically evaluated to measure the program's 

effectiveness in producing competent employees. The trainees will be encouraged to offer 

feedback after completion of classroom training sessions to provide data for program 
4-9 

improvements. Training program evaluations will identify program strengths and weaknesses, 

determine whether the program content matches current job needs, and determine whether 

corrective actions are needed to improve the program’s effectiveness. The training organization 

will be responsible for leading the training program evaluations. The RPM will review the 

evaluation information and implement changes in the training program as necessary. 

Based on the review as summarized above, the NRC staff finds that the commitments in the LA 

(IIFP, 2012a) satisfactorily address the acceptance criteria in Section 4.4.5.3 of NUREG-1520, 

Revision 0 (NRC, 2002). 

4.3.6 Ventilation and Respiratory Protection Programs 

The NRC staff reviewed the applicant’s Ventilation and Respiratory Protection Program 

commitments against the acceptance criteria in NUREG-1520, Revision 0, Section 4.4.6.3 

(NRC, 2002). The following discussion summarizes the NRC staff’s review as to whether the 

information provided by the applicant in the LA (IIFP, 2012a) meets the criteria. 

The applicant discussed ventilation commitments in Section 4.6.1 of the LA (IIFP, 2012a) and 

stated that the confinement of uranium will be a design requirement for the facility. Areas where 

uranium will be processed with the potential for dusts, mists, or fumes and areas where toxic 

chemicals are processed or produced are provided with dust collection or scrubber systems, or 

both, to maintain exposures ALARA. The design of building ventilation systems in process 

areas and control rooms will be sized with adequate flows and pressure differentials to ensure 

that potential airborne concentrations of radioactivity do not exceed derived air concentration 

(DAC) values, calculated from the dose coefficients specified in the International Commission 

on Radiological Protection (ICRP) Publication 68, “Dose Coefficients for Intakes by Workers” 

(ICRP, 1994). The applicant has requested an exemption to utilize the ICRP-68 values as 

opposed to the values presented in Appendix B, “Annual Limits on Intake (ALIs) and Derived Air 

Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations; 

Concentrations for Release to Sewerage,” to 10 CFR Part 20, “Standards for Protection against 

Radiation,” in Section 1.5 of the LA (IIFP, 2012a). The Commission has previously directed the 

NRC staff to approve the use of ICRP-68 on a case-by-case basis in the staff requirements 

memorandum, dated April 21, 1999, associated with SECY-99-0077, “To Request Commission 

Approval to Grant Exemptions from Portions of 10 CFR Part 20,” dated March 12, 1999. The 

ventilation system will also be designed so that air will flow from areas of low contamination 

potential towards areas of higher contamination potential to minimize the spread of 

contamination. 

The ventilation program, radiation detectors and alarms, process vents, and associated 

containment systems will be checked routinely as part of the operating process controls and 

Preventive Maintenance Program. Operations and maintenance relative to the Ventilation 

Program, including calibrations, change management, measurements, and analysis, will be 

performed using approved written procedures. 

Several measures will be in place to ensure effective operation of the ventilation control 

systems. Differential pressure will be monitored and alarmed for high-efficiency particulate air 

filters used for control rooms in which uranium is processed. Operating procedures will specify 

limits and set points on differential pressure consistent with manufacturer’s recommendations. 

Filters will be changed if they fail to function properly or if the differential pressure exceeds the 

manufacturer’s ratings. 
4-10 

Dust collector units in the depleted uranium tetrafluoride (DUF4) and fluorine extraction process 

buildings will be monitored and alarmed if the differential pressure exceeds established limits. 

Operating procedures will specify limits and set points for acceptable differential pressures and 

uranium sample results. Operating procedures will also specify that at least two dust collector 

units will be operated in series; if not, the process system being serviced by the dust collectors 

must be placed in a shutdown or standby mode. 

Written procedures, approved by the Plant Engineering and Maintenance Manager and the 

RPM (or designated alternates), will specify filter and dust collector inspection, testing, 

maintenance, and change-out criteria. Change-out frequency will be based on considerations of 

filter loading, operating experience, differential pressure data, and any monitoring data that 

exceeds set administrative control limits. 

Pressures will be continuously monitored and controlled for the plant offgas scrubbing system 

and across the process system vented to the scrubbing system. Limits will be set to ensure 

adequate safety margin of pressure controls for the vent gas plant scrubbing system. Operation 

procedures and operator aids will also provide for corrective response when alarms are 

received relative to the system pressure controls. 

Airflow rates at exhausted enclosures and close-capture points related to uranium processing 

and handling areas, when in use, will be adequate to preclude escape of airborne uranium and 

minimize potential for intake by workers. Airflow rates will be checked routinely when in use and 

after modification of any hood-exhausted enclosure, close-capture point equipment, or 

ventilation system serving these barriers. 

The applicant discussed the Respiratory Protection Program commitments in Section 4.6.2 of 

the LA (IIFP, 2012a). The applicant will conduct the Respiratory Protection Program in 

accordance with Subpart H of 10 CFR Part 20 and consistent with the guidance in Regulatory 

Guide 8.15, Revision 1, “Acceptable Programs for Respiratory Protection,” issued October 1999

(NRC, 1999b). 

Approved written procedures for using Respiratory Protective Equipment will be implemented 

consistent with 10 CFR 20.1703(c)(4) and will dictate the following: 

• monitoring, including air sampling and bioassays 

• supervision and training of respiratory protection users 

• fit testing of respirators 

• respirator selection 

• breathing air quality 

• inventory and control of respirators 

• cleaning of respirators 

• storage, issuance, maintenance, repair, testing, and quality assurance of respiratory 

protection equipment 
4-11 

• recordkeeping 

• limitations on respirator use and relief from respirator use. 

The applicant will maintain records of the Respiratory Protection Program (including training for 

respiratory use and maintenance) in accordance with the Records Management Program, as 

described in Chapter 11 of the LA (IIFP, 2012a). 

Based on the analysis as summarized above, the NRC staff finds that the commitments in the 

IIFP’s LA (IIFP, 2012a) satisfactorily address the application acceptance criteria in 

Section 4.4.6.3 of NUREG-1520, Revision 0 (NRC, 2002). 

4.3.7 Radiation Survey and Monitoring Programs 

The NRC staff reviewed the applicant’s Radiation Survey and Monitoring Program commitments 

against the acceptance criteria in NUREG-1520, Revision 0, Section 4.4.7.3 (NRC, 2002). The 

following discussion summarizes the NRC staff’s analysis as to whether the information 

provided by the applicant in the LA (IIFP, 2012a) meets the criteria. 

The applicant discussed the Radiation Survey and Monitoring Programs in Section 4.7 of the LA 

(IIFP, 2012a). IIFP stated that it will conduct necessary and reasonable surveys, in accordance 

with 10 CFR 20.1501(a) and 10 CFR 20.1501(b), that satisfy the applicable regulations and are 

adequate to identify the extent of radiation levels, concentrations, or quantities of radioactive 

material and the potential radiological hazards. Instruments and equipment will be calibrated 

periodically in accordance with 10 CFR 20.1501(b). Personnel dosimeters will be processed by 

a vendor accredited by the National Voluntary Laboratory Accreditation Program (NVLAP), in 

accordance with 10 CFR 20.1501(c). IIFP will also monitor exposure to radiation and 

radioactive material to demonstrate compliance with occupational dose limits, in accordance 

with 10 CFR 20.1502, “Conditions Requiring Individual Monitoring of External and Internal 

Occupational Dose.” Written procedures will address survey and monitoring objectives, 

sampling procedures and data analysis methods, types of equipment and instrumentation to be 

used, frequency of measurements, recordkeeping and reporting requirements, and actions to be 

taken in case measurements exceed administrative or regulatory limits. 

The applicant will measure external occupational dose in accordance with 10 CFR 20.1501(a). 

Individually assigned dosimeters will be used to determine deep dose equivalent and shallow 

dose equivalent from external sources of radiation. In accordance with written procedures, 

personnel dosimeters will be distributed to individuals based on their job functions, 

commensurate with the amount of time an individual spends working with or near radioactive 

materials. An NVLAP-accredited vendor will process personnel dosimeters. The capability will 

exist to process dosimeters expeditiously if there is an indication of an exposure in excess of 

established action guides. Written procedures will establish action guides for external 

exposures. 

IIFP will design and implement its Personnel Monitoring Program for internal occupational 

radiation exposures based on the requirements of 10 CFR 20.1201, 10 CFR 20.1204, 

“Determination of Internal Exposure,” 10 CFR 20.1502(b), and 10 CFR 20.1704(i). Intakes will 

be assigned to individuals based on one or more types of measurements as follows: air 

sampling, in vitro bioassay (i.e., urinalysis or fecal); and in vivo bioassay (i.e., lung counting). 

The type and frequency of measurements for an individual will be determined by his or her job 
4-12 

function and properties of the licensed material associated with a known or suspected intake. 

The measurements will be commensurate with the amount of time an individual spends working 

with or near radioactive material. Intakes will be converted to committed dose equivalent and 

committed effective dose equivalent for the purposes of limiting and recording occupational 

doses. 

Approved written procedures will establish action levels to prevent an individual from exceeding 

the occupational exposure limits specified in 10 CFR 20.1201. Work activity restrictions will be 

imposed when an individual’s exposure exceeds 80 percent of the 10 CFR 20.1201 limit. 

Control actions will include temporarily restricting the individual from working in an area 

containing airborne radioactivity, and actions are taken as necessary to prevent recurrence. 

Exposure to airborne radioactive material may also be controlled through limiting access to 

areas, limiting exposure time, and use of respiratory equipment. 

Conduct of the in vitro bioassay program will primarily evaluate the intake of soluble uranium to 

ensure that the 10 CFR 20.1201(e) intake limit of 10 milligrams per week is not exceeded. 

Personnel assigned to work in areas where soluble airborne uranium compounds will be present 

in concentrations likely to result in intakes exceeding 10 percent of the applicable limits in 

10 CFR 20.1201 will be monitored by urinalysis or fecal bioassay methods, or both. Written 

procedures will specify the minimum sampling frequency for these individuals. In vitro

monitoring may will also be used for individuals involved in non-routine operations, 

perturbations, or incidents. The applicant will use an offsite laboratory that meets the 

performance standards specified in American National Standards Institute/Health Physics 

Society (ANSI/HPS) N13.22-1995, “Bioassay Program for Uranium” (ANSI/HPS, 1995), and 

ANSI/HPS N13.30-1996, “Performance Criteria for Radiobioassay” (ANSI/HPS, 1996), to 

process and analyze in vitro bioassay samples. 

The applicant will conduct in vivo lung counting as necessary to supplement or verify in vitro 

bioassay results. Qualified contractors will perform in vivo lung counting in accordance with 

ANSI/HPS N13.35-2009, “Specifications for the Bottle Manikin Absorption Phantom” 

(ANSI/HPS, 2009) performance standards. 

In accordance with procedure, the summation of external and internal occupational radiation 

exposure will be reported as a TEDE and will be calculated consistent with the guidance in 

Regulatory Guide 8.34, “Monitoring Criteria and Methods to Calculate Occupational Radiation 

Doses,” issued July 1992 (NRC, 1992a). 

The applicant will design and implement an Air Sampling Program in areas of the IIFP facility 

that are identified as potential airborne radioactivity areas. This program will include procedures 

to conduct air surveys and to calibrate and maintain RP airborne sampling equipment in 

accordance with the manufacturers’ recommendations. Evaluations of air sampling 

effectiveness will be performed in accordance with the methods and acceptance criteria in 

Regulatory Guide 8.25, Revision 1, “Air Sampling in the Workplace,” issued June 1992

(NRC, 1992c). 

Air samples will be continuously taken from each main process area where airborne 

concentrations could potentially exceed 0.3 DAC when averaged over 40 hours to assess the 

concentrations of uranium in the air. In accordance with procedures, the air samples will be 

collected in such a way that the concentrations of uranium measured are representative of the 

air which workers breathe. Filters from air samplers will be changed each shift during normal 

operating periods or at more frequent intervals following the detection of an event that may have 
4-13 

released airborne uranium, based upon knowledge of the particular circumstances. Filters will 

be changed less frequently during periods when no work is in progress. Grab samples will be 

obtained during maintenance activities that are known to have the potential to generate airborne 

radioactivity levels in excess of 1.0 DAC. 

Each air sampler will be equipped with a flow meter to indicate flow rate of air sampled. These 

flow meters will be calibrated or replaced every 18 months, at a minimum. Air sampling results 

in excess of 2.5 DAC (8-hour sample), and not resulting from specific known causes, will be 

investigated to determine the probable cause. Operations or equipment will be shut down, and 

immediate corrective action will be taken at locations where an air sample exceeds 10 DAC 

without a specific known cause. 

The IIFP facility will be designed and operated in accordance with 10 CFR 20.1406, 

“Minimization of Contamination,” to minimize contamination, facilitate eventual 

decommissioning, and minimize where practicable the generation of radioactive waste. The 

following are examples of methods that will be employed for minimizing contamination: 

• containment of radioactive material throughout the facility 

• monitoring for equipment leaks 

• training on proper techniques for handling radioactive material 

• airflow from areas of low radioactivity to higher radioactivity. 

Routine surveys will be performed in areas that are most likely to be contaminated or where 

contamination from licensed processes, radioactive decay products or other radionuclide 

contaminants may concentrate. RP technicians will perform routine contamination surveys in 

the change rooms, plant exit walkways, and the laboratory area. The RP staff will determine 

survey frequencies, compare the survey results to action guide values specified in procedures, 

and ensure that the appropriate responses are taken. If the results exceed the action guide 

values, the RPM (or designee) will be informed and will determine whether an investigation or 

corrective actions are necessary. 

Personnel contamination surveys will be required for external contamination on clothing and the 

body by personnel exiting the change rooms. If contamination is found in excess of background 

levels, the individual will attempt self-decontamination (except for facial contamination) at the 

facilities provided in the change rooms. If decontamination attempts are not successful, or if 

facial contamination is detected, decontamination assistance will be provided by the RP function 

(typically an RP technician). If skin or personal clothing is still contaminated above background 

levels, the individual will not be permitted to leave the area without the prior approval (in 

accordance with procedure) of the RPM. 

The applicant will implement and document corrective actions for airborne occupational 

exposure based on the frequency and magnitude of events causing releases of airborne 

uranium that exceed administrative limits. Portable air sample surveys will supplement routine 

air sampling as required to evaluate nonroutine activities or breaches in containment. RP and 

operations staff will investigate the cause of the release and implement recommended actions 

to prevent future releases. 
4-14 

Appropriate radiation detection instruments will be available in sufficient number to ensure that 

adequate radiation surveillance can be accomplished. Selection criteria for portable and 

laboratory counting equipment will be based on the types of radiation detected, maintenance 

requirements, ruggedness, interchangeability, and upper and lower limits of detection 

capabilities. 

Portable instrumentation will be calibrated in accordance with manufacturing recommendations 

before initial use, after major maintenance, and on a routine basis specified by the 

manufacturer. Calibration will consist of a performance check on each range scale of the 

instrument with a radioactive source of known activity traceable to a recognized standard, such 

as the National Institute of Standards and Technology. Before each use, operability checks will 

be performed on monitoring and laboratory counting instruments. The background and 

efficiency of laboratory counting instruments will be determined on a daily basis when in use. 

The applicant discussed clearance of materials from RCAs in Section 4.7.13 of the LA 

(IIFP, 2012a). When removing equipment and materials from RCAs, with the exception of 

hazardous chemicals produced from licensed operations, IIFP staff will follow the guidance in 

NRC’s branch technical position, “Guidelines for Decontamination of Facilities and Equipment 

Prior to Release for Unrestricted Use or Termination of Licenses for Byproduct, Source, or 

Special Nuclear Material” (NRC, 1993b). Volumetrically contaminated materials will be released 

if the uranium concentration of the material does not exceed 30 picocuries per gram or the dose 

to a member of the public, taking into consideration the subsequent use of the material, does 

not exceed 1 mrem per year. The RP staff must approve the release of equipment or materials 

from RCAs. Written procedures will govern the equipment and material screening and 

evaluation process. 

Hazardous chemicals produced from licensed materials, as defined in 10 CFR 70.4, 

“Definitions,” will be considered “separated from licensed materials” by meeting the 

concentration levels described in 10 CFR 40.13(a) for “unimportant quantities of source 

material.” The term “unimportant quantities of source material” is defined in 10 CFR 40.13(a) as 

“...source material in any chemical mixture, compound, solution, or alloy in which the source 

material is by weight less than one-twentieth of 1 percent (0.05 percent) of the mixture, 

compound, solution, or alloy.” Written procedures will govern the analytical methods applied to 

determine the concentration of source material in hazardous chemicals. IIFP will conduct 

statistical sampling to provide confidence that the quantity of source materials remains below 

the “unimportant quantities” threshold. 

The applicant discussed leak testing of sealed sources in Section 4.7.14 of the LA (IIFP, 2012a) 

and made commitments consistent with branch technical positions that are acceptable to staff 

(“License Condition for Leak-Testing Sealed Byproduct Material Sources” (NRC, 1993c); 

“License Condition for Leak-Testing Sealed Plutonium Sources” (NRC, 1993d); “License 

Condition for Plutonium Alpha Sources” (NRC, 1993e); “License Condition for Leak-Testing 

Sealed Source Which Contains Alpha and/or Beta-Gamma Emitters” (NRC, 1993f); and 

“License Condition for Leak-Testing Sealed Uranium Sources” (NRC, 1993g), all of which were 

issued April 1993). When not in use, sources will be stored in a closed container adequately 

designed and constructed to contain radioactive material that may otherwise be released during 

storage. The sources will be tested for leakage using the dry wipe test method in accordance 

with International Standardization Organization (ISO) 9978, “Radiation Protection—Sealed 

Radioactive Sources—Leakage Test Methods” (ISO, 1992). Sealed sources will be leak 

checked at 6-month intervals, not to exceed that specified on the sealed source and device 

registration certificate. The check will use a quantitative analysis capable of detecting 
4-15 

185 becquerels (Bq) (0.005 microcuries (μCi)) of radioactivity. Leak tests will not be required if 

any one of the following conditions is true: 

• sources contain only tritium 

• sources contain only licensed material with a half-life of less than 30 days 

• sources contain only a radioactive gas 

• sources contain 3.7 megabecquerels (100 μCi) or less of beta-emitting or 

gamma-emitting material or 370 kilobecquerels (10 μCi) or less of alpha emitting 

material 

• sources are stored and are not being used (must be leak tested before use or transfer). 

Sources that exhibit removable contamination in excess of 185 Bq (0.005 μCi) will be removed 

from service and disposed of in accordance with regulations. 

Section 4.7.15 of the LA (IIFP, 2012a) discusses personnel access control to RCAs. For most 

RCAs, routine access points will be established through change rooms. Each change room will 

include a step-off area provided between the contamination controlled and noncontrolled areas. 

Instructions controlling entry and exit from RCAs will be posted at the entry points. Survey 

meters will be provided in the step-off area of each change room for use by personnel leaving 

the RCA. Posted instructions will address the use of the survey meters, donning and doffing of 

protective clothing, and appropriate decontamination methods. Alternate access points to RCAs 

will be established for specific activities not accommodated by the change rooms. Such access 

will be governed by approved written procedures or RWPs, which will establish controls to 

prevent the spread of contamination to noncontrolled areas. 

RCAs that may pose a risk to employees will be identified and posted in compliance with the 

requirements in 10 CFR 20.1901, “Caution Signs”; 10 CFR 20.1902, “Posting Requirements”; 

and 10 CFR 20.1903, “Exceptions to Posting Requirements.” Access to these areas will be 

controlled so that only appropriately trained individuals are allowed entry. Signs will be regularly 

inspected for conformance. The following areas will be identified and posted if applicable in 

accordance with definitions provided in 10 CFR 20.1003, “Definitions”: 

• radiation area 

• high radiation area (unlikely to have but a sealed calibration source may require) 

• airborne radioactivity area 

• radioactive material area 

In addition, contamination areas will be posted in accordance with approved written procedures. 

Signs will be posted at the entry points of areas requiring protective clothing. Radiation safety 

training and approved written procedures will instruct employees on requirements for entering 

and working in posted areas. 
4-16 

The applicant discussed the Radiation Reporting Program in Section 4.7.16 of the LA 

(IIFP, 2012a). IIFP will establish a Radiation Reporting Program to maintain records of the RP 

Program, radiation survey results, Corrective Action Program referrals, RWPs, and planned 

special exposures. The Radiation Reporting Program will be consistent with the guidance in 

Regulatory Guide 8.7, Revision 2, “Instructions for Recording and Reporting Occupational 

Radiation Exposure Data,” issued November 2005 (NRC, 2005). 

The Radiation Reporting Program will report to the NRC any event resulting in an occupational 

exposure to radiation exceeding the dose limits in 10 CFR 20.1201, within the time specified in 

10 CFR 20.2202, “Notification of Incidents”; 10 CFR 40.60, “Reporting Requirements”; and 

10 CFR 70.74, “Additional Reporting Requirements.” The Radiation Reporting Program is also 

committed to preparing and submitting to the NRC an annual report of individual monitoring 

results, as required by 10 CFR 20.2206(b). 

Radiation exposure data for an individual, and the results of any measurements, analyses, and 

calculations of radioactive material deposited or retained in the body of an individual, will be 

reported to the individual as specified in 10 CFR 19.13, “Notifications and Reports to 

Individuals.” During basic radiation safety training, individuals will be advised of their right to 

request radiation exposure data. In accordance with 10 CFR 19.11, “Posting of Notices to 

Workers,” IIFP management will post current copies at locations where they may be reviewed of 

the following documents: 

• the regulations in 10 CFR Part 19 and 10 CFR Part 20 

• the license, or documents incorporated into the license by reference and amendments 

thereto 

• the operating procedures applicable to licensing activities 

Based on the review as summarized above, the NRC staff finds that the commitments in the 

IIFP’s LA (IIFP, 2012a) satisfactorily address the acceptance criteria in Section 4.4.7.3 of 

NUREG-1520, Revision 0 (NRC, 2002). 

4.3.8 Additional Program Requirements 

The NRC staff reviewed the applicant’s additional program commitments against the 

acceptance criteria in NUREG-1520, Revision 0, Section 4.4.8.3 (NRC, 2002). The following 

discussion summarizes the NRC staff’s analysis as to whether the information provided by the 

applicant in the LA (IIFP, 2012a) meets the criteria. 

The applicant discussed additional program commitments in Section 4.8 of the LA (IIFP, 2012a). 

IIFP will maintain records of the RP Program (including program provisions, audits, and reviews 

of the program context and implementation); radiation survey results (air sampling, bioassays, 

external exposure data from monitoring individuals, and internal intakes of radioactive material); 

and results of Corrective Action Program referrals, RWPs, and planned special exposures. 

Records will be maintained consistent with the commitments in Section 11 of the LA 

(IIFP, 2012a). 

Procedures will dictate that IIFP will report to the NRC, within the time specified by 

10 CFR Part 20, Subpart M, “Reports,” and 10 CFR 70.74, any event resulting in an 

occupational exposure to radiation exceeding the dose limits in 10 CFR Part 20. 
4-17 

IIFP will prepare and submit an annual report to the NRC of the results of individual monitoring, 

as required by 10 CFR 20.2206(b). 

The IIFP Corrective Action Program will evaluate any radiation incident resulting in an 

occupational exposure that exceeds the dose limits in 10 CFR 20.1201, or is required to be 

reported in accordance with Subpart M of 10 CFR Part 20, 10 CFR 40.60, and 10 CFR 70.74. 

The applicant will report to the NRC the corrective actions taken (or planned) to protect against 

a recurrence and the proposed schedule to achieve compliance. 

Based on the review as summarized above, the NRC staff finds that the commitments in the LA 

(IIFP, 2012a) satisfactorily address the acceptance criteria in Section 4.4.9.3 of NUREG-1520, 

Revision 0 (NRC, 2002). 

4.3.9 Control of Radiological Risk Resulting from Accidents 

The applicant submitted the ISA Summary (IIFP, 2012b), as well as the accident consequence 

evaluations (IIFP, 2010) supporting the ISA Summary. These documents sufficiently described 

the radiological hazards, accident sequences, consequences, and methodology for staff to 

initially evaluate compliance with the radiological provisions in 10 CFR Part 70, Subpart H. 

Descriptions and justifications of the initiating events, preventions, and mitigating factors were 

also presented to an acceptable extent. In addition, the applicant submitted an EP (IIFP, 2011) 

to address response to accidents at the facility. This plan generally conforms to guidance in 

Regulatory Guide 3.67, “Standard Format and Content for Emergency Plans for Fuel Cycle and 

Materials Facilities,” issued January 1992 (NRC, 1992b). 

As discussed in Appendix B to this SER, the NRC staff independently evaluated a sampling of 

the applicant’s accident sequences to verify the methodologies employed and consequence 

categorizations. The NRC staff’s results were consistent with consequence categories 

determined by the applicant. Overall, the applicant concluded in the accident consequence 

evaluations (IIFP, 2010) that radiological dose, soluble uranium exposures, and environmental 

releases could be categorized as “low” for all credible accident sequences evaluated consistent 

with categorizations in 10 CFR 70.61. The chemicals associated with the processes present the 

primary hazard expected at the facility, which the consequence analysis determined as having 

the greater impact relative to the performance criteria in 10 CFR 70.61. Because the applicant 

determined that radiological consequences from credible process accidents are “low,” there 

were no IROFS identified to reduce or eliminate radiological exposures and no management 

measures identified to maintain the IROFS. The NRC staff also noted that the applicant 

identified IROFS and associated management measures to prevent or mitigate chemical 

exposures resulting from process accidents, as addressed in Chapter 6 of this SER. 

Based on the NRC staff’s evaluation of the analysis as summarized above, the NRC staff finds 

that the radiological evaluations the applicant performed to support the ISA, as well as the EP 

developed for the facility, demonstrate compliance with 10 CFR Part 70, Subpart H. 

4.4 Evaluation Findings

The applicant has committed to establish and maintain an acceptable RP Program that includes 

the following: 

• an effective, documented program to ensure that occupational radiological exposures 
4-18 

are ALARA 

• an organization with adequate qualification requirements for the RP personnel 

• approved, written RP procedures and RWPs for RP activities 

• RP training for all personnel who have access to restricted areas 

• a program to control airborne concentrations of radioactive material with engineering 

controls and respiratory protection 

• a radiation survey and monitoring program that includes requirements for controlling 

radiological contamination within the facility and monitoring external and internal 

radiation exposures 

• other programs implement corrections at the facility, maintain records, and report to the 

NRC in accordance with 10 CFR Part 20, 10 CFR Part 40, and 10 CFR Part 70. 

In addition, the applicant evaluated, in the ISA Summary (IIFP, 2012b), accident sequences with 

potential intermediate and high radiological consequences. Because none of the accident 

sequences resulted in intermediate or high radiological consequences, the radiological portion 

of the ISA Summary meets the performance criteria of 10 CFR 70.61. 

The NRC staff concludes that the applicant’s RP Program is adequate and meets the 

requirements of 10 CFR Part 19, 10 CFR Part 20, 10 CFR Part 40, and Subpart H of 

10 CFR Part 70. Conformance to the analysis and commitments to the LA will ensure safe 

operation. 

4.5 References

(ANSI/HPS, 2009) American National Standards Institute/Health Physics Society, 

“Specifications for the Bottle Manikin Absorption Phantom,” ANSI/HPS N13.35-2009, 2009. 

(ANSI/HPS, 1996) American National Standards Institute/Health Physics Society, “Performance 

Criteria for Radiobioassay,” ANSI/HPS N13.30-1996, 1996. 

(ANSI/HPS, 1995) American National Standards Institute/Health Physics Society, “Bioassay 

Program for Uranium,” ANSI/HPS N13.22-1995, 1995. 

(ICRP, 1994) International Commission on Radiological Protection, “Dose Coefficients for 

Intakes by Workers,” Publication 68, 1994. 

(IIFP, 2012a) International Isotopes Fluorine Products, Inc., “Fluorine Extraction Process and 

Depleted Uranium Deconversion Plant (FEP/DUP) License Application, Revision B,” May 2012, 

Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML12123A245. 

(IIFP, 2012b) International Isotopes Fluorine Products, Inc., “ISA Summary Rev. B for IIFP,” 

May 2012, Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML12123A245. 
4-19 

(IIFP, 2011) International Isotope Fluorine Products, Inc., “(NonPublic) Emergency Plan Rev B 

of IIFP License Application,” May 2012, Agencywide Documents Access and Management 

System (ADAMS) Accession No. ML12123A245. 

(IIFP, 2010) International Isotope Fluorine Products, Inc. (IIFP), “Letter from INIS, RE: 

Integrated Safety Analysis Supporting Documents Including: Consequence Evaluation 

Calculations and Accident Consequence Evaluations TAC L32739,” June 2010. Agencywide 

Documents Access and Management System (ADAMS) Accession No. ML102560265 

(ISO, 1992) International Organization for Standardization, “Radiation Protection—Sealed 

Radioactive Sources—Leakage Test Methods,” ISO 9978, 1992. 

(NRC, 2008) U.S. Nuclear Regulatory Commission, “Minimization of Contamination and 

Radioactive Waste Generation: Life-Cycle Planning,” Regulatory Guide 4.21, June 2008. 

(NRC, 2005) U.S. Nuclear Regulatory Commission, “Instructions for Recording and Reporting 

Occupational Radiation Exposure Data,” Regulatory Guide 8.7, Rev. 2, November 2005. 

(NRC, 2002) U.S. Nuclear Regulatory Commission, “Standard Review Plan for the Review of a 

License Application for a Fuel Cycle Facility,” NUREG-1520, Rev. 0, March 2002. 

(NRC, 1999a) U.S. Nuclear Regulatory Commission, “Instruction Concerning Prenatal 

Radiation Exposure,” Regulatory Guide 8.13, Rev. 3, June 1999. 

(NRC 1999b) U.S. Nuclear Regulatory Commission. “Acceptable Programs for Respiratory 

Protection,” Regulatory Guide 8.15, Rev. 1, October 1999. 

(NRC, 1996) U.S. Nuclear Regulatory Commission, “Instruction Concerning Risk from 

Occupational Radiation Exposure,” Regulatory Guide 8.29, Rev. 1, February 1996. 

(NRC, 1993a) U.S. Nuclear Regulatory Commission, “ALARA Levels for Effluents from 

Materials Facilities,” Regulatory Guide 8.37, July 1993. 

(NRC, 1993b) U.S. Nuclear Regulatory Commission, Branch Technical Position, “Guidelines for 

Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or 

Termination of Licenses for Byproduct, Source, or Special Nuclear Material,” 1993. 

(NRC, 1993c) U.S. Nuclear Regulatory Commission, Branch Technical Position, “License 

Condition for Leak-Testing Sealed Byproduct Material Sources,” April 1993. 

(NRC, 1993d) U.S. Nuclear Regulatory Commission, Branch Technical Position, “License 

Condition for Leak-Testing Sealed Plutonium Sources,” April 1993. 

(NRC, 1993e) U.S. Nuclear Regulatory Commission, Branch Technical Position, “License 

Condition for Plutonium Alpha Sources,” April 1993. 

(NRC, 1993f) U.S. Nuclear Regulatory Commission, Branch Technical Position, “License 

Condition for Leak-Testing Sealed Source Which Contains Alpha and/or Beta-Gamma 

Emitters,” April 1993. 
4-20 

(NRC, 1993g) U.S. Nuclear Regulatory Commission, Branch Technical Position, “License 

Condition for Leak-Testing Sealed Uranium Sources,” April 1993. 

(NRC, 1992a) U.S. Nuclear Regulatory Commission, “Monitoring Criteria and Methods to 

Calculate Occupational Radiation Doses,” Regulatory Guide 8.34, July 1992. 

(NRC, 1992b) U.S. Nuclear Regulatory Commission, “Standard Format and Content for 

Emergency Plans for Fuel Cycle and Materials Facilities,” Regulatory Guide 3.67, January 1992. 

(NRC, 1992c) U.S. Nuclear Regulatory Commission, “Air Sampling in the Workplace,” 

Regulatory Guide 8.25, Rev. 1, June 1992. 

(NRC, 1977) U.S. Nuclear Regulatory Commission, “Operating Philosophy for Maintaining 

Occupational Radiation Exposures as Low as Is Reasonably Achievable,” Regulatory 

Guide 8.10, Rev. 1, May 1977. 

(NRC, 1973) U.S. Nuclear Regulatory Commission, “Guide for Administrative Practice in 

Radiation Monitoring,” Regulatory Guide 8.2, February 1973. 
5-1 

5.0 NUCLEAR CRITICALITY SAFETY 

The IIFP facility is a source material facility licensed under Title 10 of the Code of Federal 

Regualtions (10 CFR) Part 40 and the Integrated Safety Analysis (ISA) requirements in 

10 CFR Part 70, Subpart H. As such the facility does not process or possess special nuclear 

material (SNM) other than check sources used for calibration licensed by the Agreement state. 

International Isotopes Fluorine Products, Inc. (IIFP), does not have a Nuclear Criticality Safety 

(NCS) program or procedures as described in Chapter 5, “Nuclear Criticality Safety” of the 

license application (IIFP, 2012a). 

5.1 Regulatory Requirements

Title 10 CFR Part 40 does not contain criticality requirements for source material due to the 

absence of enriched material. IIFP must demonstrate compliance with the 10 CFR Part 70, 

Subpart H, including requirements involving criticality. 

As part of the ISA, IIFP did identify one criticality accident scenario which involves the 

inadvertent receipt and processing of a shipment of fissile material. IIFP has identified controls 

which will prevent this type of accident. The cylinders typically processed at the facility consist 

of the 14-ton uranium hexafluoride tails cylinders, not the typical 21⁄2-ton enriched cylinders. In 

addition, an item relied on for safety (IROFS). 

5.2 Evaluation of Findings

Based on this review, the U.S. Nuclear Regulatory Commission (NRC) staff finds acceptable the 

applicant’s assessment that IIFP is not authorized to possess SNM under the NRC license and, 

therefore, is not required to implement an NCS program. Consistent with the ISA requirements 

in 10 CFR 70.61(d) and 10 CFR 70.64(a), the applicant has identified adequate controls and 

IROFS to prevent a criticality involving an inadvertent shipment of fissile material. 

Therefore, the NRC staff concludes that the applicant’s information regarding NCS meets the 

requirements of 10 CFR Part 70, Subpart H, and provides reasonable assurance for the 

protection of public health and safety, including workers and the environment. 

5.3 References

(IIFP, 2012a) International Isotopes Fluorine Products, Inc., “Fluorine Extraction Process and 

Depleted Uranium Deconversion Plant (FEP/DUP) License Application, Revision B,” May 2012, 

Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML12123A245. 

(IIFP, 2012b) International Isotopes Fluorine Products, Inc., “ISA Summary Rev. B for IIFP,” 

May 2012, Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML12123A245. 

 
5-2 

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6-1 

6.0 CHEMICAL PROCESS SAFETY 

The purpose of the U.S. Nuclear Regulatory Commission’s (NRC’s) review of the International 

Isotopes Fluorine Products, Inc. (IIFP or applicant), Chemical Safety Program and the facility’s 

design is to evaluate whether the applicant will adequately protect workers, the public, and the 

environment during normal operations against chemical hazards of licensed material and its 

byproducts. The Chemical Safety Program and the facility’s design must also protect against 

facility conditions and operator actions that can affect the safety of licensed materials and thus 

present an increased chemical and radiological risk. 

6.1 Regulatory Requirements

In Staff Requirements Memorandum SECY-07-0146, “Regulatory Options for Licensing New 

Uranium Conversion and Depleted Uranium Deconversion Facilities,” dated October 10, 

2007 (NRC, 2007a), the Commission stated, “If new license applications are submitted before 

the completion of the rulemaking [to Title 10 of the Code of Federal Regulations (10 CFR) Part 

40], the NRC staff shall impose 10 CFR Part 70, Subpart H, performance requirements as part 

of the licensing basis for the application review.” Therefore, since revisions to 10 CFR Part 40 

are not completed, the regulatory bases for the review must address the chemical process 

safety requirements in 10 CFR 70.65, “Additional Content of Applications.” In addition, the 

chemical process safety review should provide reasonable assurance of compliance with 

10 CFR 70.61, “Performance Requirements”; 10 CFR 70.62, “Safety Program and Integrated 

Safety Analysis”; and 10 CFR 70.64, “Requirements for New Facilities or New Processes at 

Existing Facilities.” 

6.2 Regulatory Guidance and Acceptance Criteria

Chapter 6 of NUREG-1520, “Standard Review Plan for the Review of a License Application for a 

Fuel Cycle Facility,” issued March 2002 (NRC, 2002), contains the guidance applicable to 

NRC’s review of chemical process safety for the proposed facility. This chapter is applicable in 

its entirety. The NRC staff also used the following as guidance documents for this review: 

NUREG-1601, “Chemical Process Safety at Fuel Cycle Facilities,” issued August 1997 

(NRC, 1997), and NUREG-1513, “Integrated Safety Analysis Guidance Document,” issued May 

2001 (NRC, 2001). Section 6.4.3 of NUREG-1520 (NRC, 2002) provides the acceptance 

criteria applicable to this review. 

6.3 Staff Review and Analysis

The NRC staff reviewed the license application (LA) (IIFP, 2012a) and the Integrated Safety 

Analysis (ISA) Summary (IIFP, 2012b) submitted by the applicant and considered the following 

eight areas as they relate to chemical process safety: 

(1) process description 

(2) chemical accident sequences 

(3) chemical accident consequences 

(4) chemical process items relied on for safety (IROFS) 

(5) chemical process management measures 

(6) emergency management 

(7) baseline design criteria (BDC) 

(8) defense-in-depth 
6-2 

The NRC staff reviewed the applicant’s responses to requests for additional information (RAI) 

(IIFP, 2011) and ISA documents during an in-office review (NRC, 2010), as necessary, to gain a 

better understanding of the process and safety requirements. The NRC staff evaluated the 

information to determine whether the facility’s design complied with the chemical protection 

BDC and defense-in-depth requirements as they relate to chemical process safety specified in 

10 CFR 70.64(a)(5) and 10 CFR 70.64(b), respectively. Section 6.3.6 of this Safety Evaluation 

Report (SER) discusses compliance with these regulations in more detail. The following 

sections summarize the NRC staff’s evaluation and general information about the proposed 

fluorine extraction process and depleted uranium deconversion plant (FEP/DUP). 

6.3.1 Process Description 

The applicant described the proposed FEP/DUP in Section 1.1.3 of the LA (IIFP, 2012a) and 

provided a more detailed description in Chapter 3 of the ISA Summary (IIFP, 2012b). The 

process will begin at the proposed depleted uranium hexafluoride (DUF6) to depleted uranium 

tetrafluoride (DUF4) deconversion plant. The proposed DUF6 to DUF4 deconversion plant, which 

is described in Section 6.3.1.1 of this SER, will produce DUF4 and anhydrous hydrogen fluoride 

(AHF). DUF4 will be transferred to one of two proposed fluorine extraction processes (FEPs), 

which will produce fluoride products and uranium oxides. AHF will be sold as a chemical 

commodity. The two proposed processes in which DUF4 will be transferred are the silicon 

tetrafluoride (SiF4) process and boron trifluoride (BF3) process. The products that will be 

obtained from the proposed FEP are fluoride gases (SiF4 and BF3), and uranium oxides 

(uranium dioxide [UO2] or triuranium octoxide [U3O8]). The applicant plans to sell the fluoride 

gases to several industries. The uranium oxides are more stable forms of uranium and are 

suitable for disposal at licensed waste facilities. Sections 6.3.1.2 and 6.3.1.3 of this SER 

describe the proposed SiF4 and BF3 processes, respectively. The proposed IIFP facility will also 

consist of support processes, which include process offgas treatment (i.e., potassium hydroxide 

[KOH] scrubbing system), environmental protection process, fluoride products storage and 

packaging, AHF staging containment, and fluoride products trailer loading. This section 

describes these support processes. 

The applicant provided engineering drawings as part of the LA submittal. The engineering 

drawings are process flow diagrams (PFDs) of each of the facility processes. The PFDs are 

publicly available and can be obtained through the NRC’s Agencywide Documents Access and 

Management System (ADAMS) using the Accession Number ML100120768. The descriptions 

provided below are based on the applicant’s description and these PFDs. 

6.3.1.1 DUF6 to DUF4 Deconversion Plant 

The DUF6 will arrive at the proposed IIFP facility in 14-ton, type 48-Y cylinders owned by the 

IIFP deconversion customer. The DUF6 in these cylinders will be in the solid state. These 

cylinders will be built to the standards of the American National Standards Institute (ANSI) 

(ANSI, 2001) and will be transported by truck trailers approved by the U.S. Department of 

Transportation (DOT). 

Upon receipt, IIFP staff will visually inspect full cylinders of DUF6 for damage and survey them 

for radiation and removable contamination. Documents that accompany the shipment and 

contain information regarding cylinder identification, weight, and uranium assay will be reviewed 

and verified for accuracy. Uranium assay will be qualitatively verified by performing a 

nondestructive gamma survey measurement. Once accepted for receipt, the facility cylinder 
6-3 

hauler vehicle will unload the cylinder and place it in the full DUF6 storage pad area until it is 

scheduled for feed to the deconversion process. 

Once the cylinder is scheduled to be fed into the deconversion process, the cylinder hauler 

vehicle will retrieve the cylinder from the full DUF6 storage pad area and place it onto a cylinder 

scale cart for movement into the autoclave room. The cylinder bridge crane will lift the cylinder 

from the scale cart and place it on the accountability scale where the cylinder weight will be 

measured, recorded, and compared against the applicable weight limits (i.e., 27,560 pounds [lb] 

of DUF6 for type 48-Y cylinders). If the cylinder weight is within safety limits, the cylinder bridge 

crane will lift it and place into one of two steam-heated containment type autoclaves. These 

autoclaves will be cylindrical pressure vessels designed to meet the requirements of 

Section VIII, Division 1, of the American Society of Mechanical Engineers (ASME) Boiler and 

Pressure Vessel Code (ASME, 2004). The autoclaves will have a maximum allowable working 

pressure of 200 pounds per square inch gauge (psig) at 250 degrees Fahrenheit (F). 

The cylinder will be connected to the feed manifold using a connection line (pigtail) compatible 

with uranium hexafluoride (UF6). The pigtail will be visually inspected and tested for leaks. 

Before heating the cylinder, IIFP staff will perform tests to ensure that it is safe to heat the 

cylinder. These tests will include opening the cylinder valve to ensure valve clarity and checking 

the cylinder pressure (cold pressure check) to ensure that the cylinder does not contain 

significant amounts of noncondensable or impurity gases. Once all requirements for feeding 

have been met, the autoclave hydraulic closing system will close the autoclave door. 

The DUF6 cylinder will be heated using saturated steam and, as a result, the gaseous DUF6 will 

flow through the heat-traced and insulated piping of the feed header to the DUF6 surge tank. 

When the quantity of DUF6 in a cylinder has been reduced to the extent that the desired flow of 

DUF6 cannot be maintained, the residual DUF6 will be removed from the cylinder using a 

process called heeling. Heeling will be accomplished by connecting the cylinder to a low 

pressure source, such as the purge and evacuation system. The residual DUF6 will be 

transferred to another 48-Y cylinder located inside a cylinder cold box. The cylinder cold box 

will be located in the autoclave building. After the heeling process is complete, the cylinder 

weight will be measured to verify that it is empty. Once verified, the cylinder hauler vehicle will 

transport the cylinder to the empty cylinder area where it is stored until it is returned to the 

deconversion customer. 

The DUF6 will be fed from the surge tank to a high-temperature reaction vessel where it will 

react with hydrogen (H2) to produce solid DUF4 and gaseous AHF. The reaction is exothermic 

and is shown in Equation 6-1 below. The gaseous H2 supply will be produced onsite by steam 

reforming of natural gas followed by purification using pressure swing adsorption. Steam 

reforming of natural gas is one of the most developed and commercially used technologies 

(Kutz, 2007) for hydrogen production. 

 UF6 (gas) + H2 (gas) Æ UF4 (solid) + 2HF(gas) Eq. 6-1 

The solid DUF4 powder will be continuously withdrawn from the bottom of the reaction vessel 

using a cooling screw mechanism and transferred to storage hoppers. The DUF4 in the storage 

hoppers will be transferred to the FEP plant for use as raw material feed in the production of 

SiF4 and BF3. A two-stage dust collector system will be provided to control and recycle DUF4

dusts that are internal to the solids-handling equipment and generated by air or gas flow 

associated with the handling equipment. The collected fines from the dust collector system will 

be discharged into the vacuum transfer system to the DUF4 feed hopper located in the proposed 
6-4 

FEP plant. The applicant will continuously monitor the vent stack flow to the environment for 

uranium emissions. 

The cooling screw mechanism will have an offgas plenum where the entrained DUF4 particles, 

unreacted DUF6, gaseous AHF, unreacted H2, and nitrogen (N2) (used as a buffer on the cooling 

screw bearing seals) will flow and pass through a filter system to remove entrained DUF4

particles. The filter system will comprise primary and secondary stages configured in series. 

The primary stage will be a cyclone separator with a section of sintered metal filter tubes. 

Cyclone separators have been used in the United States for about 100 years and are one of the 

most widely used of all industrial gas-cleaning devices (Benítez, 1993). The cyclone separator 

will remove the larger entrained particles, while finer particles will be screened by the sintered 

metal filter tubes and fall down to the cyclone separator. The collected solids will be discharged 

to a fines hopper for transfer to the DUF4 powder stream. The secondary stage is a backup to 

the first-stage filter. It will capture the particles that may pass the first-stage filter. The solids 

from the secondary stage will be discharged into the vacuum transfer system to the DUF4 feed 

hopper located in the proposed FEP plant. 

Following the filter system, the offgas will flow through one of the two parallel banks of 

carbon-bed traps. The carbon-bed trap material will be activated carbon. The two parallel 

banks will provide operational flexibility, and each bank will have three carbon-bed traps 

configured in series. The carbon-bed traps will remove any trace quantities of unreacted DUF6

left in the offgas . Once a carbon-bed trap is determined to be full, the flow to the carbon-bed 

trap will be isolated; and the carbon will be removed and transferred to the decontamination 

building for removal, replacement, and eventual approved disposal of uranium-contaminated 

activated carbon. After exiting the carbon-bed traps, the offgas will enter a series of two heat 

exchangers. The first heat exchanger will be a partial hydrogen fluoride (HF) condenser. The 

offgas from the partial HF condenser will flow into the second heat exchanger, the total HF 

condenser. The condensed HF from both condensers will be drained to the HF storage tanks. 

The AHF staging containment building will house the HF storage tanks; this building will be 

separated from licensed materials. Section 6.3.1.7 of this SER describes the AHF staging 

containment building. The HF will be temporarily stored and then loaded into truck trailers 

inside the AHF staging containment building for shipment to customers. DOT will approve the 

trailers for shipment. 

Residual offgases (mainly unreacted H2, N2, and small amounts of HF) will leave the total HF 

condenser to a hydrogen burner system to combust any unreacted H2. The small amounts of 

HF in the offgas leaving the hydrogen burner system will flow to the KOH scrubbing system. 

The KOH scrubbing system is designed to remove the small amounts of HF before venting the 

offgases to the atmosphere. Section 6.3.1.4 of this SER describes this system. 

The NRC staff finds that the ISA Summary (IIFP, 2012b) adequately describes the DUF6 to 

DUF4 deconversion plant because it provides information that allows the NRC staff to 

understand the process and the chemical hazards that could result from potential accident 

sequences. 

6.3.1.2 SiF4 Process 

The proposed SiF4 process will contain two equal production trains. For simplification, this 

section will only describe one of these production trains. 
6-5 

DUF4 will be transferred from the proposed DUF6 to DUF4 deconversion plant to a hopper in the 

proposed SiF4 process. Silicon dioxide (SiO2) will be stored in another hopper. The contents 

from the DUF4 and SiO2 hoppers will be mixed and then fed to a rotary calciner. In the rotary 

calciner, DUF4 and SiO2 will react to form gaseous SiF4 and solid UO2. Equation 6-2, below, 

shows this reaction. Small amounts of moisture (i.e., water vapor) may be contained in the 

powder mixture and, because of the high temperatures in the rotary calciner DUF4 and water, 

will react to form gaseous HF and solid UO2. Equation 6-3 shows this reaction. 

 UF4 (solid) + SiO2 (solid) Æ SiF4 (gas) + UO2 (solid) Eq. 6-2 

 UF4 (solid) + 2H2O(vapor) Æ 4HF(gas) + UO2 (solid) Eq. 6-3 

The resulting SiF4 gas and trace impurities (mainly HF) will exit the rotary calciner as an offgas , 

while the UO2 powder will be discharged at the end of the rotary calciner through a cooling 

screw mechanism and transferred to storage hoppers. A gas plenum that provides a redundant 

flow path for the offgases will be located at the top of the cooling screw mechanism. A 

two-stage dust collector system will control and recycle uranium oxide dusts that are internal to 

the solids-handling equipment and generated by air or gas flows associated with the handling 

equipment. The proposed SiF4 and BF3 processes will share the two-stage dust collector 

system, and the collected solids will be discharged to an uranium oxide hopper. The vent stack 

flow to the environment will be continuously monitored for uranium emissions. The uranium 

oxide in the storage hoppers will be packaged into DOT-approved shipping containers and 

transported to an offsite licensed disposal facility. 

Offgas leaving the rotary calciner and the cooling screw mechanism gas plenum will flow 

through two-stages of filters to capture entrained particles. The two-stage filter system will be 

similar to the filter system in the proposed DUF6 to DUF4 deconversion plant, described in 

Section 6.3.1.1 of this SER. The collected solids from both filters will be returned to the fines 

hopper to become part of the powder mix fed to the rotary calciner. After exiting the filter 

system, the offgas will flow to one of two precondensers to remove HF and other trace 

impurities from the SiF4 product stream. The collected HF and other trace impurities will be 

batch transferred to the KOH scrubbing system. Two precondensers will be provided for 

operational flexibility (i.e., one precondenser can transfer its contents to the KOH scrubbing 

system while the other is operating normally). 

The SiF4 product offgas will leave the precondenser and flow into a series of two product cold 

traps. The series of product cold traps will be configured into two parallel banks (i.e., two 

primary product cold traps and two secondary product cold traps). Again, this arrangement will 

be provided for operational flexibility. The primary product cold trap will solidify SiF4, and the 

secondary product cold trap will solidify any SiF4 remaining in the product offgas after passing 

the primary product cold trap. The offgas from the primary and secondary product cold traps 

will be sent to the KOH scrubbing system. After a product cold trap signals that it is full of solid 

SiF4 product, the cold trap will be set to an unloading mode by increasing the coolant 

temperature, thus causing the solid SiF4 to turn into liquid or gaseous SiF4. The liquid or 

gaseous SiF4 will flow to an evaporator vessel using a transfer compressor. The applicant 

stated that the evaporator vessel and transfer compressor will be located in the proposed FEP 

product storage and packaging area. The proposed FEP product storage and packaging area 

will be physically separated from the SiF4 process and licensed material. Section 6.3.1.6 of this 

SER describes the FEP product storage and packaging area. 
6-6 

The NRC staff finds that the ISA Summary (IIFP, 2012b) adequately describes the SiF4 process 

because it provides information that allows the NRC staff to understand the process and the 

chemical hazards that could result from potential accident sequences. 

6.3.1.3 BF3 Process 

DUF4 will be transferred from the proposed DUF6 to DUF4 deconversion plant to a hopper in the 

BF3 process. Diboron trioxide (B2O3) will be stored in another hopper. The contents from the 

DUF4 and B2O3 hoppers will be mixed, fed to a preheater, and then fed to a rotary calciner. The 

temperature in the preheater will be controlled to cause a reaction of small amounts of DUF4

with moisture (shown in Equation 6-3 above) that may be contained in the powder mixture. This 

will minimize the amount of HF impurities in the product gas stream. The HF will leave the 

preheater, move through a series of filters, and then enter the KOH scrubbing system. The filter 

system will be similar to the two-stage filter system of the proposed DUF6 to DUF4 deconversion 

plant. The collected solids will be transferred to the fines hopper. The contents of the fines 

hopper will mix with the DUF4 and B2O3 powders before entering the preheater. In the rotary 

calciner, DUF4 and B2O3 will react to form gaseous BF3 and solid UO2. Equation 6-4 shows this 

reaction. 

 UF4 (solid) + B2O3 (solid) Æ BF3 (gas) + UO2 (solid) Eq. 6-4 

The resulting BF3 gas and trace impurities (mainly HF) will exit the rotary calciner as an offgas , 

while the UO2 powder will discharge at the end of the rotary calciner through a cooling screw 

mechanism and then transfer to storage hoppers. A gas plenum that will provide a redundant 

flow path for the offgases will be located at the top of the cooling screw mechanism. A 

two-stage dust collector system will control and recycle uranium oxide dusts internal to the 

solids handling equipment and generated by air or gas flows associated with the handling 

equipment. The SiF4 and BF3 processes will share this two-stage dust collector system, which 

is described in Section 6.3.1.2 of this SER. The uranium oxide in the storage hoppers will be 

packaged into DOT approved shipping containers and transported to an offsite licensed 

disposal facility. 

Offgas leaving the rotary calciner and the cooling screw mechanism gas plenum will flow 

through two-stages of filters to capture entrained particles. The two-stage filter system will be 

similar to the filter system in the proposed DUF6 to DUF4 deconversion plant. The collected 

solids from both filters will be returned to the fines hopper to become part of the powder mix fed 

to the preheater and then to the rotary calciner. After exiting the filter system, the offgas will 

flow to one of two precondensers to remove HF and other trace impurities from the BF3 product 

stream. The collected HF and other trace impurities will be batch transferred to the KOH 

scrubbing system. As in the SiF4 process, the two precondensers will be provided for 

operational flexibility (i.e., one precondenser can transfer its contents to the KOH scrubbing 

system while the other is operating normally). 

The BF3 product offgas will leave the precondenser to a series of two product cold traps. The 

series of product cold traps will be configured into two parallel banks (i.e., two primary product 

cold traps and two secondary product cold traps). Again, this arrangement will be provided for 

operational flexibility. The primary product cold trap will solidify BF3, and the secondary product 

cold trap will solidify any BF3 remaining in the product offgas after passing the primary product 

cold trap. The offgas from the primary and secondary product cold traps will be sent to the 

KOH scrubbing system. After a product cold trap signals that it is full of solid BF3 product, the 

cold trap will be set to an unloading mode by increasing the coolant temperature, thus causing 
6-7 

the solid BF3 to turn into liquid or gaseous BF3. The liquid or gaseous BF3 will flow to an 

evaporator vessel using a transfer compressor. The applicant stated that the evaporator vessel 

and transfer compressor will be located in the FEP product storage and packaging area. The 

FEP product storage and packaging area will be physically separated from the BF3 process and 

licensed material. Section 6.3.1.6 of this SER describes the FEP product storage and 

packaging area. 

The NRC staff finds that the ISA Summary (IIFP, 2012b) adequately describes the BF3 process 

because it provides information that allows the NRC staff to understand the process and the 

chemical hazards that could result from potential accident sequences 

6.3.1.4 KOH Scrubbing System 

The KOH scrubbing system will comprise two parallel lines, and each line will have three 

scrubber stages. One of the lines will process HF from the proposed DUF6 to DUF4

deconversion plant and from the SiF4 and BF3 precondensers. The other line will process the 

final offgas coming from the SiF4 and BF3 processes, containing uncollected SiF4 and BF3 and 

trace quantities of other fluorides. KOH will be used as the scrubber liquor in each of the three 

scrubber stages. The three scrubber stages will be (1) a primary wet venturi scrubber, (2) a 

secondary countercurrent flow gas liquid-packed tower, and (3) a bed of sized coke. 

The system equipment will consist of a KOH storage tank, KOH pump tank, regenerated KOH 

tank, two venturi scrubbers, two spare venturi scrubbers, two packed towers, and two coke 

boxes. Redundant pumps will be available for each scrubber, pump tank, and storage tank. To 

remove the fluoride components from the offgas stream, the following reactions will occur: 

 HF(anhydrous) + H2O Æ HF(aqueous) Eq. 6-5 

 HF + KOH Æ KF + H2O Eq. 6-6 

 3SiF4 + 4KOH Æ 2K2SiF6 + 2H2O + SiO2 Eq. 6-7 

 SiO2 + 2KOH Æ K2SiO3 + H2O Eq. 6-8 

 4BF3 + 3KOH Æ 3KBF4 + B(OH)3 Eq. 6-9 

The KOH scrubbing system will vent the treated offgases through a single stack, and it will be 

designed to remove fluoride-bearing components in the offgas stream at approximate 

efficiencies of greater than 80 percent, 95 percent, and 99 percent for the first, second, and third 

stages, respectively. The overall system removal efficiency will be designed at greater than 

99.9 percent. The system stack will be continuously sampled and routinely analyzed to 

measure for traces of fluorides or uranium in the vent gas. 

The KOH solution will be used and recycled within each of the scrubbers until the KOH 

concentration needs replenishment. The concentration of KOH in the scrubber equipment will 

be maintained at a safe margin to ensure that it effectively reacts with fluoride components in 

the offgas stream. When the KOH concentration needs replenishment, some of the spent 

scrubbing solution containing potassium fluoride (KF), water, and some excess KOH will be 

pumped from the scrubber recycle tanks to the environmental protection process (EPP). 

Section 6.3.1.5 of this SER describes the EPP. 
6-8 

6.3.1.5 Environmental Protection Process 

The EPP will primarily be a means of treating two types of solutions that result from the 

production processes—KF solutions (KOH regeneration process) and weak aqueous HF (HF 

neutralization process). Each of these materials will originate from scrubbing systems designed 

to prevent the emission of fluoride compounds into the air. The KF solution will be a byproduct 

of using KOH as a scrubbing medium (see Equation 6-6 above). In the KOH regeneration 

process of the EPP, the KF, water, and excess KOH spent solution from the plant’s KOH 

scrubbing system will be reacted with a lime slurry. Calcium fluoride (CaF2) and regenerated 

KOH solution will be produced. The regenerated KOH will be recycled and reused in the plant’s 

scrubbing process. The CaF2 will be filtered, dried, and packaged for shipment to an approved 

commercial waste burial site, to an HF producer, or to other potential users. 

The other stream treated in the EPP will be weak aqueous HF solution, water, or KOH solution 

that may contain a low concentration of fluorides. Also, small spills that potentially occur and 

require cleanup from spill control containment areas may contain weak fluoride concentrations. 

In this case, the fluoride-bearing liquids may have too much water to send to the KOH 

regeneration and recycle system. The HF neutralization process will use lime slurry to react 

with weak HF to produce CaF2 and water. 

HF Neutralization

The HF neutralization process will be designed to operate intermittently, as needed. A lime silo 

will be provided, including an installed dust collector. The silo will hold an inventory of hydrated 

lime. Lime will be fed to a mix tank where it will be mixed with harvested water. The slurry 

generated will be approximately 30 percent solids. Dilute HF solution will be transferred from 

the weak HF solution tank to an agitated acid reaction vessel. The lime slurry from the mix tank 

will be also transferred to the acid reaction vessel. The materials in the acid reaction vessel will 

require a retention time of about 1 hour or greater for reaction completion. With the reaction 

complete, materials from the acid reaction vessel will be transferred to a thickener tank for 

settling. After thickening, CaF2 and excess lime will be transferred by a slurry-type pump from 

the bottom of the thickener to a rotary drum vacuum filter. Solids will be discharged from the 

filter to a dryer capable of removing excess water. Liquors from the rotary vacuum filter will be 

recycled to the weak HF solution tank for recycling. After drying, CaF2 will be packaged suitable 

for sale or disposal in an appropriate offsite licensed Resource Conservation Recovery Act 

(RCRA) disposal facility. Equation 6-10 shows the primary chemical reactions: 

 2HF + Ca(OH)2 Æ CaF2 + 2H2O Eq. 6-10 

KOH Regeneration

Lime will be fed to an agitated mix tank where it will mix with harvested water. The slurry 

generated will contain approximately 30 percent solids. Spent KOH solution (KF solution 

containing a weak concentration of KOH) will be transferred from a spent KOH storage tank to 

an agitated reaction vessel. The lime slurry from the mix tank will be also transferred to the 

reaction vessel. The materials in the reaction vessel tank will be given a retention time of about 

1 hour or greater for reaction completion. With the reaction complete, materials from the 

reaction vessel will be transferred to a thickening tank for settling. Excess lime and CaF2 will be 

transferred by a slurry pump from the bottom of the thickener to a rotary drum vacuum filter. 

Solids will be discharged from the filter to a dryer capable of processing excess water. Liquors 

will be transferred to a clarifier where trace solids will be settled. Regenerated KOH will be 
6-9 

removed from the top of the clarifier and passed through a set of filters to the regenerated KOH 

storage tank. The regenerated KOH solution will be pumped to the plant’s KOH scrubbing 

system as needed for reuse by the scrubbers. Solids will be transferred via a slurry pump from 

the bottom of the clarifier to the rotary drum vacuum filter and subsequently transferred to the 

dryer. The dried material will be packaged and stored for sale or sent to an approved offsite 

licensed RCRA disposal facility. Equation 6-11 shows the primary chemical reaction: 

 2KF + Ca(OH)2 Æ CaF2 +2KOH Eq. 6-11 

6.3.1.6 Fluorine Extraction Process Storage and Packaging Area 

As described in Sections 6.3.1.2 and 6.3.1.3 of this SER, the FEP storage and packaging area 

will be physically separated from licensed material. Therefore, the applicant stated that this 

process is outside the scope of NRC’s licensing review. However, the applicant stated that it 

conducted a process hazard analysis in accordance with the Occupational Safety and Health 

Administration (OSHA) regulations in 29 CFR 1910.119, “Process Safety Management of Highly 

Hazardous Chemicals.” For completeness, this section will briefly describe the FEP storage 

packaging area. 

From the evaporator and transfer compressor, the SiF4 will flow into a cooler and then into the 

storage tubes. There will be 15 storage tubes for SiF4 storage and one dump tube for accepting 

SiF4 gas in the event a relief valve from one of the other storage tubes actuates. Five of the 

15 storage tubes will be spares to be used when necessary. For BF3, the storage process will 

be essentially the same as that for SiF4. The difference is that there will be 10 storage tubes 

instead of 15. Five of the 10 tubes will be spares. The storage tubes for both products will be 

ASME Code rated pressure vessels. 

The SiF4 or BF3 gas will be packaged into small container cylinders or large container tube 

trailers. The small container cylinders and large container tube trailers will arrive at the 

applicant’s proposed facility from IIFP’s customers. When packaging is scheduled, the cylinder 

or tube trailer will be connected to the respective SiF4 or BF3 packaging manifold. Each 

packaging manifold will be enclosed in a containment-type packaging station that will be 

connected via a hood and ducts to the emergency KOH scrubbing system. Detectors will be 

located inside and outside the packaging station that will alarm and automatically shut off the 

flow of product gas at the storage tube manifold from which the package is being filled, and it 

will shut off the flow at the packaging manifold to the cylinder or tube trailer being filled. If a leak 

is detected, the packaging station will open valves on the station to vent to the emergency KOH 

scrubbing system. 

The small container cylinders and large container tube trailers will be ASME Code rated 

pressure vessels and will require DOT’s approval for shipment. 

6.3.1.7 AHF Staging Containment and Fluoride Products Trailer Loading Buildings 

Similar to the FEP storage and packaging area, the AHF staging containment and fluoride 

products trailer loading buildings will be separated from licensed material. Safeguards and 

operational controls will be provided to meet the requirements of OSHA regulations 

(29 CFR 1910.119) or Federal and State of New Mexico environmental permit requirements. 

Again, for completeness, this section will briefly describe the processes in these buildings. 
6-10 

As described in Section 6.3.1.1 of this SER, AHF from the partial and total HF condensers will 

be transferred to the AHF storage tanks located in the AHF staging containment building. Dikes 

will be provided to each AHF storage tank. Each dike will be sized to hold the contents of the 

entire tank and minimize the surface area of evaporation of AHF in the case of a leak from an 

AHF storage tank. The AHF will be temporarily stored, and when the inventory in a storage tank 

reaches a level for shipment, the AHF will be loaded into a customer-owned tank trailer staged 

in the fluoride products trailer product building. The tank trailers will be DOT approved for AHF 

shipment to customers. 

The fluoride products trailer loading building will serve two purposes. As mentioned above, it 

will serve as a loading place for AHF tank trailers from storage. The second purpose will be to 

load gas-tube trailers with SiF4 or BF3 transferred from the FEP product storage and packaging 

building (briefly described in Section 6.3.1.6 of this SER). 

The AHF staging containment and fluoride products trailer loading buildings will be connected 

and will have a fluoride detection system and water-spray deluge system. If a leak of AHF or 

fluoride product gas occurs, the fluoride detection system will automatically initiate an alarm and 

isolate the transfer of products at the storage tanks and tank trailer fill lines. The system will 

also automatically initiate the water-spray deluge system to knock down the fluoride vapors 

within the buildings. The water from the water-spray deluge system is gravity drained to a 

holding tank that vents to the KOH scrubbing system. If the holding tank contains aqueous HF, 

it would be sent to the EPP (described in Section 6.3.1.5 of this SER). 

6.3.1.8 Hazardous Chemicals and Chemical Interactions 

The hazardous chemicals present in significant quantities in the proposed IIFP facility will be 

UF6, uranium tetrafluoride (UF4), UO2, HF, SiF4, and BF3. Any UF6 that is released to the 

environment will react exothermically with moisture in the air producing solid uranyl fluoride 

(UO2F2) and HF gas. Equation 6-12 shows the reaction of gaseous UF6 with water vapor at 

elevated temperatures: 

 UF6 (gas) + 2H2O(vapor) Æ UO2F2 (solid) + 4HF(gas) + heat Eq. 6-12 

At room temperature, depending on the relative humidity of the air, the products of this reaction 

are UO2F2, hydrates, and HF-H2O fog, which will be seen as a white cloud. Equation 6-13 

shows a typical reaction with excess water: 

 UF6 (gas) + (2+4x)H2O(vapor) Æ UO2F2*2H2O (solid) + 4 HF*xH2O(fog) + heat Eq. 6-13 

These reactions, if occurring in the gaseous phase at ambient or higher temperatures, are very 

rapid and near instantaneous. 

The chemical characteristics of UF4 do not have the same reactivity as UF6 when released to 

the environment; that is, it reacts slowly with moisture at ambient temperature (DOE, 1999). 

UO2 does not react with water (Benedict, 1981), and it will oxidize to U3O8 in air at ambient 

temperature. Nonetheless, uranium is a heavy metal that, in addition to being radioactive, can 

have toxic chemical effects, primarily on the kidneys, if it enters the blood stream by means of 

ingestion or inhalation. HF is an extremely corrosive gas that can damage the lungs and cause 

death if inhaled at sufficiently high concentrations. 
6-11 

Appendix A, “List of Highly Hazardous Chemicals, Toxics and Reactives (Mandatory),” to 

29 CFR 1910.119 does not list SiF4 as a highly hazardous chemical. However, the applicant 

stated that SiF4 is corrosive to the eyes, skin, and respiratory tract (IIFP, 2012a). It may react 

violently and decompose exothermically with water or moisture in the air to form HF and silicic 

acid (NOAA, 2011a). Equation 6-14 shows the SiF4 decomposition by water: 

 SiF4 + 4H2O Æ 4HF + Si(OH)4 + heat Eq. 6-14 

Appendix A to 29 CFR 1910.119 lists BF3 as a highly hazardous chemical with a threshold 

quantity of 250 lb. The applicant stated that BF3 is corrosive and can cause irritation of the 

eyes, nose, throat, and skin (IIFP, 2012a). It hydrolyzes in moist air to form toxic and corrosive 

hydrogen fluoride, fluoroboric acid (HBF4), and boric acid (H3BO3) (IIFP, 2012a; 2012b, 

and NOAA, 2011b). Equation 6-15 shows the BF3 decomposition by water: 

 2BF3 + 3H2O Æ 2HF + HBF4 + B(OH)3 Eq. 6-15 

6.3.1.9 Materials of Construction 

The applicant considered the interactions between process equipment and process fluids/gases 

in the design of the proposed IIFP facility. IIFP stated in the LA (IIFP, 2012a) that the materials 

used for construction are compatible with the process operational physical parameters of 

temperature and pressure and corrosion resistant to UF6, AHF, SiF4, and BF3. The applicant 

commits to use materials for construction in accordance with applicable codes and standards, 

such as the 2009 edition of the International Building Code (IBC, 2009). 

The cylinders used to store and transport DUF6 consist of carbon steel and are standard 

DOT-approved containers, designed and fabricated in accordance with ANSI N14.1, “Uranium 

Hexafluoride Packaging for Transport” (ANSI, 2001). The cylinders are painted to resist 

corrosion caused by being stored outside in open air where they are exposed to atmospheric 

conditions. Also, the cylinders are routinely inspected to assess corrosion and corrosion rates. 

Packages of SiF4 and BF3 include small container cylinders and tube trailers; packages for AHF 

include tube trailers. The small cylinders and tube trailers are ASME Code rated pressure 

vessels and require DOT’s approval for transportation. 

6.3.1.10 Process Description Conclusion 

Based on the review and verification of the information provided in the LA (IIFP, 2012a) and the 

ISA Summary (IIFP, 20012b), including the consistency between the description of the process 

and the associated hazards, the NRC staff finds that the applicant has provided detailed 

process descriptions sufficient to enable an understanding of the chemical process hazards and 

the chemical hazards that could result from potential chemical interactions. The rationale for 

these findings is that the applicant provided expected process operating conditions, interactions 

between chemicals, and interactions between process chemicals and the equipment material 

that will be used to contain these process chemicals. In addition, the applicant’s information 

allowed the development of potential accident sequences. Therefore, the information that the 

applicant provided, as described above, meets the guidance in Section 6.4.3.1, bullets (1) and 

(2), of NUREG-1520 (NRC, 2002) and is acceptable. 
6-12 

6.3.2 Chemical Accident Sequences 

Tables 3-7, 3-8, 3-9, and 3-10 of the ISA Summary (IIFP, 2012b) describe the chemical accident 

sequences in the proposed DUF6 to DUF4 deconversion plant, SiF4 process, BF3 process, and 

auxiliary process systems, respectively. Tables 4-3, 4-4, 4-5, and 4-6 summarize the accident 

sequences and assign the indices used to determine the overall risk of each accident sequence 

in the proposed DUF6 to DUF4 deconversion plant, SiF4 process, BF3 process, and auxiliary 

process systems, respectively. The consequences of the chemical accident sequences 

identified in these tables have the potential to exceed the performance requirements of 

10 CFR 70.61. Table 5-1 shows the areas/systems for which the applicant identified chemical 

accident sequences with the potential to have intermediate or high consequences. The auxiliary 

process systems are located in the FEP oxide staging building, which is a common area that 

receives uranium oxide drums from both the SiF4 and BF3 processes. 

The regulations in 10 CFR 70.61(b) and 10 CFR 70.61(c) define high- and intermediateconsequence events, respectively. A high-consequence event is the release of licensed 

material or hazardous chemicals produced from licensed material that, if an individual were to 

be exposed, could endanger the life of a worker or lead to irreversible or other serious, 

long-lasting health effects to a member of the public. A high-consequence event also includes 

an intake of 30 milligrams (mg) of soluble uranium by a member of the public. An 

intermediate-consequence event is the release of licensed material or hazardous chemicals 

produced from licensed material that, if an individual were to be exposed, could lead to 

irreversible or other serious, long lasting health effects to a worker or could cause mild transient 

health effects to a member of the public. The NRC staff performed a risk-informed evaluation of 

the chemical accident sequences by reviewing the ISA Summary which contains the 

intermediate- and high-consequence events postulated by the applicant. Also, the NRC staff 

performed a vertical slice review of certain accident sequences. 

The NRC staff concludes that IIFP has identified appropriate chemical accident sequences 

based on the applicant’s use of a recommended process hazards analysis method (i.e., the 

What-if) contained in NUREG-1513 (NRC, 2001). The information provided by the applicant, as 

described above and in Chapter 3 of this SER, includes a list of the accident sequences with 

their respective consequence and likelihood identified in the ISA Summary (IIFP, 2012b) that 

involve hazardous chemicals produced from licensed material and chemical risks of plant 

conditions that affect the safety of licensed material. The applicant also described the 

postulated high-consequence events, how they will be detected, and the mitigative measures in 

Sections 4.1, 5.3, and 8.3, respectively, of the Emergency Plan (EP) (IIFP, 2012c). The actions 

described on the EP (IIFP, 2012c) are consistent with the consequences of the accident 

sequences identified in the ISA Summary (IIFP, 2012b). Based on the above, the information 

provided by the applicant meets the guidance in Section 6.4.3.1, bullet (2), of NUREG-1520 

(NRC, 2002) and is, therefore, acceptable. 

6.3.3 Chemical Accident Consequences 

The chemical exposure limits proposed by the applicant for UF6, UF4, HF, SiF4, and BF3 are 

based on the acute exposure guideline limit (AEGL) values. Specifically, for AHF, SiF4, and BF3

these values apply, from an NRC licensing perspective, before these compounds are separated 

from licensed material because these compounds have licensed material as precursors. Once 

separated from licensed material, OSHA regulates these compounds. It should be noted that 

HF is under NRC’s regulatory jurisdiction if it is produced from the reaction of UF6 with water 

(see Equations 6-12 and 6-13 above) and from the reaction of SiF4 and BF with water (see 
6-13 

Equations 6-14 and 6-15 above) if it forms because of a release of SiF4 and BF3 in the FEP 

process building. The conversion to HF could affect the safe handling of licensed material in 

this building (i.e., UF4 and UO2). The applicant proposed to use the 10-minute AEGL values for 

exposures to workers with a duration of 10 minutes or less. The worker is trained to take proper 

actions (i.e., escape) upon sensing the initial effects of released hazardous material. For the 

public, the exposure duration was assumed to be 30 minutes. This is consistent with one of the 

conservative factors listed in NUREG-1140, “A Regulatory Analysis on Emergency 

Preparedness for Fuel Cycle and Other Radioactive Material Licensees,” issued January 1988 

(NRC, 1988). 

There are no AEGL values for UO2. For this reason, the applicant derived the chemical 

exposure limits for UO2 from the emergency response planning guidelines (ERPGs). ERPG 

values are based on exposures for up to 60 minutes, and the applicant has scaled the 

exposures to 10 minutes and 30 minutes for the worker and public, respectively. 

For soluble uranium (i.e., UF6 or UO2F2) intakes, the applicant proposed to use 75 mg of soluble 

uranium intake as the high-consequence threshold for the worker. The NRC staff finds this 

value acceptable because it is based on the model in ICRP 66 (ICRP-66), “Human Respiratory 

Tract Model for Radiological Protection,” that derives the systematic burned value of 

0.3 milligram per kilogram of body weight (ICRP, 1994). The NRC staff approved the use of the 

ICRP-66 model and determined that using 75 mg of soluble uranium as the threshold for 

permanent renal damage is consistent with a high-consequence event to a worker, as defined in 

10 CFR 70.61(b)(4)(i) (NRC, 2007b). For the public, the high-consequence threshold for a 

soluble uranium intake is 30 mg. This value is acceptable because it is based on the 

performance requirement in 10 CFR70.61(b)(3). 

The comparison of the AEGLs or ERPGs with the calculated airborne concentrations allows the 

applicant to classify a high, intermediate, or low chemical consequence level. Table 5-2, 

reproduced from the LA (IIFP, 2012a), presents the chemical consequence levels and values. 

For worker and public chemical exposures, Tables 5-3 and 5-4, taken from the ISA Summary 

(IIFP, 2012b), present more detailed chemical consequence values for licensed material and 

hazardous chemicals produced from licensed material, respectively. 

Table 6-1 Chemical Consequence Levels and Values (IIFP, 2012a) 

Consequence Workers Offsite Public 

High CE > AEGL-3 

(10-minute exposure) 

CE > AEGL-2 

(30-minute exposure) 

Intermediate AEGL-2 < CE ≤ AEGL-3 

(10-minute exposure) 

AEGL-1 CE ≤ AEGL-2 

(30-minute exposure) 

Low Lower than CE above Lower than CE above 

Note: CE = chemical exposure 

Table 6-2 Worker Chemical Exposure Values (10 Minutes) 

(IIFP, 2012b) 

Chemical High Consequences 

AEGL-3/(mg/m3

) 

Intermediate 

Consequences 

AEGL-2/(mg/m3

) 

UF6 216 28 

UF4 216 28 

UO2F2 216 28 
6-14 

HF 139 77.7 

SiF4 81 27 

BF3 140 47 

UO2 (ERPG) 300 201 

Table 6-3 Public Chemical Exposure Values (30 Minutes) 

(IIFP, 2012b) 

Chemical High Consequences 

AEGL-2/(mg/m3

) 

Intermediate 

Consequences 

AEGL-1/(mg/m3

) 

UF6 19 3.6 

UF4 19 3.6 

UO2F2 19 3.6 

HF 28 0.82 

SiF4 18 0.21 

BF3 47 2.5 

UO2 (ERPG) 32 0.68 

The NRC staff’s review of the ISA and supporting documentation found that the source term 

values are reasonable because the applicant provided the material at risk values for each 

building and followed the methodology described in NUREG/CR-6410, “Nuclear Fuel Cycle 

Facility Accident Analysis Handbook,” issued March 1998 (NRC, 1998). The applicant used a 

computer code named HGSYSTEM, Version 3.0, to model the site boundary atmospheric 

dispersion of HF. The applicant also used modeling methods for source term determination, 

release fraction, dispersion factors, and meteorological conditions. The NRC staff performed an 

independent evaluation of the consequences of potential accidents identified in the ISA 

Summary. The NRC staff evaluated a representative selection of the types of accidents 

postulated for the proposed IIFP facility. Appendix B to this SER discusses this evaluation. The 

information provided by IIFP meets the guidance of Section 6.4.3.1, bullets (3) and (5), of 

NUREG-1520 (NRC, 2002) because the applicant (1) identified and used appropriate 

techniques in estimating the concentration of hazardous chemicals produced from licensed 

material, (2) used the performance requirements criteria of 10 CFR 70.61, and (3) ensured that 

the consequence analysis conformed to the guidance in NUREG/CR-6410 (NRC, 1998). 

Therefore, the NRC staff finds the applicant’s proposed methodology for source term 

determination and consequence analysis to be acceptable. 

6.3.4 Items Relied on for Safety and Management Measures 

6.3.4.1 Chemical Process Items Relied on for Safety 

Table 6-1 of the ISA Summary (IIFP, 2012b) is a descriptive list of all IROFS. It includes (1) the 

type of control of the IROFS (i.e., administrative control, enhanced administrative control, active 

or passive-engineered control); (2) the safety function and how it is implemented; (3) the 

management measures applied to ensure that the IROFS is available and reliable when 

needed; and (4) the index used to determine the overall risk of a particular accident sequence 

and its basis. Table 8-1 of the ISA Summary (IIFP, 2012b) identifies the sole IROFS credited to 

prevent a high-consequence event, and Table 8-2 identifies the sole IROFS credited to prevent 

an intermediate-consequence event. All of the chemical process sole IROFS are passiveengineered controls. The applicant stated in the ISA Summary that a passive-engineered 

control does not require human intervention to carry out its safety function; and it relies on 
6-15 

natural forces, such as gravity or natural convection, to maintain safe process conditions. This 

description is consistent with the term as used in NUREG-1520, Glossary (NRC, 2002). 

Specifically for chemical process safety, the passive-engineered control is the structural design 

criteria of certain process equipment in the DUF6 to DUF4 deconversion plant and SiF4 and BF3

processes. The NRC reviewed the IROFS and management measures described above and 

finds they provide reasonable assurance that process equipment will remain available and 

reliable. In addition, NRC staff finds that the process equipment will not suddenly fail, leading to 

a release of licensed material or hazardous chemicals produced from licensed material, without 

a detectable process upset or external impact. The applicant has adequate detection systems, 

procedures, and shutdown capabilities to ensure that the structural design criteria of certain 

process equipment is not challenged. 

Relying on multiple IROFS for an accident sequence provides the protection against releases of 

licensed material or hazardous chemicals produced from licensed material. For example, to 

prevent a release of licensed material caused by heating an overfilled UF6 cylinder, the applicant 

proposed an IROFS procedure that prevents an overfilled cylinder to be placed in the autoclave. 

The IROFS procedure is complemented by an IROFS accountability scale and by the IROFS 

autoclave integrity. 

The applicant postulated accident sequences that rely on sole IROFS. The chemical process’s 

sole IROFS prevent intermediate-consequence events, and it is the structural design criteria of 

certain process equipment (discussed in the first paragraph of this section). 

The NRC staff reviewed the listed IROFS, process descriptions, and process flow diagrams 

provided in Chapters 6 and 3 of the ISA Summary (IIFP, 2012b) and the engineering drawings, 

respectively, to identify where each IROFS would be used and how the IROFS would function to 

prevent the accident sequence or mitigate its consequences. The identified IROFS protect 

against a loss of confinement of licensed material or hazardous chemicals produced from 

licensed material during operation of the facility. 

Based on this system-level review, the NRC staff concludes that the applicant meets the 

guidance of Section 6.4.3.2, bullet (2), of NUREG-1520 (NRC, 2002) because the ISA Summary 

(IIFP, 2012b) identified chemical process IROFS to prevent or mitigate the consequences of 

accident sequences that involve the chemical hazards of licensed material and hazardous 

chemicals produced from licensed material. In addition, the ISA Summary (IIFP, 2012b) 

identified the hazards being mitigated and the risk category of each accident sequence. 

6.3.4.2 Management Measures 

The following sections briefly discuss the integration of chemical safety into the elements of 

management measures. The applicant identified management measures to ensure that 

chemical safety IROFS would be available and reliable to perform their safety function when 

needed. The applicant proposed to implement a graded approach based on the level of 

protection needed by each IROFS to meet the performance requirements. Chapter 11 of this 

SER provides an overall evaluation of the management measures applied to the proposed IIFP 

facility. 

6.3.4.2.1 Configuration Management 

The Configuration Management Program’s proposal in Section 11.1 of the LA for the proposed 

IIFP facility, includes those elements that ensure that the facility’s technical baseline is 
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thoroughly documented and maintained. The Configuration Management Program also ensures 

that all safety, security, and licensing organizations review and approve changes to the 

technical baseline. The technical baseline consists of facility drawings, procedures, 

specifications, and other technical documents—including the ISA (IIFP, 2012b). 

The applicant stated that it will submit changes to the technical baseline that require prior NRC 

approval in a license amendment request (as required by 10 CFR 70.72(d)(1)) and will not 

implement such changes without prior NRC’s approval. The regulations in 10 CFR 70.72(c) 

include the criteria for changes that require NRC pre-approval. The applicant also stated that it 

will submit changes that do not require prior NRC approval annually, as well as the revised ISA 

Summary pages, as required by 10 CFR 70.72(d)(2) and 10 CFR 70.72(d)(3), respectively 

(IIFP, 2012b). 

6.3.4.2.2 Maintenance 

The applicant outlined the maintenance and functional testing programs in Section 11.2 of the 

LA. Maintenance activities provide reasonable assurance that IROFS will be available and 

reliable to perform their safety functions when needed. Appropriate plant management is 

responsible for ensuring operational readiness of IROFS under this control. For this reason, the 

maintenance function is administratively closely coupled to the engineering function 

(IIFP, 2012a). The applicant will evaluate the impact of maintenance activities on the specific 

systems and on other nearby systems (IIFP, 2012b). 

Maintenance activities generally fall into the following categories: 

• surveillance and monitoring 

• testing 

• preventive maintenance 

• corrective maintenance 

Surveillance and monitoring identify conditions that require corrective maintenance and ensure 

that preventive maintenance is effective. The chemical process safety discipline within the 

applicant’s organization will evaluate the results of surveillance and monitoring activities to 

determine any impact on the ISA and any updates needed. The applicant stated that testing 

includes functional tests, performance tests, software checks and updates, and instrument 

calibration. In addition, the applicant will develop testing plans for IROFS (including chemical 

safety) before the IROFS installation, and IROFS will be tested following maintenance and on 

an annual basis (unless otherwise noted in Chapter 6 of the ISA Summary [IIFP, 2012b]). The 

applicant stated that preventive maintenance includes periodic refurbishment or like-kind 

replacement of IROFS at a frequency based on the expected life of the IROFS along with the 

relative importance of the IROFS in meeting the performance requirements and the results of 

surveillance and monitoring. Sole IROFS will be refurbished or replaced at a greater frequency. 

Corrective maintenance includes repair of like-kind replacement of equipment that has failed to 

perform or is performing outside of desired safety and process parameter limits (IIFP, 2012b). 
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6.3.4.2.3 Training and Qualifications 

Section 11.3 of the LA describes the training program for operations of the facility, including 

preoperational functional testing and initial startup testing. The applicant proposed to train and 

qualify employees to ensure safe operation of the IIFP facility. The training program 

requirements are applicable, but not limited, to employees who perform activities that affect 

IROFS or items that may affect the function of IROFS (IIFP, 2012a). Technical training is 

provided to assist employees in gaining an understanding of applicable fundamentals, 

procedures, and practices related to IROFS. Employees are provided with formal classroom 

training along with specific on-the-job training. IIFP will use a job task analysis, as needed, to 

supplement training when tasks associated with IROFS are involved (IIFP, 2012b). 

Chemical safety (hazard communication) training is provided to all employees as one of the 

multiple topics included in the general employee training. This general employee training also 

covers general industrial safety. The applicant also conducts more specific training in various 

aspects of industrial and chemical safety protection to train new employees in specific job duties 

and to provide refresher training topics to workers depending on employee job responsibilities. 

In particular, any employee or contractor using hazardous materials is trained to ensure safe 

handling, use, and disposal of such materials. 

IIFP provides annual training to emergency response personnel, as discussed in Section 10.2.5 

of the IIFP EP (IIFP, 2012c). The applicant stated that the emergency response personnel are 

prepared to respond to various emergency conditions, including a chemical accident. 

6.3.4.2.4 Procedures 

The applicant proposed that all activities involving IROFS are conducted in accordance with 

approved procedures. Procedures are used to ensure that activities involving IROFS are 

carried out in a safe manner and in accordance with regulatory requirements (IIFP, 2012a). 

Section 11.4 of the LA (IIFP, 2012a) describes and outlines the types of procedures, including 

(1) operating procedures, (2) administrative procedures, (3) maintenance procedures, and 

(4) emergency procedures. 

The operating procedures are used to directly control production and to clearly identify 

applicable safety limits and IROFS. Administrative procedures are used to perform activities 

that support production, including management measures such as, but not limited to, chemical 

safety. Maintenance procedures are used to maintain, test, and calibrate facility IROFS. 

Finally, the emergency procedures outline duties, responsibilities, action levels, and actions to 

be taken by responders pertinent to specific accident scenarios and other categorized 

non-routine operational events. In addition, administrative procedures ensure that individuals 

and groups with assigned responsibilities in an emergency have easy access to a current copy 

of each procedure that pertains to their functions (IIFP, 2012a). 

6.3.4.2.5 Audits and Assessments 

Chemical safety is one of the areas in which the applicant proposed to conduct audits and 

assessments. Section 11.5 of the LA (IIFP, 2012a) describes audits and assessments. Audits 

are focused on verifying compliance with regulatory and procedural requirements and licensing 

commitments. Assessments are focused on evaluating the effectiveness of activities and 

ensuring that IROFS, and any items that affect the function of IROFS, are reliable and available 

to perform their intended safety functions (IIFP, 2012a). 
6-18 

The results of audits and assessments are documented and reported as specified in plant 

procedures, which include the managers responsible for the activities audited or assessed 

(IIFP, 2012a). The Corrective Action Program tracks audit and assessment results 

(IIFP, 2012a). 

6.3.4.2.6 Incident Investigation and Corrective Actions 

The applicant described the incident investigations and corrective action process in Section 11.6 

of the LA (IIFP, 2012a). The applicant proposed to use an incident investigation process to 

report deficiencies, abnormal events, and potentially unsafe conditions or 

activities (IIFP, 2012a). When an incident occurs, facility management will form a qualified team 

to determine root causes of the event and develop recommendations to reduce the likelihood of 

recurrence. Lessons learned will be developed, and unaffected organizations can review their 

operations for similar type initiators. Written incident investigation and corrective action 

procedures address the process of incident identification, investigation, root-cause analysis, 

recording, and followup; IIFP staff use these procedures when performing incident 

investigations and corrective actions. These procedures also address hazardous chemical 

safety requirements (IIFP, 2012a). Corrective actions are assigned and tracked 

programmatically to ensure that timely and adequate corrections to deficiencies are 

incorporated. 

6.3.4.2.7 Records Management 

The applicant will maintain the following records, which are described in Section 11.7 of the LA 

(IIFP, 2001a), for the proposed IIFP facility: 

• results of surveys to determine the dose from external sources and used in the 

assessment of individual dose equivalents 

• results of measurements and calculations used to determine individual intakes of 

radioactive material and used in the assessment of internal dose 

• results of air sampling, surveys, and bioassays 

• results of measurements and calculations used to evaluate the release of radioactive 

effluents to the environment 

• records of spills or other unusual occurrences involving the spread of contamination in 

and around the facility, equipment, or site 

• as-built drawings and modifications of structures and equipment in restricted areas 

where radioactive materials are used or stored 

• IROFS design specifications and maintenance records 

• training and qualification records 

• audit, assessment, and inspection results 
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• incident investigation reports 

• quality assurance records. 

6.3.4.3 Items Relied on for Safety and Management Measures Conclusion 

The information IIFP provided, as described above, meets the guidance in Section 6.4.3.2, 

bullet (2), of NUREG-1520 (NRC, 2002) and is acceptable because the applicant has 

adequately identified the administrative and engineered controls (IROFS) to prevent chemical 

accident sequences or mitigate their consequences at the proposed facility. The applicant also 

identified the hazards being mitigated and the risk category. 

In addition, the information IIFP provided, as described above and in Chapter 11 of this SER, 

meets the guidance in Section 6.4.3.2, bullet (3), of NUREG-1520 (NRC, 2002) and is 

acceptable because the applicant sufficiently described its procedures to ensure the reliable 

operation of engineered controls and the correct implementation of administrative controls. 

6.3.5 Emergency Management 

As described further in Chapter 8 of this SER and consistent with the requirements in 

10 CFR 40.31(j)(1)(ii), IIFP has submitted an EP and program which includes response to 

mitigate the potential impact of any process chemical release, including requirements for 

notification and reporting of accidental chemical releases. The emergency response team is 

outfitted, equipped, and trained for hazardous material response; and local agencies can 

supplement the response with additional response teams. The applicant has an agreement with 

the Fire Department of the City of Hobbs, NM, to supplement the IIFP facility’s emergency 

response team with additional response teams, if needed. The applicant stated in 

Section 10.2.4 of the EP (IIFP, 2012c) that IIFP facility personnel meet at least every 2 years 

with each offsite assistance group to accomplish training and review items of mutual interest, 

including relevant changes to the Emergency Management Program. 

In Chapter 13 of the EP (IIFP, 2012c) for the proposed IIFP facility, the applicant committed to 

meeting the requirements of the Emergency Planning and Community Right-to-Know Act of 

1986 (Title III, Public Law 99-499), as required by 10 CFR 40.31(j)(3)(xiii). The applicant 

identified the material safety data sheets (MSDSs) that will be available to local agencies. The 

applicant will provide the MSDSs for UF6, UF4, UO2, U3O8, HF, SiF4, and BF3 to local agencies. 

The applicant also committed to providing additional MSDSs to local agencies if any significant 

quantities of hazardous chemicals are added to the facility inventory. Table 9 of the EP 

(IIFP, 2012c) provides the major chemicals used at the IIFP facility and their locations. 

The NRC staff finds that the applicant has provided reasonable assurance that measures used 

to mitigate the consequences of accident sequences identified in the ISA Summary 

(IIFP, 2012b) are consistent with actions described in Chapter 8 of NUREG-1520 (NRC, 2002). 

The information IIFP provided, as described above, meets the guidance in Section 6.4.3.1, 

bullet (2), of NUREG-1520 (NRC, 2002), and is acceptable because the applicant described the 

postulated high-consequence events, how they will be detected, and the mitigative measures in 

Sections 4.1, 5.3, and 8.3, respectively, of the EP (IIFP, 2012c). The actions described in the 

EP (IIFP, 2012c) are consistent with the consequences of the accident sequences identified in 

the ISA Summary (IIFP, 2012b). 
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6.3.6 Baseline Design Criteria and Defense-in-Depth 

The applicant provided design-basis information for chemical process safety IROFS for the 

proposed facility in the LA (IIFP, 2012a) and ISA Summary (IIFP, 2012b). For chemical 

protection, 10 CFR 70.64(a)(5) states the following: 

Chemical protection. The design must provide for adequate protection against 

chemical risks produced from licensed material, facility conditions which affect 

the safety of licensed material, and hazardous chemicals produced from licensed 

material. 

Applicable to new facilities, the regulations in 10 CFR 70.64(b) require the facility and system 

design and facility layout to be based on defense-in-depth practices. The regulations also 

require that the design must incorporate, to the extent practicable, preference for the selection 

of engineered controls over administrative controls to increase overall system reliability and 

features that enhance safety by reducing the challenges to IROFS. The applicant provided 

information in the process description of the ISA Summary (IIFP, 2012b) regarding non-IROFS 

safeguards to be used as defense-in-depth to reduce the challenge to IROFS. 

The chemicals of concern are UF6, UF4, UO2, HF, SiF4, BF3, and H2. Chapters 3 and 6 of the 

ISA Summary (IIFP, 2012b), respectively, provide details of the design and safety features of all 

chemical process systems. The applicant’s design of the chemical process systems includes 

numerous controls, in addition to the IROFS, for maintaining safe conditions during operation. 

The applicant accomplishes this through several means, including the following: 

• managing the arrangement and size of material containers and processes 

• selecting and using materials compatible with process chemicals 

• providing inherently safer operating conditions (e.g., limiting steam temperature in the 

autoclave, ensuring a slight vacuum or near atmospheric pressure in the rotary 

calciners) 

• providing process interlocks, controls, and alarms within the process. 

The NRC staff reviewed the applicant’s proposed design of the proposed FEP/DUP contained in 

Section 3.1 of the ISA Summary and the process hazards description in Section 3.2 

(IIFP, 2012b). The NRC staff notes that the applicant preferred engineered controls over 

administrative controls in the selection of IROFS. This is demonstrated by the fact that no 

accident sequence is prevented or mitigated only by administrative controls. Usually the 

accident sequences are prevented or mitigated by a combination of administrative and 

engineered controls. As discussed in Section 6.3.4 of this SER, when sole IROFS are credited 

to prevent an accident sequence, the selected IROFS is an engineered control. The NRC staff 

also notes that the applicant preferred to prevent an accident sequence rather than mitigate it. 

The majority of IROFS are preventive. In contrast, mitigative IROFS are credited in a few of the 

accident sequences identified by the applicant. 

The NRC staff notes that the applicant’s design of the UF6 feeding system uses a cylinder 

qualified under ANSI N14.1 (ANSI, 2001) as the primary confinement vessel and an ASME 

Code pressure vessel as a secondary confinement system. The applicant proposed a limit of 
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235 degrees F for the temperature of the steam used to vaporize the UF6 in the cylinder inside 

the autoclave. This temperature limit provides a margin of safety from the limit used in the ANSI 

N14.1 standard (i.e., 250 degrees F) (ANSI, 2001). The rationale for the limit of 250 degrees F 

is, if a full UF6 cylinder (i.e., 27,560 lb for a 48-Y cylinder) is heated to this temperature, the 

cylinder would have 5 percent of its total volume available. The NRC staff concludes that this 

design approach for the feeding portion of the process is acceptable because it uses recognized 

nuclear fuel cycle industry codes and standards and provides an adequate safety margin. 

The NRC staff reviewed the results of the applicant’s What-if analysis, as discussed in 

Chapter 3 of this SER. This method is widely used in the chemical industry during the design 

phase to identify operability and safety issues and is identified as an acceptable method in 

Section 2.4 of NUREG-1513 (NRC, 2001). As applied to the proposed depleted uranium 

deconversion and fluorine extraction processes, the What-if analysis considered a variety of 

internal process, facility, and external hazards that could breach the process and release 

licensed material and hazardous chemicals produced from licensed material. Tables 4-3, 4-4, 

4-5, and 4-6 of the ISA Summary (IIFP, 2012b) present the results of the applicant’s ISA. The 

tables contain information concerning the accident sequences identified as a result of the 

What-if analysis, the unmitigated risk of each applicant-identified accident sequence, and the 

IROFS applied to prevent the accident sequence or mitigate its consequences. The NRC staff 

also reviewed selected high-consequence and intermediate-consequence accident scenarios to 

confirm that the applicant addressed chemical events that could exceed the performance 

requirements of 10 CFR 70.61. 

Based on the above, the NRC staff concludes that the information IIFP provided and its 

proposed design meet the guidance in Section 6.4.3.3 of NUREG-1520 (NRC, 2002) and 

provide for adequate protection against chemical risks produced from licensed materials, facility 

conditions which affect the safety of licensed material, and hazardous chemicals produced from 

licensed material, thus meeting the requirements of 10 CFR 70.64(a)(5) and 10 CFR 70.64(b). 

6.4 Evaluation Findings

The NRC staff evaluated the application using the criteria previously listed. Based on the 

review of the LA (IIFP, 2012a) and ISA Summary (IIFP 2011b), the NRC staff concludes that the 

applicant has described and assessed accident sequences that can result from the handling, 

storage, or processing of licensed materials and that can potentially have significant chemical 

consequences and effects. The applicant has prepared a hazard analysis that identified and 

evaluated those chemical process hazards and potential accidents and established safety 

controls providing reasonable assurance of safe facility operation. To ensure that the 

performance requirements in 10 CFR Part 70 are met, the applicant stated that controls are 

maintained, available, and reliable to perform their safety-related functions when needed. The 

NRC staff has reviewed these safety controls and the applicant’s plan to managing chemical 

process safety and finds them acceptable. 

The NRC staff concludes that the applicant’s plan for managing chemical process safety and 

chemical process safety controls meets the requirements of 10 CFR Part 70, Subpart H, and 

10 CFR Part 40 and therefore provides reasonable assurance that the public health and safety 

and the environment will be protected. 
6-22 

6.5 References

(ASME, 2004) American Society of Mechanical Engineers, “Boiler and Pressure Vessel Code,” 

Section VIII, Division 1, 2004 

(ANSI, 2001) American National Standards Institute, “Uranium Hexafluoride—Packaging for 

Transport,” ANSI N14.1, February 2001. 

(Benedict, 1981) Benedict, Mason, Thomas H. Pigford, and Hans Wolfgang Levi, Nuclear 

Chemical Engineering (2nd edition), (pp: 223). McGraw-Hill, 1981. 

(Benítez, 1993) Benítez, Jaime, Process Engineering and Design for Air Pollution Control, 

(pp: 333). Prentice-Hall. 1993. 

(ICRP, 1994) International Commission on Radiological Protection, “Human Respiratory Tract 

Model for Radiological Protection,” Publication 66, 1994. 

(IBC, 2009) International Code Council, International Building Code, 2009. 

(IIFP, 2012a) International Isotopes Fluorine Products, Inc., “Fluorine Extraction Process and 

Depleted Uranium Deconversion Plant (FEP/DUP) License Application, Revision B,” May 2012, 

Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML12123A245. 

(IIFP, 2012b) International Isotopes Fluorine Products, Inc., “ISA Summary Rev. B for IIFP,” 

May 2012, Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML12123A245. 

(IIFP, 2012c) International Isotope Fluorine Products, Inc., “Emergency Plan Rev B of IIFP 

License Application,” May 2012, Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML12123A245. 

(IIFP, 2011) International Isotopes Fluorine Products, Inc., “Official Responses to Chemical 

Process Safety RAIs,” March 2011 Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML110950330. 

(Kutz, 2007) Kutz, Myer, Environmentally Conscious Alternative Energy Production, John Wiley 

& Sons. 2007. 

(NOAA, 2011a) National Oceanic and Atmospheric Administration, “Chemical Datasheet—

Silicon Tetrafluoride,” http://cameochemicals.noaa.gov/chemical/1449, Access Date: 

January 21, 2011. 

(NOAA, 2011b) National Oceanic and Atmospheric Administration, “Chemical Datasheet—

Boron Trifluoride,” http://cameochemicals.noaa.gov/chemical/255, Access Date: 

January 21, 2011. 
6-23 

(NRC, 2010) U.S. Nuclear Regulatory Commission, “July 6–8, 2010, Summary of Site Visit to 

Conduct Integrated Safety Analysis Horizontal and Vertical Slice and Discussion Draft Requests 

for Additional Information with International Isotopes Fluorine Products, Inc.,” July 2010. 

ADAMS Accession No. ML102070210 

(NRC, 2007a) U.S. Nuclear Regulatory Commission, Staff Requirements Memorandum to 

SECY-07-0146, “Regulatory Options for Licensing New Uranium Conversion and Depleted 

Uranium Deconversion Facilities,” October 10, 2007. 

(NRC, 2007b) U.S. Nuclear Regulatory Commission, “Reply to Proposed Usage of International 

Commission on Radiological Protection 66 in Determination of Soluble Uranium Intake 

Threshold,” July 2007. ADAMS Accession No. ML072010285 

(NRC, 2002) U.S. Nuclear Regulatory Commission, “Standard Review Plan for the Review of a 

License Application for a Fuel Cycle Facility,” NUREG-1520, Rev. 0, March, 2002. 

(NRC, 2001) U.S. Nuclear Regulatory Commission, “Integrated Safety Analysis Guidance 

Document,” NUREG-1513, May 2001. 

(NRC, 1998) U.S. Nuclear Regulatory Commission, “Nuclear Fuel Cycle Facility Accident 

Analysis Handbook,” NUREG/CR-6410, March 1998. 

(NRC, 1997) U.S. Nuclear Regulatory Commission, “Chemical Process Safety at Fuel Cycle 

Facilities,” NUREG-1601, August 1997. 

(NRC, 1988) U.S. Nuclear Regulatory Commission, “A Regulatory Analysis on Emergency 

Preparedness for Fuel Cycle and Other Radioactive Material Licensees,” NUREG-1140, 

January 1988. 

 
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7-1 

7.0 FIRE SAFETY 

The purpose of this review is to determine, with reasonable assurance, whether International 

Isotopes Fluorine Products, Inc. (IIFP or the applicant), has designed a facility with adequate 

protection against fires and explosions that could affect the safety of licensed materials and thus 

present an increased radiological risk. The review should also establish that the applicant has 

considered the radiological consequences of fires and will institute suitable safety controls to 

protect workers, the public, and the environment. 

Details related to items relied on for safety (IROFS) have been marked by the applicant as 

"Security-Related Information," pursuant to Title 10 of the Code of Federal Regulations 

(10 CFR) 2.390, and the Commission has determined to include this information in the public 

version of this report. 

7.1 Regulatory Requirements

The regulatory basis for the fire safety review should be the general and additional contents of 

application, as required by 10 CFR 40.32, entitled “General Requirements for Issuance of 

Specific Licenses.” In addition, the fire safety review should focus on providing reasonable 

assurance of compliance with the following regulations: 

• 10 CFR 70.61, “Performance Requirements” 

• 10 CFR 70.62, “Safety Program and Integrated Safety Analysis” 

• 10 CFR 70.64, “Requirements for New Facilities or New Processes at Existing Facilities” 

• 10 CFR 70.65, “Additional Content of Applications.” 

7.2 Regulatory Acceptance Criteria

Chapter 7 of NUREG-1520, “Standard Review Plan for the Review of a License Application for a 

Fuel Cycle Facility” (NRC, 2002), contains the guidance applicable to the U.S. Nuclear 

Regulatory Commission’s (NRC’s) review of the fire safety description section of the license 

application (LA) (IIFP, 2012a). This chapter is applicable in its entirety. Sections 7.4.3.1 

through 7.4.3.5 of NUREG-1520 (NRC, 2002) provide the acceptance criteria applicable to this 

review. 

7.3 Staff Review and Analysis

This section addresses the NRC staff’s review of facility fire protection, including fire safety 

management measures, fire hazards analysis, facility fire protection, process fire safety, and fire 

safety and emergency response, as presented in the LA (IIFP, 2012a); the Integrated Safety 

Analysis (ISA) Summary (IIFP, 2012b). 
7-2 

7.3.1 Fire Safety Management Measures 

The applicant will implement fire safety management measures as described in Chapter 11 of 

the LA (IIFP, 2012a). Management measures applicable to fire safety include configuration 

management, maintenance, training and qualifications, procedures development and 

implementation, audits and assessments, incident investigations and corrective actions process, 

records management and document control, and quality assurance program elements. These 

measures will ensure that IROFS related to fire protection are available and reliable during 

normal operations, anticipated (off-normal) events, and accidents. The applicant will follow the 

codes and standards, as listed in Table 7-1 of the LA (IIFP, 2012a), which are applicable to the 

individual fire safety management measures. Chapter 11 of this Safety Evaluation Report 

(SER) evaluates management measures. 

7.3.1.1 Management Policy and Direction 

The Regulatory Affairs and Quality Assurance Director is responsible for fire protection and is 

assisted by the Environmental Safety and Health Manager, who is responsible for the 

day-to-day safe operation of the facility, including fire safety. These individuals are part of the 

Facility Safety Review Committee which, with managers from other disciplines, integrates facility 

modifications. Fire safety personnel who are trained in fire protection and have nuclear fire 

safety experience assist the Environmental Safety and Health Manager. The fire protection staff 

is responsible for the following: 

• fire protection program and procedural requirements 

• fire prevention activities (i.e., administrative controls and training) 

• maintenance, surveillance, and quality of the facility fire protection features 

• control of design changes, as related to fire protection 

• documentation and recordkeeping, as related to fire protection 

• organization and training of the fire brigade 

• pre-fire planning. 

Fire prevention at the facility consists of administrative controls to (1) govern the handling of 

transient combustibles, (2) control ignition sources, (3) ensure that open flames or combustion 

generated smoke is not used for leak testing, (4) conduct periodic fire prevention inspections, 

(5) perform periodic housekeeping inspections, and (6) implement a system to control the 

disarming of the various types of fire detection or fire suppression systems. The inspection, 

testing, and maintenance of fire protection systems will comply with nationally recognized 

industry standards. Section 7.3.3 of this chapter provides further information concerning the fire 

detection and suppression systems. 

7.3.1.2 Fire Safety Management Measures Conclusions 

Based on its review, the NRC staff concludes that the applicant’s fire safety management 

measures are acceptable for the following reasons: 
7-3 

• Consistent with the acceptance criteria in Section 7.4.3.1 of NUREG-1520 (NRC, 2002), 

the applicant’s fire safety management measures identify a senior level manager who 

has the authority and staff to ensure that fire safety receives appropriate priority. 

• Consistent with the acceptance criteria in Section 7.4.3.1 of NUREG-1520 (NRC, 2002), 

the applicant’s fire safety management measures identify a facility safety committee 

staffed by managers of different disciplines to integrate facility modifications. 

• Consistent with the acceptance criteria in Section 7.4.3.1 of NUREG-1520 (NRC, 2002), 

the applicant’s fire safety management measures include fire prevention; inspection, 

testing, and maintenance of fire protection systems; fire brigade qualifications, drills, and 

training; and pre-fire plans as recommended by National Fire Protection Association 

Standard (NFPA) 801, “Standard for Fire Protection for Facilities Handling Radioactive 

Materials” (NFPA, 2008c). 

• Consistent with the acceptance criteria in Section 7.4.3.1 of NUREG-1520 (NRC, 2002), 

the applicant’s fire safety management measures are documented in sufficient detail to 

identify their relationship to, and functions for, normal operations; anticipated (off-normal) 

events; and accident safety (i.e., IROFS). 

• Consistent with the acceptance criteria from Section 7.4.3.1 of NUREG-1520 

(NRC, 2002), the applicant’s fire safety management measures will ensure that the 

IROFS, as identified in the ISA Summary (IIFP, 2012b), are available and reliable and 

will ensure that the facility maintains fire safety awareness among employees, controls 

transient ignition sources and combustibles, and maintains a readiness to extinguish or 

limit the consequences of fire. 

Based on the above findings, the NRC staff finds that the applicant's fire safety management 

measures meet the requirements of 10 CFR 40.32, 10 CFR 70.61, 10 CFR 70.64, and 

10 CFR 70.65 as they pertain to the fire protection aspects of the facility. 

7.3.2 Fire Hazards Analysis 

As part of its ISA, the applicant performed a fire hazards analysis (FHA), consistent with the 

guidance in NUREG-1520 (NRC, 2002) and NFPA 801 (NFPA, 2008c), which considered each 

process area and described, by fire area, the fuel loading, fire scenarios, methods of 

consequence analysis, potential consequences, and mitigative controls. 

Using the results from the FHA, the ISA Summary (IIFP, 2012b) describes, qualitatively, the 

potential credible fire accident scenarios and associated risks for the facility. The applicant 

postulated and evaluated the following key fire accident scenarios: 

• fire involving a transport truck or external wild lands (including an evaluation of the 

cylinder staging and storage pads) 

• fire involving a cylinder hauler 

• fire or explosion involving underground natural gas and propane lines 
7-4 

• fire in various areas of the depleted uranium hexafluoride (DUF6) conversion plant 

(depleted uranium tetrafluoride (DUF4) and DUF6 buildings) 

• fire in various areas of the silicon tetrafluoride (SiF4) plant (fluorine extraction process 

[FEP] building) 

• fire in various areas of the boron trifluoride (BF3) plant (FEP building). 

7.3.2.1 Items Relied on for Safety Related to Fire Safety 

The applicant identified a set of IROFS that would ensure that the likelihood of a fire causing 

high-consequence events is highly unlikely and the likelihood of a fire causing intermediate 

consequence events is unlikely. Table 6-1 of the ISA Summary (IIFP, 2012b) lists these IROFS. 

The NRC staff considers the failure probability indices assigned to these IROFS to be 

achievable, with their respective bases, as described in Section 6.1 of the ISA Summary 

(IIFP, 2012b). Section 6.1 of the ISA Summary (IIFP, 2012b) provides proposed surveillance 

frequencies, safety margins, and other measures that will support the low failure probabilities 

assigned to these IROFS, which are in accordance with the NRC guidance provided in 

Table A-9 of NUREG-1520 (NRC, 2002). Additionally, the applicant has followed the 

requirements of 10 CFR 70.64(b) by incorporating the preference for engineered controls over 

administrative controls into the design. As described in Section 4.2 of the ISA Summary 

(IIFP, 2012b), general management measures will provide further support for all of the listed 

IROFS. Chapter 3 of this SER provides additional discussion of the ISA methodology. 

The LA also describes fire protection measures that provide overall defense-in-depth protection 

of fire safety for operations. The following subsections, which discuss each fire-related accident 

scenario, evaluate all fire protection measures. The applicant’s ISA Summary (IIFP, 2012b) 

adequately identified the IROFS and features of fire protection in accordance with the 

regulations and the guidance established in NUREG-1520 (NRC, 2002). 

7.3.2.2 Fire Involving a Transport Truck or External Staging and Storage Pads 

Concerning the cylinder staging and storage pads, the applicant will administratively limit 

transient combustible loading on the pads, to ensure cylinder integrity. Additionally, as a 

defense-in-depth measure, the cylinder pads are sloped to prevent the accumulation of 

flammable or combustible liquids. 

The potential for a wild land fire to impact the site is not considered to be a credible event given 

the dry desert topography and lack of forestry and vegetation. Regardless, the applicant 

provides for vegetation control to ensure that brush is kept to a minimum. 

The applicant determined that the likelihood of a fire being initiated and the IROFS failing, 

causing a release exceeding the consequence threshold of 10 CFR 70.61(b) or 

10 CFR 70.61(c), is highly unlikely. The applicant based its analysis for the transport truck fire 

and fire on the cylinder staging and storage pads on the guidance provided in NUREG-1805, 

“Fire Dynamics Tools (FDTs

) Quantitative Fire Hazard Analysis Methods for the U.S. Nuclear 

Regulatory Commission Fire Protection Inspection Program,” issued December 2004 

(NRC, 2004), and NUREG-1520 (NRC, 2002). The NRC staff’s evaluation determined that the 

applicant’s analysis meets the acceptance criteria in Section 7.4.3.2 of NUREG-1520 

(NRC, 2002) and is, therefore, acceptable. 
7-5 

The NRC staff finds that the applicant has demonstrated that the facility will comply with the 

performance requirements of 10 CFR 70.61 in the event of a fire from a transport truck or on 

external staging and storage pads. 

7.3.2.3 Fire Involving a Cylinder Hauler 

The applicant determined that the likelihood of a fire being initiated and IROFS failing, so that a 

release exceeding the consequence threshold of 10 CFR 70.61(b) or 10 CFR 70.61(c) occurs, 

is highly unlikely. The NRC staff evaluated the applicant’s calculations and finds the results 

acceptable, given that the calculations were performed utilizing guidance previously approved 

by the NRC in “Confirmatory Calculations for Fire Protection Review of National Enrichment 

Facility Integrated Safety Analysis (ISA) Summary” (NRC, 2005). The NRC staff’s evaluation 

determined that the applicant’s analysis meets the acceptance criteria in Section 7.4.3.2 of 

NUREG-1520 (NRC, 2002) and is, therefore, acceptable. 

The NRC staff finds that the applicant demonstrated that the facility will comply with the 

performance requirements of 10 CFR 70.61 in the event of a fire on a cylinder transporter used 

to transport DUF6 cylinders to and from the DUF6 autoclave building and the cylinder 

staging/storage pads. 

7.3.2.4 Fire or Explosion Involving Underground Natural Gas and Propane Lines 

Several underground natural gas pipelines are located in the vicinity of the proposed 40-acre 

site. The applicant performed an analysis based on the guidance provided in NRC Regulatory 

Guide 1.91, Revision 1, “Evaluations of Explosions Postulated to Occur on Transportation 

Routes Near Nuclear Power Plants,” issued February 1978 (NRC, 1978). The leak or rupture of 

an underground natural gas pipeline, followed by detonation, would generate a blast pressure 

wave. The source magnitude of the blast would depend on several factors (pipe size, gas 

pressure, gas temperature, depth of the pipe beneath the ground surface). Atmospheric 

conditions (wind speed and stability class) would affect dispersion of the natural gas, which 

would also strongly influence the magnitude of the blast. The magnitude of the blast pressure 

wave generated by the blast would rapidly diminish with distance. 

The annual probability that a nearby, fossil fuel pipeline could rupture, detonate, and cause a 

1-psi blast pressure wave at a process building is less than 4×10-6. The probability of this event 

qualifies as highly unlikely, in accordance with the guidance in NUREG-1520 (NRC, 2002); 

therefore, no further analysis of this event is required. 

Based on the above, the applicant determined that the likelihood of a fire or explosion being 

initiated and causing a release exceeding the consequence threshold of 10 CFR 70.61(b) or 

10 CFR 70.61 (c) is highly unlikely. The NRC staff conducted a detailed onsite review of the 

applicant’s analysis and supporting documents (NRC, 2011). The applicant based its analysis 

on the guidance provided in Regulatory Guide 1.91 (NRC, 1978). The NRC staff’s evaluation 

determined that the applicant’s analysis meets the acceptance criteria in Section 7.4.3.2 of 

NUREG-1520 (NRC, 2002) and is, therefore, acceptable. 

The NRC staff finds that the applicant has demonstrated that the facility will comply with the 

performance requirements of 10 CFR 70.61 in the event of a fire or explosion involving 

underground natural gas and propane lines. 
7-6 

7.3.2.5 Fire in Various Areas of the DUF6 Conversion Plant 

The applicant will receive DUF6 from customers in solid form contained in 14-ton steel cylinders 

that are unloaded and temporarily stored in the full cylinder staging pad area until scheduled for 

feeding to the process. The DUF6 cylinders are placed in a containment autoclave where the 

contents are vaporized. Steam-heated autoclaves are used to heat the DUF6 cylinders to 

vaporize the material allowing it to enter the DUF6 feed system to the reactor for processing. 

DUF6 does not react with oxygen, nitrogen, carbon dioxide, or dry air; but it does react with 

water. For this reason, DUF6 is handled in leak tight containers and processing equipment. 

The DUF6 vapor is fed to a specially designed reaction vessel where it undergoes an exothermic 

reaction to produce DUF4 and anhydrous hydrogen fluoride (AHF). The DUF4 solids are 

continuously withdrawn from the reactor bottom through a special material cooling screw 

mechanism. The DUF4 is transferred to storage hoppers for use as raw material feed in the 

FEP building for producing SiF4 and BF3. 

Offgases from the reaction vessel leave the cooling screw equipment and pass through sintered 

metal and carbon-bed trap systems in series to remove entrained particulates and residual 

traces of unreacted DUF6. Particulate DUF4 that is collected in the sintered metal filter system 

is recycled back to the DUF4 powder system. The offgas flow, exiting the carbon-bed traps, is 

cooled by passing it through heat exchangers for collecting the byproduct AHF. Residual 

offgases exit the condenser equipment to a scrubbing system designed for removing trace 

quantities of fluorides. 

The AHF that liquefies in the condenser equipment is drained to storage tanks that are located 

in the AHF staging containment building. The AHF is temporarily stored and then loaded into 

tank-truck trailers inside the fluoride truck loading building for shipment to customers. 

A portion of the DUF6 conversion plant (the DUF6 autoclave building) is characterized as having 

ordinary or moderate fire hazards. Such a fire would dissipate shortly given the limited 

combustibles and sprinklers in the area. 

A second fire scenario involves the cylinder hauler vehicle used to transport DUF6 cylinders to 

and from the cylinder staging pads. The cylinder hauler remains external to the building and an 

overhead bridge crane moves the cylinder into or out of the autoclave. A fuel leak by the 

cylinder hauler or another vehicle adjacent to the autoclave area would have minimal impact as 

the vehicle would not enter the building and the fuels would be drained away from the building 

interior, given the slope of the vehicle pathways. The radiant heat from this limited fire would 

not affect the DUF6 cylinders because of their robust construction. The likelihood of the radiant 

heat from this limited fire affecting the DUF6 cylinders, especially given their robust construction, 

qualifies as highly unlikely in accordance with the guidance in NUREG-1520 (NRC, 2002); 

therefore, no further analysis of this event is required. 

Steam is the heating medium used in the autoclaves to bring the DUF6 cylinders to the proper 

temperature for release. A natural gas fired package boiler located in the utility building 

provides the steam. No open flames or steam are produced in the autoclave area. 

Hydraulic fluid is a limited combustible material used in the autoclave door opening and closing 

mechanisms. A leak in the system might spray hydraulic fluid onto an ignition source that would 

ignite the fluid. The capacities of the hydraulic system, hydraulic fluid properties, and operating 

pressures are not yet determined but are expected to be small, which is supported by analysis 
7-7 

of similar operations in the FHA. Because of limited combustibles in the area, a fire would be 

small and would be contained by the automatic sprinklers in the building. 

The DUF4 building contains only minor fire hazards, with the greatest single hazard being the 

hydrogen gas that is injected into the reaction vessel. The process hazards section of this SER 

(i.e., Section 7.3.4.1) discusses hydrogen. 

The reaction vessel shell is enclosed in a series of four clam-shell-type electric heaters within an 

insulated jacket. Ducting for cooling air penetrates the insulation jacket in selected locations. 

The heaters are controlled through program logic controllers to provide heating for startup. After 

the reaction vessel has reached its operating temperature, the heaters are used in conjunction 

with a cooling air blower and ducting for fine control of reaction vessel temperatures for optimum 

performance. The reaction vessel shell heaters are not considered to be a major ignition source 

for the area. Combustibles are limited, and no hydraulic fluids are located in this area. 

A vacuum system will transfer the finished product from the DUF4 process area to the FEP 

process. A failure in the vacuum system and a system rupture could allow material to fall freely 

in the area. A fire in the area from other sources could possibly be increased as a result of the 

loose particles of DUF4 (although DUF4 is not combustible) in the area, but the automatic 

sprinklers in the building would contain such a fire. DUF4, upon exposure to water, can very 

slowly release some hydrogen fluoride (HF) by hydrolysis. Post fire cleanup of the water 

minimizes the likelihood of any reactions, given the very slow reaction time, and no additional 

significant fire consequences are created. 

Granular-sized carbon is used in the carbon-bed traps for adsorption of trace of amounts of 

DUF6 from the reaction vessel offgas stream. The carbon is enclosed in six separate trap 

housings. Each trap contains about 0.34–0.42 cubic m (12–15 cubic ft) of the granular carbon 

material. The carbon trap housing (shell) is of robust mechanical integrity, meets the 

requirements of the American Society of Mechanical Engineers (ASME) Code, and operates at 

relatively low pressures (about 16–25 pounds-force per square inch absolute). The likelihood of 

rupturing the trap housing is low. When a trap requires replacement, it is isolated with valves 

and is moved to the decontamination building for replacement of the carbon material inside an 

enclosed hooded booth system. Carbon material is not changed inside the DUF4 process 

building’s process area. The potential of spillage of any particulate carbon inside a process 

building that would result in a significant source of combustible material is very low. 

Based on the above, the applicant determined that the likelihood of a fire being initiated and the 

IROFS failing, so that a release exceeding the consequence threshold of 10 CFR 70.61(b) or 

10 CFR 70.61(c) occurs, is highly unlikely. The applicant performed this analysis using the 

guidance provided in NUREG-1520 (NRC, 2002). In addition, the analysis meets the 

requirements of NFPA 801 (NFPA, 2008c). The NRC staff’s evaluation determined that the 

applicant’s analysis meets the acceptance criteria in Section 7.4.3.2 of NUREG-1520 

(NRC, 2002) and is, therefore, acceptable. 

The NRC staff finds that the applicant has demonstrated that the facility will comply with the 

performance requirements of 10 CFR 70.61 in the event of a fire in the DUF6 conversion plant. 

7.3.2.6 Fire in Various Areas of the SiF4 Plant 

Silicon dioxide (SiO2) powder is mixed with the DUF4 and continuously fed to an electrically 

heated rotary calciner in which the reaction temperature is about 480–650 degrees C (900–
7-8 

1,200 degrees F). Solid particles of depleted uranium dioxide (DUO2) discharge from the rotary 

calciner via a cooling screw mechanism and are cooled and sent to the storage hopper. Alarms 

and controls are in place to shut down the rotary calciner and cooling screw in the event of a 

loss of cooling on the transfer screw. The product (SiF4) offgas stream is noncombustible. 

Stored and packaged product is also noncombustible. 

DUO2 is pyrophoric, but only at elevated temperatures, specific particle sizes, and with 

adequate oxygen to support combustion. Within the storage hopper, a nitrogen gas buffer is 

maintained as a defense-in-depth measure. A spill or release of depleted uranium oxide could 

potentially be an issue. Exposed DUO2 does not spontaneously combust; however, if the 

material is exposed to an existing fire, it increases the amount of fuel loading in the area. The 

specific heat released is not anticipated to be of sufficient quantity to create structural damage. 

The automatic sprinklers in the area would prevent other combustibles from igniting; however, 

the reaction from the introduction of water could further distribute burning material in the 

building. Given the limited combustibles, any spread of the fire would be highly unlikely to 

create a scenario in which the consequence thresholds of 10 CFR 70.61 would be exceeded. 

Based on the above, the applicant determined that the likelihood of a fire being initiated and of 

the IROFS failing, so that a release exceeding the consequence threshold of 10 CFR 70.61(b) 

or 10 CFR 70.61(c) occurs, is highly unlikely. The applicant performed this analysis using the 

guidance provided in NUREG-1520 (NRC, 2002). In addition, the analysis meets the 

requirements of NFPA 801 (NFPA, 2008c). The NRC staff’s evaluation determined that the 

applicant’s analysis meets the acceptance criteria in Section 7.4.3.2 of NUREG-1520 

(NRC, 2002) and is, therefore, acceptable. 

The NRC staff finds that the applicant has demonstrated that the facility will comply with the 

performance requirements of 10 CFR 70.61 in the event of a fire in the SiF4 plant. 

7.3.2.7 Fire in Various Areas of the BF3 Plant 

Diboron trioxide (B2O3) is mixed with DUF4 and continuously fed to an electrically heated 

preheater followed by reaction of the mixture in an electrically heated rotary calciner in which 

the reaction temperature is about 480–650 degrees C (900-1,200 degrees F). Product (BF3) 

offgas streams and stored product are noncombustible. Solid particles of DUO2 discharge from 

the rotary calciner via a cooling screw mechanism and are cooled and sent to the storage 

hopper. Alarms and controls are in place to shut down the rotary calciner and cooling screw in 

the event of a loss of cooling on the screw. 

DUO2 is pyrophoric, but only at elevated temperatures, specific particle sizes, and with 

adequate oxygen to support combustion. Within the storage hopper, a nitrogen gas buffer is 

maintained as a defense-in-depth measure. A spill or release of depleted uranium oxide could 

potentially be an issue. Exposed DUO2 does not spontaneously combust; however, if the 

material is exposed to an existing fire, it increases the amount of fuel loading in the area. The 

specific heat released is not anticipated to be of sufficient quantity to create structural damage. 

The automatic sprinklers in the area would prevent other combustibles from igniting; however, 

the reaction from the introduction of water could further distribute burning material in the 

building. Given the limited combustibles, any spread of the fire would be highly unlikely to 

create a scenario in which the consequence thresholds of 10 CFR 70.61 would be exceeded. 

Based on the above, the applicant determined that the likelihood of a fire being initiated and of 

the IROFS failing, so that a release exceeding the consequence threshold of 10 CFR 70.61(b) 
7-9 

or 10 CFR 70.61(c), occurs is highly unlikely. The applicant performed this analysis using the 

guidance provided in NUREG-1520 (NRC, 2002). In addition, the analysis meets the 

requirements of NFPA 801 (NFPA, 2008c). The NRC staff’s evaluation determined that the 

applicant’s analysis meets the acceptance criteria in Section 7.4.3.2 of NUREG-1520 

(NRC, 2002) and is, therefore, acceptable. 

The NRC staff finds that the applicant has demonstrated that the facility will comply with the 

performance requirements of 10 CFR 70.61 in the event of a fire in the BF3 plant. 

7.3.2.8 Fire Hazards Analysis Conclusions 

The NRC staff has reasonable assurance that the applicant has identified and evaluated all firerelated accident scenarios credible for the proposed extraction and deconversion processes. 

The applicant has reasonably identified and evaluated possible fire initiators and consequences 

and has identified IROFS for preventing or mitigating fire accident scenarios that could result in 

intermediate or high consequences, in accordance with 10 CFR Part 70, “Domestic Licensing of 

Special Nuclear Material,” and as described by the guidelines established in NUREG-1520 

(NRC, 2002). 

Consistent with the acceptance criteria in Section 7.4.3.2 of NUREG-1520 (NRC, 2002), the 

applicant performed an FHA for each process area. The FHA described, by fire area, the fuel 

loading, fire scenarios, methods of consequence analysis, potential consequences, and 

mitigative controls, which meets the requirements of NFPA 801 (NFPA 2008c) and is, therefore, 

acceptable. In addition, based on its review of the above information, the NRC staff concludes 

with reasonable assurance that the applicant's performance of its FHA meets the requirements 

of 10 CFR 40.32, 10 CFR 70.61, and 10 CFR 70.64, as they pertain to the fire protection 

aspects of the facility. 

7.3.3 Facility Design 

7.3.3.1 Facility Passive-Engineered Fire Protection Systems 

Buildings containing uranium are within the DUF6 conversion plant (DUF6 autoclave building, 

DUF4 process building, decontamination building, DUF4 container storage building, and the 

DUF4 container staging building) and within the SiF4/BF3 plants (FEP building and FEP oxide 

staging building), which have noncombustible structural steel beams and columns, on concrete 

slabs, with built-up composite roofing on metal deck. This construction is equivalent to 

Type II-000, in accordance with NFPA 220, “Standard for Types of Building Construction” 

(NFPA, 2006b). Thus, all structural members are of noncombustible or limited combustible 

construction, but no fire rating is required. To meet building code requirements and as a 

defense-in-depth measure, some buildings are separated into different fire areas by rated walls 

constructed in accordance with NFPA 221, “Standard for High Challenge Fire Walls, Fire Walls, 

and Fire Barrier Walls” (NFPA, 2006c). 

To ensure life safety protection for the occupants, buildings are provided with means of egress, 

illumination, and protection in accordance with the New Mexico Commercial Building Code 

(NMCBC) (NMCBC, 2006). Barriers with fire resistance ratings consistent with the NMCBC and 

the FHA are provided to prevent fire propagation. All of the buildings are provided with 

emergency lighting for the illumination of the primary exit paths and the essential operations 

areas where personnel are required to operate valves, dampers, and other controls in an 

emergency. Emergency lighting is considered a critical load. Marking of means of egress, 
7-10 

including illuminated exit signs with battery backup, are provided in accordance with the 

NMCBC and NFPA 101, “Life Safety Code” (NFPA, 2006a). 

The potential for lightning strikes to the buildings is considered possible; however, the structural 

design with metal beam and columns and connections to the underground grounding loops 

surrounding the buildings is permissible in lieu of air terminals under NFPA 780, “Standard for 

the Installation of Lightning Protection Systems” (NFPA, 2008b), and is considered to be 

effective lightning protection. 

The applicant’s ISA Summary (IIFP, 2012b) has adequately addressed passive-engineered fire 

protection systems in accordance with the guidance established in NUREG-1520 (NRC, 2002) 

and meets the requirements of NFPA 801 (NFPA, 2008c). In addition, the applicant’s 

passive-engineered fire protection systems meet the requirements of 10 CFR 40.32, 

10 CFR 70.64, and 10 CFR 70.65 as they pertain to the fire protection aspects of the facility. 

7.3.3.2 Facility Active Engineered Fire Protection Systems 

7.3.3.2.1 Electrical Installation, Ventilation, and Fire Alarm System 

All electrical systems at the facility are installed in accordance with NFPA 70, “National Electric 

Code” (NFPA, 2008a). Switchgear, motor control centers, panel boards, variable frequency 

drives, uninterruptible power supply (UPS) systems, and control panels are mounted in metallic 

enclosures and contain limited amounts of combustible material. Cable trays and conduits are 

metallic, and the cable in cable trays is flame retardant and tested in accordance with industry 

guidance found in Institute of Electrical and Electronics Engineers (IEEE) Standard 383, 

“Standard for Type Test of Class 1E Electric Cables, Field Splices, and Connections for Nuclear 

Power Generating Stations” (IEEE, 2003). 

Roof-mounted exhaust fans and wall-mounted intake louvers provide ventilation for the DUF4

and FEP process buildings. Steam is used as the main heat source for the process building 

environment. Process control room areas are heated, ventilated, and cooled by electrical heat 

pump units with electric auxiliary heat. The control room heating, ventilation, and 

air-conditioning (HVAC) units create positive pressure in each of the control rooms with alarms 

to indicate loss of pressure. Process equipment areas are open and of large volumes, so steam 

heating is practical. Cooling of other process and storage areas is provided with wall-mounted 

exhaust fans and intake louvers. The ventilation and HVAC systems meet NFPA 90A, 

“Installation of Air Conditioning and Ventilating Systems” (NFPA, 2009b), and NFPA 90B, 

“Installation of Warm Air Heating and Air Conditioning Systems” (NFPA, 2009c). The ventilation 

systems are not engineered for smoke control, but are designed to shut down in the event of a 

fire. The offsite fire department provides smoke control using portable smoke removal 

equipment. 

7.3.3.2.2 Portable Fire Extinguishers 

Portable fire extinguishers are installed and inspected throughout all buildings in accordance 

with NFPA 10, “Standard for Portable Fire Extinguishers” (NFPA, 2007a). Multipurpose fire 

extinguishers are provided generally for Class A (ordinary combustibles), Class B (flammable 

and combustible liquids), and Class C (electrical equipment) fires. The NRC staff finds 

acceptable the applicant’s commitment to provide portable fire extinguishers appropriate for the 

type of fire in accordance with national fire standards. 
7-11 

7.3.3.2.3 Fire Water Supply and Drainage 

The facility fire water supply consists of two 378,541-L (100,000-gal) water storage tanks 

designed and constructed in accordance with NFPA 22, “Standard for Water Tanks for Private 

Fire Protection” (NFPA, 2003). Separate storage tanks are used for the sanitary water supply. 

The site’s well water supply provides fill and make up to the tanks, and the water supply is 

capable of filling fire protection water inventory in a single tank within an 8-hour period. The fire 

pumps consist of one electric-driven pump and one diesel-driven pump, both rated for 

2,271 liters per minute (lpm) (600 gallons per minute [gpm]), at 6.89-bar (100-pounds-force per 

square inch absolute) pumps. Both pumps are horizontal centrifugal pumps designed and 

installed in accordance with NFPA 20, “Standard for Installation of Stationary Pumps for Fire 

Protection” (NFPA, 2007c). The combination of two water tanks and two fire pumps provides 

100-percent redundancy for fire protection. The tanks are arranged such that one will be 

available for suction at all times. In addition to fixed standpipes and fire hose stations, the 

facility will be provided with fire hose on mobile apparatus or at strategic locations throughout 

the facility. The amount of hose provided will be sufficient to ensure that all points within the 

facility will be able to be reached by at least two backup hoses with a diameter of 64 millimeters 

(mm) (2.5 inches [in]), consistent with NFPA 1410, “Standard on Training for Initial Emergency 

Scene Operations” (NFPA, 2010b). These lines will have a minimum nozzle pressure of 4.5 bar 

(65 pounds-force per square inch gauge [psig]) for the fixed hose and 6.9 bar (100 psig) for the 

64-mm (2.5-in) hose. The NRC finds acceptable the applicant’s commitment to provide a fire 

water supply system consistent with national standards, which provides adequate redundancy 

for water supply to all points within the facility. 

Buildings, building aprons, and process area outdoor pads, where chemicals or licensed 

materials are stored or processed, have curbs or dikes—or both—to prevent drainage of 

contaminated liquids outside the spill controlled areas. Water from activation of the sprinkler 

system or from firefighting activities could contain contaminated materials or flammable and 

combustible liquids. During the initial period of sprinkler activation or generation of fire water in 

an area, the water collects in the spill controlled area and is handled and treated as any other 

type of spillage or liquid. Areas that have dikes for nonuranium hazardous chemical or oil spill 

control have installed pumps that can either automatically or manually be activated to pump 

spilled liquids or water for treatment. Areas where licensed materials are processed or stored, 

and have curbing or dikes, are not automatically pumped. If fire water accumulates in those 

areas in excess of the holding capacity, the water may be pumped either to another licensed 

material curb or dike area, or to holding tanks in the decontamination building, or to the large HF 

recycle tank where it can be sampled and its disposition determined. Portable pumps are also 

available for emergency pumping of liquids to other holding areas, tanks, or treatment, if 

necessary. If the volume of the fire water reaches a level that exceeds the respective spill 

control area, it is pumped to other outside spill control areas not directly affected by the fire 

response. If the water drains and enters the plant storm sewer drain system, it then flows to the 

storm water retention basin where it is sampled and a decision is made about its disposal. The 

NRC finds acceptable the applicant’s drainage system and back-up pumping capability to 

control contamination from fire suppression liquids. 

7.3.3.2.4 Engineered Automatic Fire Suppression Systems 

Automatic wet pipe sprinkler systems are provided in all buildings in the facility, with the 

exception of the lime silo storage shed, which is not normally occupied, has limited amounts of 

combustibles, and contains no licensed material. 
7-12 

These systems are designed and tested in accordance with NFPA 13, “Standard for the 

Installation of Sprinkler Systems” (NFPA, 2007b). Sprinkler system control valves are 

monitored under a periodic inspection program, and their proper positioning is supervised in 

accordance with NFPA 801 (NFPA, 2008c). 

The applicant’s ISA Summary (IIFP, 2012b) adequately addressed active-engineered fire 

protection systems in accordance with the guidance established in NUREG-1520 (NRC, 2002) 

and meets the requirements of NFPA 801 (NFPA, 2008c). In addition, the applicant’s 

active-engineered fire protection system meets the requirements of 10 CFR 40.32, 

10 CFR 70.64, and 10 CFR 70.65 as they pertain to the fire protection aspects of the facility. 

7.3.3.3 Facility Design Conclusions 

The NRC staff finds the applicant adequately addressed building construction, fire area 

determination, electrical installation, life safety, drainage, and lightning protection in the 

application. The FHA described ventilation characteristics as they relate to fire protection and 

fire hazards. 

Consistent with the acceptance criteria in Section 7.4.3.3 of NUREG-1520 (NRC, 2002), the 

application documents the fire safety considerations used in the general design of the facilities 

containing licensed material or facilities that have the potential to impact the safety of 

radiological facilities and is, therefore, acceptable. In addition, the NRC staff’s evaluation finds 

that the fire protection features of the applicant’s proposed facility meet the requirements of 

10 CFR 40.32, 10 CFR 70.61, 10 CFR 70.64, and 10 CFR 70.65 as they pertain to the fire 

protection aspects of the facility. 

7.3.4 Process Fire Safety 

The applicant plans to convert (deconversion) DUF6 into uranium oxide for long-term, stable 

disposal. The company will also include a commercial plant to produce specialty fluoride gas 

products for sale. Using the fluorine derived from the deconversion process, the applicant will 

manufacture high-purity SiF4 and BF3 in the IIFP facility. The fluoride gas products are utilized 

in the electronic, solar, and semiconductor markets. In addition, AHF is a product of the 

deconversion process and is sold as a chemical for use in various industrial applications. 

The applicant’s FHA, which is discussed in Section 7.2.2 of this SER, analyzed many of the 

specific hazards within the process. The following subsection discusses the aspects of each 

process chemical that has fire safety implications. 

7.3.4.1 Process Descriptions 

Depleted Uranium Hexafluoride (DUF6): DUF6 is not flammable and does not disassociate to 

flammable constituents under conditions at which it will be handled at the facility. DUF6 does not 

react with oxygen, nitrogen, carbon dioxide, or dry air; but does react with water or water vapor. 

Hydrocarbons can be explosively oxidized if they are mixed with DUF6 in the liquid state or at 

elevated temperatures. For this reason, the DUF6 processes at the facility do not use 

nonfluorinated hydrocarbon lubricants. DUF6 pumps are lubricated using a perfluoropolyether 

(PFPE) oil that is noncombustible. The used PFPE oil will be collected, packaged, and shipped 

offsite for disposal at a licensed low-level radioactive waste facility. 
7-13 

Hydrogen Fluoride (HF): HF is a byproduct of the chemical reaction of DUF6 with water vapor. 

HF is extremely reactive in both gaseous and aqueous form. HF alone is neither flammable nor 

combustible. It can, however, react exothermically with water to generate sufficient heat to 

ignite nearby combustibles. 

Depleted Uranyl Fluoride (DUO2F2): DUO2F2 is also a byproduct of the chemical reaction of 

DUF6 with water vapor. DUO2F2 is stable in air to 300 degrees C (572 degrees F). It is neither 

flammable nor combustible and will not decompose to combustible constituents under 

conditions that will exist at the facility. 

Depleted Uranium Dioxide (DUO2): DUO2 is pyrophoric, but only at elevated temperatures, 

specific particle sizes, and with adequate oxygen to support combustion. Within the process, a 

nitrogen gas buffer is maintained as a defense-in-depth measure. A spill or release of DUO2

could potentially be an issue. Exposed DUO2 does not spontaneously combust; however, if the 

material is exposed to an existing fire, it increases the amount of fuel loading in the area. The 

specific heat released is not anticipated to be of sufficient quantity to create structural damage 

to any of the various process buildings. The automatic sprinklers in the process areas would 

prevent other combustibles from igniting; however, the reaction from the introduction of water 

could further distribute burning material in the building. Given the limited combustibles in the 

process areas, the spread of the fire would become highly unlikely. 

Storage and Handling of DUF6: DUF6 cylinders are stored or handled in the cylinder 

staging/storage pads and in the DUF6 autoclave building. On the staging/storage pads, fire 

concerns include the cylinder transport vehicle, a fire exposure from nearby vegetation, and fire 

exposure from a nearby vehicle accident. The applicant evaluated these various fire scenarios 

and either concluded that they did not pose a threat to the stored cylinders or that, with 

adequate controls, the threats could be mitigated by IROFS. Combustible loadings in the DUF6 

autoclave building are limited, and transient combustibles will be controlled. Therefore, any fire 

originating in these areas will be limited. 

Hydrogen Control: Hydrogen is produced for use in the DUF4 plant and may also be generated 

at battery charging stations in the facility (all forklifts are battery powered). The gaseous 

hydrogen supply for the various processes is generated onsite using a vendor-supplied, 

pre-engineered system which reforms natural gas followed by purification using pressure swing 

absorption. 

Hydrogen control in battery-charging stations will be provided by measures identified in 

NFPA 70E, “Standard for Electrical Safety in the Workplace” (NFPA, 2009a), including hydrogen 

detection. Natural and mechanical ventilation will be provided to ensure that hydrogen 

concentrations do not exceed 25 percent of the lower explosive limit. Redundant ventilating 

fans with explosive proof motors are run continuously or are interlocked to start upon the 

detection of hydrogen in the area. Redundant airflow sensors will indicate fan run status. 

Combustible Material Hazards: Materials of construction for the process buildings are 

predominantly noncombustible (e.g., steel, aluminum, and concrete floors). A minimum of fixed 

combustibles is expected to be present in the operations areas, and the applicant plans to 

control transient combustibles to minimize potential fire hazards, and general housekeeping 

procedures. Other quantities of combustible materials are as follows: 

• Minor quantities of hydraulic fluids or lube oils are used in the overhead cranes and 

small forklifts. 
7-14 

• Combustible coke is used in the coke box scrubber; however, it is kept wet as part of the 

process and is highly unlikely to become sufficiently dry to become a com'bustible 

hazard. 

• Natural gas is the primary fuel for the steam boilers. 

• Diesel fuel supplies the backup generators and the diesel powered fire pump. In 

addition, quantities of diesel fuel are onsite at the remote, dedicated fueling station for 

the cylinder hauler. 

• Granular-sized carbon is used in the carbon-bed traps in various process areas. 

The NRC staff finds the applicant’s list of combustible materials accurate for the described 

operations and finds acceptable the applicant’s commitments discussed above to limit 

combustible loading. 

7.3.4.2 Process Fire Safety Conclusions 

No fire safety IROFS are directly inherent to the process design. Section 7.2.2 of this SER 

discusses the fire safety IROFS used to protect the licensed material within the processes. In 

its review, the NRC staff has taken into account the potential presence of the identified 

combustibles in the various accident scenarios. The applicant’s ISA Summary (IIFP, 2012b), 

which provides the supporting safety basis for the LA (IIFP, 2012a), documents the identification 

of fire hazards and related analyses. 

Consistent with the acceptance criteria in Section 7.4.3.4 of NUREG-1520 (NRC, 2002), the 

application identifies the hazardous chemicals, processes, and design standards used to ensure 

safety in areas that have fire hazards that may threaten licensed material and is, therefore, 

acceptable. In addition, the NRC staff’s evaluation finds that the fire protection features of the 

applicant's facility meet the requirements of 10 CFR 40.32, 10 CFR 70.61, 10 CFR 70.64, and 

10 CFR 70.65 as they pertain to the process fire safety aspects of the facility. 

7.3.5 Fire and Emergency Response 

The facility will maintain a fire brigade made up of employees trained in firefighting techniques, 

first-aid procedures, and emergency response. The fire brigade is organized, operated, trained, 

and equipped in accordance with NFPA 600, “Standard on Industrial Fire Brigades” 

(NFPA, 2010a) for incipient firefighting capability. The intent of the facility fire brigade is to be 

able to handle all minor fires and to provide a first response effort designed to supplement the 

local fire department for major fires at the plant. The field incident commander, working with the 

plant’s emergency director, will coordinate offsite fire department activities. The fire brigade is 

staffed so as to ensure a minimum of five brigade members available per shift. Section 7.3.3.2 

of this SER describes the onsite water supply available to the fire brigade and other responders. 

The applicant provides periodic training on the facility’s emergency training procedures to 

personnel of offsite organizations. Facility emergency response personnel meet at least 

biennially with each offsite assistance group to accomplish training and review items of mutual 

interest, including relevant changes to the program. The primary agencies that will be available 

for this response are the fire and rescue agency of the City of Eunice, NM, and fire department 
7-15 

of the City of Hobbs, NM. These agencies are signatories to the Lea County, NM, mutual aid 

agreement and can request assistance from any of several adjacent municipal fire departments 

and the fire/emergency response services. The applicant has received letters from the 

referenced fire departments, which include a commitment to fire protection and emergency drills 

and define the fire protection and emergency response commitments between the 

organizations. 

The Hobbs Fire Department is the primary response agency, and is comprised of a roster of 

approximately 70 paid personnel, staffing three fire stations in a three-shift rotation. The 

department has structural engines, ladder truck, heavy rescue truck, grass fire trucks, a water 

tanker, and several command vehicles and ambulances, each equipped to provide advanced 

level life support. Firefighters are trained to Firefighter Level I and Emergency Medical 

Technician (EMT) Basic as a minimum, in accordance with New Mexico standards. 

Shift-assigned ambulance personnel are EMT paramedics under New Mexico standards. 

Eunice Fire and Rescue is the secondary response agency and comprises a roster of 

approximately 20 volunteers. Eunice has structural fire engines, grass fire trucks, a water 

tanker, command vehicles, and ambulances, each equipped to provide intermediate level life 

support. Firefighters are trained, as a minimum, to Firefighter Level I and ambulance personnel, 

as a minimum, to EMT Basic, in accordance with New Mexico standards. In the event of a fire, 

the IIFP fire brigade responds and the Hobbs and Eunice Fire and Rescue Departments are 

notified to respond. If the fire is incipient, the IIFP fire brigade fights the fire using hand 

portable/wheeled fire extinguishers and 38-mm (1.5 in) hose lines. The estimated response 

time to the facility is less than 15 minutes for the Hobbs Fire Department and between 20 and 

30 minutes from Eunice Fire and Rescue. 

The NRC staff finds that the onsite fire brigade, onsite water supply, onsite hose lines, and 

mutual aid from adequately equipped fire departments can provide defense-in-depth protection 

from releases from all identified and credible fire scenarios, satisfy the requirements of 

10 CFR 70.64(b), and are in accordance with the guidance in NUREG-1520 (NRC, 2002). 

Consistent with the acceptance criteria in Section 7.4.3.5 of NUREG-1520 (NRC, 2002), the LA 

(IIFP, 2012a) documents the available fire protection systems and fire emergency response 

organizations and is, therefore, acceptable. In addition, the applicant’s emergency response 

capability meets the requirements of 10 CFR 40.32, 10 CFR 70.61, 10 CFR 70.64, and 

10 CFR 70.65, as they pertain to the fire protection aspects of the facility. 

7.3.6 Baseline Design Criteria 

The NRC staff reviewed the Fluorine Extraction Process and Depleted Uranium Deconversion 

Plant and its fire safety program to determine applicability and level of compliance with 

NFPA 801 (NFPA, 2008c) and applicable standards referenced within. Table 7-1 of the LA 

(IIFP, 2012a) lists the fire codes and standards considered by the applicant to be applicable to 

the facility. The NRC staff finds the use of these consensus codes and standards to be in 

accordance with the guidance of Section 7.4.3 of NUREG-1520 (NRC, 2002) with regard to 

nationally recognized codes and standards that may be used to measure reasonable assurance 

of fire safety. The design provides for adequate protection against fires and explosions. 

Therefore, the NRC staff considers that the use of the above codes and standards satisfies the 

requirements of 10 CFR 70.64(a) and 10 CFR 70.64(a)(3). 
7-16 

Consistent with the acceptance criteria in Section 7.4.3 of NUREG-1520 (NRC, 2002), the 

LA (IIFP, 2012a) and ISA Summary (IIFP, 2012b) document the defense-in-depth provided to 

enhance safety by reducing the challenges to IROFS. This chapter of the SER documents the 

evaluation of the notable fire protection related defense-in-depth measures. Furthermore, the 

IROFS outlined in Section 7.2.2.1 of this SER show a clear preference for engineered controls 

over administrative controls. Administrative controls were typically used to supplement 

engineered controls or when the use of an engineered control was not practicable. Therefore, 

given the outlined methodology, the NRC staff finds that the facility’s design satisfies the 

requirements of 10 CFR 70.64(b), as they pertain to the fire protection aspects of the facility. 

7.4 Evaluation Findings

The applicant has established a fire protection function meeting the acceptance criteria from 

Chapter 7 of NUREG-1520 (NRC, 2002). The function includes a facility safety review 

committee responsible for integrating modifications to the facility and a fire safety manager 

responsible for day-to-day program implementation. Fire prevention, inspection, testing, and 

maintenance of fire protection systems and the qualification, drills, and training of facility 

personnel are in accordance with applicable NFPA codes and standards. (Note that SER 

Section 11.3.3 describes fire protection training requirements.) 

The applicant has conducted risk analyses in accordance with NFPA 801, “Standard for 

Fire Protection for Facilities Handling Radioactive Material” (NFPA, 2008c). The FHA identified 

credible fire scenarios that bound the fire risk. The ISA used these scenarios and identified fire 

protection IROFS. A memorandum of understanding with the local fire department documents 

the required assistance and the annual exercises. Procedures are in place to allow the fire 

department efficient access to process areas during fire emergencies. Worker egress is 

designed and maintained in accordance with NMCBC (NMCBC, 2006) and NFPA 101 

(NFPA, 2006a). 

The applicant demonstrated that it incorporated appropriate fire safety considerations in the 

design of its facilities. The applicant also demonstrated that the facility has appropriate active 

fire protection systems. 

The NRC staff concludes that the applicant’s capabilities meet the criteria in Chapter 7 of the 

SRP. In addition, the NRC staff concludes that the applicant provided sufficient information in 

accordance with requirements of 10 CFR 40.32, and 10 CFR 70, Subpart H, regarding potential 

fire hazards, consequences, and required controls for the proposed processes. The NRC staff 

finds that the applicant’s proposed equipment, facilities, and procedures provide assurance that 

adequate fire protection will be provided and maintained for those IROFS to meet the safety 

performance requirements and BDC of 10 CFR Part 70. 

7.5 References

(IEEE, 2003) Institute of Electrical and Electronics Engineers, Inc., “Standard for Type Test of 

Class 1E Electric Cables, Field Splices, and Connections for Nuclear Power Generating 

Stations,” IEEE Standard 383, December 12, 2003. 

(IIFP, 2012a) International Isotopes Fluorine Products, Inc., “Fluorine Extraction Process and 

Depleted Uranium Deconversion Plant (FEP/DUP) License Application, Revision B,” May 2012, 

Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML12123A245. 
7-17 

(IIFP, 2012b) International Isotopes Fluorine Products, Inc., “ISA Summary Rev. B for IIFP,” 

May 2012, Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML12123A245. 

(NFPA, 2010a) National Fire Protection Association, “Standard on Industrial Fire Brigades,” 

NFPA 600, 2010. 

(NFPA, 2010b) National Fire Protection Association, “Standard on Training for Initial 

Emergency Scene Operations,” NFPA 1410, 2010. 

(NFPA, 2009a) National Fire Protection Association, “Standard for Electrical Safety in the 

Workplace,” NFPA 70E, 2009. 

(NFPA, 2009b) National Fire Protection Association, “Installation of Air Conditioning and 

Ventilating Systems,” NFPA 90A, 2009. 

(NFPA, 2009c) National Fire Protection Association, “Installation of Warm Air Heating and Air 

Conditioning Systems,” NFPA 90B, 2009. 

(NFPA, 2008a) National Fire Protection Association, “National Electric Code,” NFPA 70, 2008. 

(NFPA, 2008b) National Fire Protection Association, “Standard for the Installation of Lightning 

Protection Systems,” NFPA 780, 2008. 

(NFPA, 2008c) National Fire Protection Association, “Standard for Fire Protection for Facilities 

Handling Radioactive Materials,” NFPA 801, 2008. 

(NFPA, 2007a) National Fire Protection Association, “Standard for Portable Fire Extinguishers,” 

NFPA 10, 2007. 

(NFPA, 2007b) National Fire Protection Association, “Standard for the Installation of Sprinkler 

Systems,” NFPA 13, 2007. 

(NFPA, 2007c) National Fire Protection Association, “Standard for Installation of Stationary 

Pumps for Fire Protection,” NFPA 20, 2007. 

(NFPA, 2007d) National Fire Protection Association, “National Fire Alarm Code,” NFPA 72, 

2007. 

(NFPA, 2006a) National Fire Protection Association, “Life Safety Code,” NFPA 101, 2006. 

(NFPA, 2006b) National Fire Protection Association, “Standard on Types of Building 

Construction,” NFPA 220, 2006. 

(NFPA, 2006c) National Fire Protection Association, “Standard for High Challenge Fire Walls, 

Fire Walls, and Fire Barrier Walls,” NFPA 221, 2006. 

(NFPA, 2003) National Fire Protection Association “Standard for Water Tanks for Private Fire 

Protection,” NFPA 22, 2003. 
7-18 

(NMCBC, 2006) New Mexico Commercial Building Code, 2006. 

(NRC, 2011) U.S. Nuclear Regulatory Commission, “January 19, 2011, Meeting Summary, 

‘Overview of the Fossil Fuel Pipeline Explosion Evaluation for the International Isotopes, Inc., 

FEP & DUF6 Deconversion Facility.’” 

(NRC, 2005) U.S. Nuclear Regulatory Commission, “Confirmatory Calculations for Fire 

Protection Review of National Enrichment Facility Integrated Safety Analysis (ISA) Summary,” 

March 2005. Agencywide Documents Access and Management System (ADAMS) Accession 

No. ML050910020 

(NRC, 2004) U.S. Nuclear Regulatory Commission, “Fire Dynamics Tools (FDTs

) Quantitative 

Fire Hazard Analysis Methods for the U.S. Nuclear Regulatory Commission Fire Protection 

Inspection Program,” NUREG-1805, December 2004. 

(NRC, 2002) U.S. Nuclear Regulatory Commission, “Standard Review Plan for the Review of a 

License Application for a Fuel Cycle Facility,” NUREG-1520, Rev. 0, March, 2002. 

(NRC, 1978) U.S. Nuclear Regulatory Commission, “Evaluations of Explosions Postulated To 

Occur on Transportation Routes Near Nuclear Power Plants,” Regulatory Guide 1.91, Rev. 1, 

February 1978. 
8-1 

8.0 EMERGENCY MANAGEMENT 

The purpose of reviewing the International Isotopes Fluorine Products, Inc.’s (IIFP or the 

applicant), Emergency Plan (EP) (IIFP, 2012a) is to determine whether IIFP has established 

adequate emergency response facilities and procedures to protect workers, the public, and the 

environment. 

8.1 Regulatory Requirements

The regulations in Title 10 of the Code of Federal Regulations (10 CFR) 40.31(j)(1) require 

applicants that are requesting authorization to possess uranium hexafluoride (DUF6) in excess 

of 50 kilograms (kg) (110 pounds [lb]) in a single container or 1,000 kg (2,200 lb) total to submit 

either (1) an evaluation that shows that the maximum intake of uranium by a member of the 

public from a release would not exceed 2 milligrams (mg) or (2) an EP which addresses the 

radiological hazards of an accidental release of source material and any associated chemical 

hazards directly incident thereto. The regulations in 10 CFR 40.31(j)(3) also outline the 

regulatory requirements for the information that must be included in an EP. The regulations in 

70.64(a)6) require emergency planning to be incorporated into the baseline design criteria. 

8.2 Regulatory Acceptance Criteria

Title 10 CFR 40.31(j)(3) and Section 8.4.3 of NUREG-1520, “Standard Review Plan for the 

Review of a License Application for a Fuel Cycle Facility” (NRC, 2002), outline the acceptance 

criteria for the U.S. Nuclear Regulatory Commission’s (NRC’s) review of the EP. 

8.3 Staff Review and Analysis

8.3.1 Facility Description 

Section 2.0 of the EP (IIFP, 2012a) describes the licensed activity, the facility, the site, and the 

area near the site. The information provided includes the following: 

• a brief description of the deconversion process 

• a discussion of chemicals of concern—including form, physical state, location, and 

quantity 

• a detailed description of the site location and layout 

• a description of the major structures to be located at the site 

• a description of the ventilation systems—including stack heights, maximum flow rates, 

and filter efficiencies 

• a description of the area near the site, including area land use information 

• demography of the area, water use, and climate 

• detailed maps of the facility and surrounding area. 
8-2 

The NRC staff finds that the information provided by the applicant as discussed above, taken 

together with the information provided by the applicant as discussed in Section 8.3.2 of this 

SER, below, meets the acceptance criteria of 10 CFR 40.31(j)(3)(i) and Section 8.4.3.1.1 of 

NUREG-1520 (NRC, 2002). Therefore, the NRC staff finds the applicant’s facility description 

adequate. 

8.3.2 Onsite and Offsite Emergency Facilities 

Section 9.0 of the EP (IIFP, 2012a) contains descriptive information regarding the emergency 

response equipment and facilities. The primary Emergency Operations Center (EOC) is the 

facility’s control room. The IIFP’s control room/EOC controls communications to all principal 

points within and outside the facility. The IIFP’s control room/EOC contains procedures and 

operational engineering information to assist in routine operations and in emergency response. 

The IIFP EOC and its personnel support the following functions: 

• Assess abnormal conditions. 

• Notify additional facility personnel, if needed. 

• Make offsite notifications. 

• Perform or direct accident mitigation. 

• Direct facility operations. 

• Implement onsite protective actions. 

If the nature and location of the emergency situation requires it, or if the control room/EOC 

becomes uninhabitable, the Emergency Director may move the emergency response personnel 

to the designated backup (alternate) EOC. The backup EOC is located in a different and 

physically separate building (identified in the IIFP facility’s EP Implementing Procedures 

[EPIPs]) from the primary control room/EOC. The same documentation and communications 

capabilities available in the primary control room/EOC will also be available in the alternate 

onsite EOC. In addition, the applicant will designate an offsite EOC location (identified in the 

facility EPIPs), arranged as part of the community offsite assistance agreements, for use in the 

event the entire facility requires evacuation. The offsite EOC will have appropriate documents 

and communications equipment. 

Section 7.4 of the EP (IIFP, 2012a) describes offsite emergency support and equipment. This 

section provides information regarding fire, emergency medical services, and local law 

enforcement. The applicant will sign a memorandum of understanding (MOU) between IIFP 

and the Hobbs, NM, fire department for fire and medical emergency services 3 months before 

the startup of operations or receipt of licensed materials or hazardous process chemicals. IIFP 

will also sign MOUs with the Lea County Medical Center, Lea County Emergency Management, 

Lea County Sheriff’s Department, and the New Mexico Department of Homeland Security and 

Emergency Management at least 3 months before the startup of operations or receipt of 

licensed materials or hazardous process chemicals. These MOUs are not part of the License 

Application, but will be mutually signed by IIFP and the appropriate agency prior to operations. 
8-3 

Sections 8.2.2, 8.5.1.3, and 10.6.1 of the EP (IIFP, 2012a) describe the emergency monitoring 

equipment that is available for assessing radioactive releases, including personnel monitoring, 

air sampling, gaseous effluent monitoring, equipment for sampling soil, groundwater and 

vegetation, and portable radiological monitors. 

Sections 5.2 and 5.3.5 of the EP (IIFP, 2012a) describe the emergency monitoring equipment 

that is available for assessing process system upsets, including alarms in the control room, and 

fluoride detectors. The applicant will inventory and test emergency equipment on a quarterly or 

monthly basis, depending on the type of equipment, as discussed in Section 10.6 of the EP 

(IIFP, 2012a). The appropriate manager will be notified of any deficiencies that cannot be 

corrected so that repair or replacement can be expedited. 

Section 9.2 of the EP (IIFP, 2012a) describes the facility’s communications equipment, which 

includes facility telephones with facsimile capability, the public address system, alarms, and 

two- way radios. The communications systems are designed so that a failure in one system 

does not leave the facility without communications capability. For offsite communications, the 

facility’s telephone system is the primary means of communication; the facility’s radio system is 

the alternate means of emergency communication with offsite authorities. Onsite and offsite 

radios are tested quarterly. 

The applicant identified and described the onsite and offsite emergency facilities. The NRC 

staff finds that the information provided by the applicant meets the acceptance criteria in 

Section 8.4.3.1.2 of NUREG-1520 (NRC, 2002) and, together with new information discussed in 

Section 8.3.3 of this SER, meets the requirements of 10 CFR 40.31(j)(3)(ii) for new facilities 

because the applicant identified and described the onsite and offsite facilities that could be 

relied on in an emergency. The applicant’s onsite and offsite emergency organization and 

facilities have sufficient equipment and resources, and the NRC staff finds them acceptable. 

8.3.3 Types of Accidents 

Section 4.0 of the EP (IIFP, 2012a) identifies postulated events that have high and intermediate 

consequences. 

Since the IIFP processes only depleted uranium, in the form of DUF6, depleted uranium 

tetrafluoride (DUF4), and a blend of depleted uranium oxides, the offsite radiological 

consequences associated with plant accidents is limited. The applicant stated that no nuclear 

criticality potential exists at the facility, and there is no material onsite that contains fission 

products or transuranic elements. A potential exists to release depleted uranyl fluoride (UO2F2) 

into the environment as a reaction product from a DUF6 release. The applicant stated that no 

credible accident has been identified to pose intermediate- or high-radiological consequences to 

the public. Revision B of the IIFP Integrated Safety Analysis (ISA) Summary (IIFP, 2012b), 

Sections 4.1.2 and 7, contain figures projecting doses and toxic substance concentrations as a 

function of distance and time for various meteorological stability classes, including computer 

models and assumptions. The NRC staff finds these statements to be accurate due to the fact 

that the facility processes depleted uranium and does not possess special nuclear material. 

The NRC staff finds that established controls are in place to prevent or mitigate significant 

chemical or radioactive release at the plant site. Credible accident initiators have been 

identified and subsequent limits and controls established to control the frequency and severity of 

credible accidents such that acceptable risk to the public is maintained. 
8-4 

As stated above, the applicant identified and described the types of accidents identified in the 

EP for which protective actions may be needed. The NRC staff finds that the information 

provided by the applicant meets the acceptance criteria in Section 8.4.3.1.3 of NUREG-1520, 

(NRC, 2002) and 10 CFR 40.31(j)(3)(ii) for new facilities because IIFP has identified and 

described the types of accidents that may require protective actions. Therefore, the NRC staff 

finds that applicant’s description of the types of accidents is acceptable. 

8.3.4 Classification of Accidents 

IIFP has established emergency action levels consistent with Appendix A, “Examples of 

Initiating Conditions,” to NRC Regulatory Guide 3.67, “Standard Format and Content for 

Emergency Plans for Fuel Cycle and Materials Facilities” (NRC, 1992). Section 6.1 of the EP 

(IIFP, 2012a) describes the criteria used to classify an emergency as either an alert or a site 

area emergency declaration and defines both types of incidents. The applicant established that 

the threshold for escalating an event from an alert to a site area emergency declaration is based 

on the significance of the release to the environment and the possibility that the event could 

require a response by an offsite response organization. The EPIPs provide the processes for 

making the appropriate classification. Sections 6.1.1 and 6.1.2 of the EP (IIFP, 2012a) provide 

examples of site-specific incidents and the emergency classification that will be declared for 

each event. The Shift Superintendent is responsible for accident classification and assumes the 

responsibility of the Emergency Director until relieved by the Chief Operations Officer/Plant 

Manager (COO/PM) or his or her designee. 

The NRC staff finds that the applicant adequately identified and described the EP classification 

system. The information provided by the applicant meets the requirements of 

10 CFR 40.31(j)(3)(iii) and the acceptance criteria in Section 8.4.3.1.4 of NUREG-1520 

(NRC, 2002) for new facilities because IIFP has identified and described the two classifications 

of an alert and site area emergency, the classification that is expected for each accident 

identified in the EP, the emergency action levels at which an alert or site area emergency will be 

declared, and the personnel positions and alternates with responsibility for accident classifications. The applicant’s process for classification of accidents is, therefore, acceptable. 

8.3.5 Detection of Accidents 

Sections 2.2 and 8.3.3 of the EP explain the methods and systems available to detect accidents 

at the facility, including the following: 

• UF6, anhydrous hydrogen fluoride, silicon tetrafluoride, boron trifluoride, or uranium 

powder releases 

• natural phenomena 

• fires propagating between areas 

• fires involving excessive transient combustibles 

• process system upsets. 

The operating procedures contain directions for manual actions to be taken to bring the facility 

to a safe condition. 
8-5 

The NRC staff finds that the applicant has adequately identified and described the methods and 

systems for the detection of accidents. The information provided by the applicant meets the 

requirements of 40.31 (j)(3)(iv) and the acceptance criteria in Section 8.4.3.1.5 of NUREG-1520 

(NRC, 2002) for new facilities because IIFP identified and described the means of detecting an 

accident, the means of detecting any release of radioactive material or hazardous chemicals, 

the means of alerting the operating staff, and the anticipated response of the operating staff. 

Therefore, the NRC staff finds that the applicant’s program for detection of accidents is 

acceptable. 

8.3.6 Mitigation of Consequences 

Section 8.3 of the EP (IIFP, 2009) describes actions and equipment that the applicant will use to 

mitigate the consequences of accidents at the facility. The major hazard is a chemical spill or 

release. The main features used at the facility to mitigate the consequences of accidents 

associated with a chemical release, natural phenomena, or fire include automatic isolation, 

interruption or termination of specific operations, fire detection and suppression systems, 

operator response to abnormal conditions or alarms, and process area dikes as a means of 

secondary containment of spilled materials. Section 8.4 of the EP (IIFP, 2012a) describes 

personnel evacuation and accountability plans, as well as the staging, maintenance, and use of 

protective equipment and supplies. 

As stated above, the applicant identified and described the mitigation of consequences. The 

information provided by the applicant meets the requirements of 10 CFR 40.31(i)(3)(v) and the 

acceptance criteria in Section 8.4.3.1.6 of NUREG-1520 (NRC, 2002) for new facilities because 

IIFP identified and described the measures and equipment to be used for safe shutdown and 

the mitigation of consequences to workers onsite, as well as the public offsite. Therefore, the 

NRC staff finds that the applicant’s program for mitigation of consequences is acceptable. 

8.3.7 Assessment of Releases 

Section 8.2 of the EP (IIFP, 2012a) describes the actions that the applicant will take to assess 

the extent of an accident at the facility. In case of an alert declaration, the spill control team 

monitors releases to provide assessment of dose or intake projections of facility personnel. The 

Environmental and Safety Officer (ESO) will advise on any additional sampling or monitoring for 

assessing effects on the environment or public, if needed. The applicant stated that, for an alert 

classification, it is unlikely that the assessment of dose and intake effects would extend beyond 

personnel within the facility. In case of a site area emergency declaration, in addition to 

monitoring performed for an alert declaration, monitoring personnel will conduct radiation and 

chemical surveys of assembly areas, the EOC, and appropriate areas around the emergency 

scene. Environmental air sampling will be conducted at selected and predetermined locations 

for uranium and fluorides. The ESO may determine the need to conduct other environmental 

sampling depending on the potential impact to the plant and offsite area. 

As stated above, the applicant identified and described the assessment of releases. The 

information provided by the applicant meets the requirements of 10 CFR 40.31(j)(3)(vi) and the 

acceptance criteria in Section 8.4.3.1.7 of NUREG-1520 (NRC, 2002) for new facilities because 

IIFP identified and described the procedures to be used to assess the release of radioactive 

material or hazardous chemicals incident to the processing of licensed material. The applicant’s 

program for assessment of releases is, therefore, acceptable. 
8-6 

8.3.8 Responsibilities 

Sections 7.0 and 7.3, respectively, of the EP (IIFP, 2012a) describe the responsibilities of facility 

personnel during normal operations and emergency situations. In case of an emergency, the 

Shift Superintendent assumes the duties of the Emergency Director until the COO/PM, or 

designee, arrives. As stated in Section 7.2 of the EP, the Emergency Director ensures the 

response effort. Specific responsibilities of the Emergency Director include the following: 

• declaring or terminating emergency events, as stated in Section 7.3.1 of the EP 

• notifying appropriate offsite organizations and the NRC, as stated in Sections 6.2.1 and 

6.2.2 for alert and site area emergency declarations, respectively 

• making recommendations to offsite authorities concerning protective actions for the 

public offsite, as stated in Section 8.4.2 of the EP 

• determining onsite protective actions for facility personnel, as stated in Section 8.4.1 of 

the EP 

• initiating special sampling and monitoring as required, as stated in Sections 6.2.1 and 

6.2.2 of the EP 

• requesting support from offsite organizations, as stated in Section 7.4 of the EP 

• approving press statements before their release, as stated in Section 7.3.1 of the EP. 

Section 7.3 of the EP (IIFP, 2012a) summarizes the responsibilities of the emergency response 

organization (ERO) positions. Section 7.4 of the EP (IIFP, 2012a) provides the local offsite 

assistance to the facility for fire, emergency medical services, and local law enforcement. IIFP 

will sign MOUs with local agencies 3 months before the startup of operations or receipt of 

licensed materials or hazardous process chemicals, which will be reviewed annually and 

renewed at least every 4 years, or more frequently if necessary. As stated in Section 10.1 of 

the EP (IIFP, 2012a), departmental administrative procedures are established which assign 

responsibility for the development, review, approval, and update of the EP and its supporting 

procedures. The EP is reviewed by emergency preparedness personnel for accuracy and 

updated annually, as needed. 

As stated above, the applicant identified and provided a description of responsibilities. The 

information provided by the applicant meets the acceptance criteria in Section 8.4.3.1.8 of 

NUREG-1520 (NRC, 2002) for new facilities because IIFP identified and described the ERO and 

administration that ensure effective planning, implementation, and control of emergency 

preparedness activities. The applicant’s description of the emergency preparedness 

responsibilities is, therefore, acceptable. 

8.3.9 Notification and Coordination 

As discussed in Section 8.3.4 of this Safety Evaluation Report (SER), the Emergency Director is 

responsible for the classification of emergencies, which is outlined in Section 6.1 of the EP 

(IIFP, 2012a). Section 6.2 of the EP (IIFP, 2012a) provides a clear commitment to promptly 

notify offsite EROs of an emergency, including the notification to the NRC Operations Center 
8-7 

within 1 hour. Section 7.4 of the EP (IIFP, 2012a) adequately describes the provisions for 

assistance from offsite EROs. 

In the NRC staff’s opinion, these sections adequately describe the agreements held between 

IIFP and local offsite EROs and agencies, procedures for access to the site for these response 

organizations, and equipment and services available from these organizations. 

As discussed in Section 8.3.8 of this SER, the Emergency Director is responsible for 

(1) declaring an alert or site area emergency, (2) activating the onsite ERO, (3) notifying offsite 

emergency response authorities of an emergency, (4) notifying the NRC Operations Center, 

(5) deciding which onsite protective actions to initiate, (6) deciding which offsite protective 

actions to recommend and providing that information to offsite authorities, (7) deciding to 

request support from offsite organizations, and (8) deciding to terminate the emergency or enter 

recovery mode. 

As stated above, the applicant identified and described the provisions for notification and 

coordination. The information provided by the applicant meets the requirements of 

10 CFR 40.31(j)(3)(viii) and the acceptance criteria in Section 8.4.3.1.9 of NUREG-1520 

(NRC, 2002) for new facilities because IIFP identified and described the emergency notification 

procedures that enable the organization to correctly classify emergencies, notify emergency 

response personnel, and initiate or recommend appropriate actions in a timely manner. The 

applicant’s program for notification and coordination is, therefore, acceptable. 

8.3.10 Information To Be Communicated 

Section 6.3 of the EP (IIFP, 2012a) adequately describes the type of information to be given to 

offsite EROs and the NRC during an emergency. IIFP will use an event notification form as a 

script for initial notification of an emergency at the facility to appropriate offsite emergency 

response authorities, which will then be transmitted electronically as a confirmation to the 

notification. The emergency notification form also serves as a communications update to offsite 

organizations during an emergency event. If a site area emergency is declared, 

recommendations will be made to offsite organizations for use in advising the public. Facility 

meteorological data provided in the EOC may be used to provide more specific information. 

As stated above, the applicant identified and described the notification and coordination 

procedures. The information provided by the applicant meets the requirements of 

10 CFR 40.31(j)(3)(ix) and the acceptance criteria in Section 8.4.3.1.10 of NUREG-1520 

(NRC, 2002) for new facilities because IIFP identified and described the use of a standard 

reporting checklist, the types of information to be provided, the preplanned protective action 

recommendations, the offsite authorities to be notified, and the recommended actions to be 

implemented by offsite organizations for each accident discussed in the EP. The applicant’s 

description of the information to be communicated is, therefore, acceptable. 

8.3.11 Training 

Section 10.2 of the EP (IIFP, 2012a) describes the training IIFP will provide to workers on how 

to respond to an emergency. All workers receive general employee training, which includes 

quality assurance, radiation protection, safety, emergency and administrative procedures, and 

applicable regulations. The applicant also provides training in radiation protection and 

emergency procedures specific to each type of job function. The facility emergency response 

personnel receive additional training to provide specific information about how the ERO 
8-8 

responds during emergency conditions and includes EOC staffing, determining and estimating 

potential offsite releases of radiation and chemicals, and interface with offsite assistance 

organizations. This training is required before an individual is assigned to the emergency 

organization, and IIFP provides refresher training at least once every year. 

The IIFP administrative procedures contain the specific topics, performance objectives, content, 

training schedules, and number of training hours required for each position. The Environmental 

Safety and Health (ESH) Manager, or designee, reviews the contents of the formal radiation 

protection and hazard communications training programs and updates the information as 

required at least every 2 years to ensure that the programs are current and adequate. 

Individuals requiring unescorted access to the plant controlled access area receive annual 

retraining. 

Periodic training is offered to offsite assistance organization personnel in accordance with IIFP 

EPIPs. IIFP facility personnel meet at least every 2 years with each offsite assistance group to 

accomplish training and review items of mutual interest, including relevant changes to the 

program. This training includes facility tours, information concerning facility access control 

(normal and emergency), potential accident scenarios, emergency action levels, notification 

procedures, exposure guidelines, personnel monitoring devices, communications, 

contamination control, and the offsite ERO role in responding to an emergency at the IIFP 

facility, as appropriate. 

As stated above, the applicant identified and described its emergency preparedness training 

program. The information provided by the applicant meets the requirements of 

10 CFR 40.31(j)(3)(x) and the acceptance criteria in Section 8.4.3.1.11 of NUREG-1520 

(NRC, 2002) for new facilities because IIFP identified and described the topics and general 

content of the training programs. The areas addressed include the administration of the training 

programs, the training provided for the use of protective equipment, the training program for 

onsite personnel who are not members of the EROs, and special instructions and orientation 

tours provided for offsite organizations. The applicant’s proposed training program is, therefore, 

acceptable. 

8.3.12 Recovery and Plant Restoration (Safe Shutdown) 

Section 12.0 of the EP (IIFP, 2012a) describes key positions and responsibilities of the recovery 

organization and describes how reentry and restoration will be accomplished. Once significant 

offsite releases and the emergency condition are terminated, deliberate reentry and restoration 

activities may be applied. Visual, chemical, and radiological monitoring activities will be 

performed as necessary to identify any ongoing minor releases and to identify any areas with 

residual chemical or radioactive contamination. Reentry into areas affected by hazardous 

chemical or radioactive releases will be authorized in accordance with plant safety and health 

physics procedures. 

Visual and engineering assessments of plant systems will be implemented to determine the 

need for repairs, modification, and functional testing. Recovery activities will be conducted in 

accordance with plant radiation protection procedures and all employee exposures will be 

controlled in accordance with the facility’s program to keep exposures and doses as low as 

reasonably achievable. Before resuming normal operations, the COO/PM, or designee, will 

notify the appropriate local, State, and Federal authorities, informing them that the facility is in 

compliance with applicable environmental regulations. Within 15 days after any emergency 

event involving the offsite release of hazardous materials, the COO/PM, or designee, will submit 
8-9 

a written report on the incident to the U.S. Environmental Protection Agency’s Regional 

Administrator that characterizes the extent of the release and provides an assessment of 

hazards to human health. 

As stated above, the applicant identified and described the safe shutdown process. The 

information provided by the applicant meets the requirements of 10 CFR 40.31(j)(3)(xi) and the 

acceptance criteria in Section 8.4.3.1.12 of NUREG-1520 (NRC, 2002) for new facilities 

because IIFP identified and described the methods and responsibilities for assessing damage, 

procedures for determining actions to reduce releases of radioactive material or hazardous 

chemicals, provisions for accomplishing required restoration actions, and key positions in the 

recovery organization. The applicant’s program for safe shutdown, recovery, and facility 

restoration is, therefore, acceptable. 

8.3.13 Exercises and Drills 

Section 10.3 of the EP (IIFP, 2012a) identifies adequate provisions for periodic drills and 

biennial exercises to test the adequacy of EPIPs and emergency equipment and 

instrumentation and to ensure that all ERO personnel are familiar with and proficient in their 

duties. 

Section 10.4 of the EP (IIFP, 2012a) provides that post exercise evaluations will be conducted 

by those involved and appropriate improvements will be implemented. Areas evaluated include 

the adequacy of the EP, procedures, equipment, facilities, personnel training, and overall 

response effectiveness. The ESH Manager is responsible for the planning, scheduling, and 

conducting of emergency response drills and exercises for the facility. Offsite EROs are invited 

to participate in the biennial exercise, and the NRC is invited to participate or observe. IIFP will 

submit exercise objectives and scenarios to the NRC for review and comment at least 30 days 

before the exercise. Section 10.3.2 of the EP (IIFP, 2012a) includes an adequate provision for 

quarterly communications checks to verify the operability of equipment used by the ERO to 

communicate with offsite agencies and response organizations. 

As stated above, the applicant identified and described its approach to exercises and drills. The 

information provided by the applicant meets the requirements of 10 CFR 40.31(j)(3)(xii) and the 

acceptance criteria in Section 8.4.3.1.13 of NUREG-1520 (NRC, 2002) for new facilities 

because IIFP identified and described the conduct of drills and exercises in a manner that 

demonstrates the capability of the organization to plan and perform an effective response to an 

emergency. 

 In Section 13 of the EP (IIFP, 2012a) the applicant stated that it will maintain compliance with 

the Emergency Planning and Community Right-to-Know Act of 1986 with respect to any 

hazardous materials possessed at the plant site. The information provided by the applicant 

meets the acceptance criteria in Section 8.4.3.1.1 of NUREG-1520 (NRC, 2002) and 

10 CFR 40.31(j)(3)(xiii); therefore, the NRC staff finds the applicant’s plan for the treatment of 

hazardous chemicals acceptable. 

8.3.14 Responsibilities for Developing and Maintaining the Emergency Plan and its 

Procedures 

Section 10.1 of the EP (IIFP, 2012a) describes the EPIPs that state the specific duties, 

responsibilities, emergency action levels, and actions to be taken by responders identified in 

Section 7.3 of the EP (IIFP, 2012a). The section also describes administrative procedures in 
8-10 

place that ensure that all individuals and groups assigned responsibilities in an emergency have 

easy access to a current copy of each procedure that pertains to their functions. It also 

describes other administrative procedures that assign responsibility for the development, 

review, approval, and update of the EP and supporting procedures. The applicant reviews and 

updates the EP annually as needed and will not implement changes that would decrease the 

effectiveness of the EP without prior NRC approval. IIFP will provide any proposed change that 

affects an offsite organization to that organization for review and comment at least 60 days 

before the change being implemented, unless mutually agreed otherwise. The applicant 

provides that revisions to the EP and supporting procedures be distributed to all affected parties 

and submitted to the NRC within 3 months of the revision. Section 10.5 of the EP also 

describes the annual audit to be performed on the EP as part of the applicant’s quality 

assurance program. 

As stated above, the applicant identified and described the responsibilities for developing and 

maintaining the emergency program and its procedures. The information provided by the 

applicant meets the acceptance criteria in Section 8.4.3.1.24 of NUREG-1520 (NRC, 2002) for 

new facilities because IIFP identified and described the responsibilities for developing and 

maintaining the emergency program and its procedures. 

8.4 Evaluation Findings

The NRC staff evaluated IIFP’s EP (IIFP, 2012a) for the proposed facility. The applicant has 

established an EP for responding to the radiological hazards resulting from a release of 

radioactive material or hazardous chemicals relating to the processing of licensed material, in 

accordance with 10 CFR 40.31(j)(1)(ii). The NRC staff reviewed IIFP’s EP with respect to the 

requirements of 10 CFR 40.31(j)(3), 70.64(a)(6) and the acceptance criteria in Section 8.4.3 of 

NUREG-1520 (NRC, 2002). The NRC staff concludes that the IIFP EP is adequate to 

demonstrate compliance with the regulatory requirements. In particular, the applicant has 

ensured that: (1) the facility is properly configured to limit releases of radioactive materials or 

hazardous chemicals in case of an accident; (2) a capability exists for measuring and assessing 

the significance of accidental releases of radioactive materials or hazardous chemicals; 

(3) appropriate emergency equipment and procedures are provided onsite to protect workers 

against radiation and other chemical hazards that might be encountered after an accident; (4) a 

system has been established to notify Federal, State, and local government agencies and to 

recommend appropriate protective actions to protect members of the public; and (5) the 

necessary recovery actions have been established to return the facility to a safe condition after 

an accident. 

The requirements of the EP are implemented through approved written procedures. Changes 

that decrease the effectiveness of the EP may not be made without NRC approval. The NRC 

will be notified of other changes that do not decrease the effectiveness of the EP within 6 

months of making the changes. 

8.5 References

(IIFP, 2012a) International Isotope Fluorine Products, Inc., “Emergency Plan Rev B of IIFP 

License Application,” May 2012, Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML12123A245. 
8-11 

(IIFP, 2012b) International Isotopes Fluorine Products, Inc., “ISA Summary Rev. B for IIFP,” 

May 2012, Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML12123A245. 

(NRC, 2002) U.S. Nuclear Regulatory Commission, “Standard Review Plan for the Review of a 

License Application for a Fuel Cycle Facility,” NUREG-1520, Rev. 0, March, 2002. 

(NRC, 1992) U.S. Nuclear Regulatory Commission, “Standard Format and Content for 

Emergency Plans for Fuel Cycle and Materials Facilities,” Regulatory Guide 3.67, 1992. 

 
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9-1 

9.0 ENVIRONMENTAL PROTECTION 

The purpose of the U.S. Nuclear Regulatory Commission’s (NRC’s) review of the International 

Isotopes Fluorine Products, Inc.’s (IIFP or the applicant), Environmental Protection Plan for its 

proposed Fluorine Extraction Process and Uranium Deconversion Process (FEP/DUP) plant is 

to determine whether the applicant’s proposed environmental protection measures are 

adequate to protect the environment and the health and safety of the public, as required by 

Title 10 of the Code of Federal Regulations (10 CFR) Part 20, “Standards for Protection against 

Radiation”; 10 CFR Part 40, “Domestic Licensing of Source Material”; 10 CFR Part 51, 

“Environmental Protection Regulations for Domestic Licensing and Related Regulatory 

Functions”; and 10 CFR Part 70, “Domestic Licensing of Special Nuclear Material.” 

NUREG-1748, “Environmental Review Guidance for Licensing Actions Associated with NMSS 

Programs,” issued August 2003, provides general procedures for the environmental review of 

licensing actions regulated by the Office of Nuclear Material Safety and Safeguards. 

9.1 Regulatory Requirements

• 10 CFR Part 20 provides the effluent control and treatment measures necessary to meet 

the dose limits and dose constraints for members of the public specified in Subpart B, 

“Radiation Protection Programs”; Subpart D, “Radiation Dose Limits for Individual 

Members of the Public”; and Subpart F, “Surveys and Monitoring.” Subpart F also 

specifies the survey requirements. Subpart K, “Waste Disposal,” specifies the waste 

disposal requirements; Subpart L, “Records,” specifies the records requirements; and 

Subpart M, “Reports,” specifies the reporting requirements. 

• 10 CFR Part 51 provides that the applicant must establish effluent and environmental 

monitoring systems to provide the information required by 10 CFR 51.60(a). 

• 10 CFR 40.32(c) provides that the application will be approved if the proposed 

equipment, facilities, and procedures are adequate to protect health and minimize 

danger to life and property. 

• 10 CFR 40.65, “Effluent Monitoring Reporting Requirements,” provides the reporting 

requirements for radiological effluent monitoring for a 10 CFR Part 40 licensee. 

• 10 CFR 70.65(b)(6) provides that an applicant for a facility must provide an Integrated 

Safety Analysis (ISA) Summary that includes a list of the items relied on for safety 

(IROFS). 

9.2 Regulatory Acceptance Criteria

Section 9.4.3 of NUREG-1520, Revision 0, “Standard Review Plan for the Review of a License 

Application for a Fuel Cycle Facility” (NRC, 2002), outlines the acceptance criteria for the NRC’s 

review of the IIFP Environmental Protection Program. 
9-2 

9.3 Staff Review and Analysis

The applicant has established and implemented personnel training and qualification 

requirements in accordance with approved written procedures. Chapter 11 of the license 

application (LA) (IIFP, 2012a) describes the applicant’s training program; Chapter 2 of the LA 

(IIFP, 2012a) describes its qualification requirements for key management positions. 

9.3.1 Personnel Training 

As noted in Section 11.3 of the LA (IIFP, 2012a), the applicant’s training programs are provided 

through shared responsibility of the environmental health and safety (EHS) disciplines and line 

management. The training programs are designed primarily to ensure that personnel who 

perform activities relied on for safety have the appropriate skills to design, operate, and maintain 

the facility in a safe manner. Training requirements are applicable primarily to those personnel 

who have a direct relationship to various aspects of IROFS. In addition to its safety focus, the 

training program is also designed to ensure that assigned personnel are trained and tested as 

necessary to perform the responsibilities important to the protection of the environment. 

Technical training consists of initial, on-the-job, continuing, and special training, as applicable to 

assist personnel gain an understanding of specific, assigned technical duties. Professional 

development, which uses internal or external professionals via formal workshops, tutorials, and 

selected training programs, assists personnel in gaining additional understanding of technical 

practices common to their assigned job functions. Job-specific training is performance based. 

Lesson plans are based on job performance requirements and are reviewed by line management and by the responsible organization for the subject matter. Under its training program, the 

applicant also provides continuing or periodic retraining to ensure that personnel remain 

proficient and retain important knowledge and skills. Retraining may be required as a result of 

facility modifications or changes in procedures or in the Quality Assurance Program, which 

would result in new or changed information. The applicant evaluates trainee understanding and 

proficiency through observation, demonstration, or examinations, and documents training 

results. 

The applicant also performs periodic evaluations of its training program to assess program 

effectiveness. These evaluations identify program strengths and weaknesses. The evaluations 

are also used to help determine whether training content matches current job needs and 

whether corrective actions are needed to improve training program effectiveness. In addition, 

the applicant may also perform independent audits to evaluate the overall training program 

effectiveness. 

9.3.2 Personnel Qualifications 

Chapter 2, “Organization and Administration” of the LA (IIFP, 2012a), identifies key 

management and supervisory positions, hierarchy, and functions, including personnel 

qualifications for each key position. Figure 2-2, “Plant Operation Organization,” of the LA 

(IIFP, 2012a), provides an overview of those management positions and functions; Section 2.2 

of the LA (IIFP, 2012a) describes those key management positions and qualifications that relate 

to, among other things, environmental protection. 

Chapter 2 of this Safety Evaluation Report (SER) describes the overall personnel qualifications. 

This section highlights the qualifications of key environmental protection personnel. 
9-3 

The project environmental assessment lead (EAL) is a professional staff contractor working 

under the project oversight and support contract that is approved by the President/Chief 

Executive Officer (CEO) of International Isotopes Inc. (INIS)/IIFP. During the design and 

construction stages, the EAL ensures that environmental technical and licensing support, as 

requested by the Chief Operations Officer (COO)/Commercial Facility Project Director or IIFP’s 

Regulatory/Quality Assurance Director, is provided for evaluation and assessment of design or 

engineering modifications or construction activities. The EAL also provides technical support for 

the Federal, State, and local environmental-related permit applications. A primary responsibility 

of the EAL during the design/construction stage of the project is to prepare responses for and 

interact with the NRC about requests for additional information related to the licensing review of 

the IIFP’s Environmental Report (ER) (IIFP, 2011). The above responsibilities transfer to the 

IIFP’s ESH Manager and designated environment staff as those positions are filled and the 

transition to operations is completed. The EAL shall have, as a minimum, a bachelor’s degree 

in engineering or a scientific field and a minimum of 5 years in a chemical, radiological, or 

nuclear facility; with at least 3 of those years in responsible environmental assignments. The 

EAL must have experience in interacting with regulatory agencies and a working knowledge of 

relevant regulatory requirements. 

The COO/Plant Manager (PM) appoints the ESH Manager at the facility, with the concurrence of 

the INIS Regulatory Affairs and Quality Assurance Director (RAQD). The ESH Manager reports 

to the IIFP COO/PM and also has a reporting and interacting relationship with the RAQD on 

matters of ESH policies, regulatory requirements, plant safety, and environmental compliance. 

In addition, the ESH Manager has the authority and responsibility to elevate any ESH concerns 

to corporate management and the INIS President/CEO. The IIFP ESH Manager establishes 

and oversees the Radiation Protection (RP), Licensing, Integrated Safety Analysis, Industrial 

Safety, Environmental Protection, Fire Protection, and Emergency Preparedness/Security 

Programs to ensure compliance with applicable Federal, State, and local regulations and laws. 

Those programs are designed to ensure the health and safety of employees and the public, as 

well as the protection of the environment. The ESH Manager shall have, as a minimum, a 

bachelor’s degree in engineering, science, or a related field and 5 years of responsible 

assignments of ESH activities at chemical, radiological, or nuclear facilities. 

The environmental lead reports to the IIFP ESH Manager and has responsibilities in support of 

the ESH Manager, which include, but are not limited to: (1) developing environmental programs 

and procedures, (2) leading monitoring and measuring activities, (3) developing and maintaining 

environmental-related permits, (4) assisting in training of employees in environmental matters, 

(5) conducting audits and inspections, (6) preparing and providing environmental data and 

reports, and (7) interacting with Federal, State, and local representatives in ensuring compliance 

with permit requirements and conditions. The environmental lead shall have a bachelor’s 

degree in engineering or a scientific field and at least 2 years of environmental-related 

experience in a chemical, radiological, or nuclear facility. 

The NRC staff reviewed the applicant’s training program identified in Section 11 of its LA 

(IIFP, 2012a) and the personnel qualifications identified in Section 2 of its LA (IIFP, 2012a). 

This review included the qualification and training of managers, supervisors, and technical staff 

who are associated with environmental protection. The NRC staff finds that the applicant has 

established that the training, testing, and qualification of these personnel meets the 

requirements found in 10 CFR 40.32(b) and with guidance found in Section 9.4.3.2 of 

NUREG-1520, Revision 0 (NRC, 2002), which in turn incorporates, by reference, the training 

and qualifications information in Section 11.4.3.3 of NUREG-1520, Revision 0 (NRC, 2002). 

Section 11.4.3.3 of NUREG-1520 (NRC, 2002) identifies training areas and provides that 
9-4 

managers and staff are to possess levels of education and experience commensurate with the 

responsibilities of their positions. The NRC staff finds the personnel qualifications of the 

proposed facility’s staff adequate because the minimum qualifications of personnel are 

commensurate with the assigned responsibility and the training of personnel who perform 

regulated activities and they also meet the acceptance criteria in Section 11.4.3.3 of 

NUREG-1520 (NRC, 2002) 

9.3.3 Radiation Safety 

9.3.3.1 As Low As Reasonably Achievable Goals for Air and Liquid Effluent Control 

Title 10 CFR 20.1101, “Radiation Protection Programs,” requires each licensee to develop, 

document and implement an RP program. This environmental review of the IIFP RP Program 

focuses on the applicant’s proposed methods to maintain public doses as low as reasonably 

achievable (ALARA), in accordance with 10 CFR 20.1101. Acceptance criteria found in 

Section 9.4.3.2.1 of NUREG-1520, Revision 0 (NRC, 2002), state that the applicant’s proposed 

ALARA Program is acceptable if the program complies with 10 CFR 20.1101 according to 

Regulatory Guide 8.37, “ALARA Levels for Effluents from Materials Facilities,” issued July 1993 

(NRC, 1993). In addition, Section 9.4.3.2.1 of NUREG-1520 (NRC 2002) states the applicant 

should establish its ALARA goals at a modest fraction (10 to 20 percent) of the values in 

Appendix B, “Annual Limits on Intake (ALIs) and Derived Air Concentrations (DACs) of 

Radionuclides for Occupational Exposure; Effluent Concentrations; Concentration for Release 

to Sewerage” to 10 CFR Part 20, Columns 1 and 2 and Table 3, and the external exposure limit 

in 10 CFR 20.1302(b)(2)(ii); or, if the applicant proposes to demonstrate compliance with 

10 CFR 20.1301, the dose limit for members of the public. The applicant’s constraint approach 

is acceptable if it is consistent with the guidance in Regulatory Guide 4.20, “Constraint on 

Releases of Airborne Radioactive Material to the Environment for Licensees Other Than Power 

Reactors,” issued December 1996 (NRC, 1996), and provides sufficient detail to demonstrate 

specific application of the guidance to proposed routine and nonroutine operation, including 

anticipated events. The applicant should also describe its proposed effluent controls to maintain 

public doses ALARA and demonstrate a commitment to reduce unnecessary exposure to 

members of the public and releases to the environment. According to Section 9.4.3.2.1 in 

NUREG-1520, Revision 0 (NRC, 2002), the applicant also needs to commit to annual reviews of 

the content and implementation of the RP Program, which includes the ALARA Program. These 

reviews should consider analysis of trends in release concentrations, environmental monitoring 

data, and radionuclide usage; determinations of whether operational changes are needed; and 

evaluations of designs for system installations or modifications. In addition, the results of the 

annual reviews should be reported to senior management, along with recommendations for 

changes to facilities or procedures necessary to achieve ALARA goals. 

Sections 4.2 and 9.2 of the applicant’s LA (IIFP, 2012a) and Section 4.12.2 of its 

ER (IIFP, 2011) describe the applicant’s ALARA and Radiation Protection Programs. The 

applicant stated that it maintains and uses treatment systems, as appropriate, to ensure that 

releases of radioactive material to unrestricted areas remain below the limits specified in 

10 CFR 20.1301 and in accordance with ALARA policy. The applicant has an Environmental 

Protection Program for the proposed IIFP facility. The applicant stated that the primary purpose 

of this program is to ensure that exposure of the workers, public, and the environment to 

radioactive materials used in facility operations is kept ALARA. As discussed in Chapter 6 of 

the applicant’s ER (IIFP, 2011), compliance with the ALARA concept is a part of the applicant’s 

Environmental Protection Program. The applicant relies on air and liquid effluent controls to 
9-5 

maintain public doses ALARA. The applicant’s ALARA goal is 20 percent of the values found in 

Appendix B to 10 CFR Part 20. 

The NRC staff finds that the applicant’s approach is sufficiently detailed to demonstrate that IIFP 

is in compliance with the regulatory dose limits found in 10 CFR 20.1301; that air and liquid 

dose constraints meet the acceptance criteria in Section 9.4.3.2.1 of NUREG-1520, Revision 0 

(NRC, 2002); and that the applicant’s ALARA Program for controlling gaseous and liquid 

effluents is consistent with the guidance found in Regulatory Guide 8.37 (NRC, 1993). The 

NRC staff determines that the ALARA Goals for Air and Liquid Effluent Control are adequate 

because the applicant committed to use effluent controls to maintain public doses ALARA 

9.3.3.2 Air Effluent As Low As Reasonably Achievable Goal 

The applicant proposed an ALARA goal for radiological effluents to air of less than 

0.1 millisievert per year (mSv/yr) [10 millirem per year (mrem/yr)] total effective dose 

equivalent (TEDE), which is within the 10 CFR 20.1101 constraint of 0.1 mSv/yr (10 mrem/yr) 

TEDE for the maximally exposed member of the public. The NRC staff determined that the 

applicant’s proposal satisfies the ALARA goal of 0.1 mSv/yr (10 mrem/yr) recommended in 

Regulatory Guide 8.37, Regulatory Position C.1.2, “ALARA Goals” (NRC, 1993), and meets the 

applicable acceptance criterion found in Section 9.4.3.2.1 of NUREG-1520, Revision 0 

(NRC, 2002). Therefore, the NRC staff finds the applicant’s proposal to be acceptable because 

ALARA goals are set to a fraction of the values in Table 2, Columns 1 and 2, and Table 3 of 

Appendix B to 10 CFR Part 20. 

9.3.3.3 Liquid Effluent As Low As Reasonably Achievable Goal 

As noted in Sections 4.2, and 9.2.1 of the applicant’s LA (IIFP, 2012a) and Section 4.12.2 of its 

ER (IIFP, 2011), no liquid effluent discharges are expected because of the anticipated low 

volume of contaminated liquid waste and the effectiveness of the treatment process. Thus NRC 

staff finds that liquid effluent does not provide a significant pathway for radiological exposure to 

the general public. The applicant stated that liquid plant effluents are maintained on the IIFP 

site, and there is no discharge of process wastewater radioactive wastes. Therefore, staff 

concluded, average annual release concentrations of liquid effluents meet the limits in 

10 CFR 20.1302, “Compliance with Dose Limits for Individual Members of the Public,” and will 

not exceed the values in Appendix B to 10 CFR Part 20. Liquid effluents result in negligible 

increases to the environmental or public radiological exposures. NRC staff determined that 

ALARA levels for liquid effluents are consistent with the guidance given in 

Regulatory Guide 8.37 (NRC, 1993). The RP procedures incorporate the ALARA philosophy 

into facility operations and ensure that exposures are kept below the 10 CFR 20.1101(d) limits. 

The level of radioactivity from liquid effluents to the public and the environment must be below 

the regulatory goal of 0.1 mSv/yr (10 mrem/yr) recommended in Regulatory Guide 8.37 

(NRC, 1993) and meet the applicable acceptance criteria found in Section 9.4.3.2.1 of 

NUREG-1520, Revision 0 (NRC, 2002), and is, therefore, acceptable to the NRC staff. 

The NRC staff finds the ALARA goal with respect to liquid effluents acceptable because it 

complies with the acceptance criteria in NUREG-1520 Section 9.4.3.2.1, Revision 0 

(NRC, 2002), and 10 CFR 20.1101(b) and (d). 
9-6 

9.3.3.4 Air Effluent Controls To Maintain Public Doses As Low As Reasonably Achievable 

 

Sections 4.6.1 and 9.2.1 of the applicant’s LA (IIFP, 2012a) and Section 4.12.2 of its 

ER (IIFP, 2011) discuss the air effluent controls for the proposed IIFP facility. Most of the 

airborne depleted uranium is removed through filtration by prefilters, high-efficiency filters, and 

carbon-bed filters before entering the three-stage scrubber system. After scrubbing, the 

gaseous effluent is discharged into the atmosphere. In Section 2.1.3 of its ER (IIFP, 2011), the 

applicant provided a detailed description of the filtration and scrubber systems, which overall 

achieve greater than 99 percent efficiency. Table 2-1 of the ER (IIFP, 2011) provides design 

efficiencies for process vent offgas treatment equipment. The stacks are sampled continuously 

and routinely analyzed to measure the radioactivity of the discharged gases. 

As noted in Section 4.12.2.2 of the ER (IIFP, 2011), the applicant estimated offsite radiological 

impacts to key receptors (critical populations) from routine effluent releases using the GENII 

model (Version 2.06). The key receptor locations for determining dose impacts included the 

resident nearest to the IIFP facility and the maximally exposed individual (MEI) (at the northwest 

boundary). The MEI is a hypothetical person living at the point of highest projected total 

uranium concentrations near the site boundary. The applicant evaluated the impact from 

inhalation of gaseous effluents, immersion in a passing effluent plume, exposure to direct 

radiation from deposited radioactivity on the ground surface (ground plane exposure), and 

ingestion of contaminated food products. Because the applicant does not anticipate 

contamination of drinking water, its analysis did not consider radiological contamination of 

drinking water. The applicant’s analysis included dose equivalent assessments for four age 

groups—adults, teens, children, and infants—for these pathways. 

9.3.3.5 Staff Evaluation of Air Effluent Controls 

The NRC staff evaluated the air effluent controls and effects described above. The applicant 

calculated the dose equivalents associated with gaseous effluents for the MEI and the nearest 

resident, by pathway, for the total body in adults, teens, children, and infants. Tables 4-25 and 

4-26, respectively, of the ER (IIFP, 2011) present these results. The calculated committed 

effective dose equivalent (CEDE) for the adult MEI from the proposed IIFP facility emissions is 

8.4×10-6 mSv/yr (8.4×10-4 mrem/yr). For the adult, full-time resident nearest the facility, the 

calculated CEDE from the IIFP facility is 6.4×10-5 mSv/yr (6.4×10-3 mrem/yr). Doses for public 

receptors at other sites of interest (e.g., schools and hospitals) would be lower than the MEI 

because the airborne concentrations of uranium are lower at these more distant locations. The 

doses for the MEI and nearest resident are well below (by orders of magnitude) the 

U.S. Environmental Protection Agency’s (EPA’s) 0.1 mSv/yr (10 mrem/yr) standard in 

40 CFR Part 190, “Environmental Protection Standard for Nuclear Power Operations.” This 

estimated maximum public dose is also well below the 0.1 mSv (10 mrem) ALARA constraint on 

air emissions described in 10 CFR 20.1101. Because public receptors at other sites of interest 

are more distant than the MEI, their doses would be even lower because of uranium dispersion 

at the more distant locations. 

The NRC staff finds that the applicant’s air effluent controls ensure that radiation levels to the 

public remain well below regulatory limits and ALARA air effluent goals, and that the applicant’s 

approach to effluent control meets the applicable acceptance criteria found in Section 9.4.3.2.1 

of NUREG-1520, Revision 0 (NRC, 2002). Therefore, the NRC staff finds these controls 

acceptable. The NRC staff also finds that the applicant has demonstrated that its air effluent 

controls reduce releases to provide adequate protection of the environment and of the health 

and safety of the public. 
9-7 

9.3.3.6 Liquid Effluent Controls to Maintain Public Doses As Low As Reasonably Achievable 

As noted in Section 4.12.2 of the applicant’s ER (IIFP, 2011), the general public may be 

impacted by radioactive material from liquid and gaseous effluent discharges associated with 

controlled releases from the uranium process lines during routine operations of the proposed 

facility. However, as noted in Section 9.3.1.3 of this SER, no liquid discharges from the 

proposed facility are expected because of the anticipated low volume of contaminated liquid 

waste and the effectiveness of treatment processes. Also, as noted in Section 4.13.3.2 of the 

applicant’s ER (IIFP, 2011), the proposed facility does not discharge any process effluents to 

natural surface waters or grounds; and there is no tie into a publicly owned treatment works. All 

process wastes are recycled. The applicant does not expect rainfall runoff from the cylinder 

storage pads to be a significant exposure pathway. Runoff water is directed to an onsite 

retention basin for evaporation of the collected water. No liquids from retention basins are 

discharged offsite. Therefore, the applicant does not anticipate any radiological contamination 

of drinking water. Thus, the applicant stated that liquid effluents do not provide a significant 

pathway for radiological exposure to the public, and is within the effluent release requirements 

in Appendix B to 10 CFR Part 20. 

9.3.3.7 Staff Evaluation of Liquid Effluent Controls 

The applicant stated, as noted above and in Section 9.2.1.2 of the LA (IIFP, 2012a), process 

water discharges are treated and are contained onsite either by recycling and reusing in the 

process or by evaporating. Cooling water is recycled. The proposed facility liquid effluent 

collection and recycle systems provide a means to control liquid waste and maintain a process 

water practical mass balance using flow-surge tanks, scrubber solution regeneration/recycle, 

and evaporation equipment. Uranium is precipitated from wastewater streams and disposed as 

low-level radioactive waste. In addition, storm water runoff is collected and transported to a 

retention basin via the plant storm water sewer system. The storm water is stored temporarily 

in the retention basin until it is sampled and then evaporated or discharged. Any discharged 

storm water contains only trace levels of radiological contamination, and the NRC staff 

determined that these releases are within the effluent release requirements in Appendix B to 

10 CFR Part 20. 

The NRC staff finds that the applicant’s liquid effluent controls ensure that radiation levels to the 

public remain well below regulatory limits and ALARA liquid effluent goals and that the 

applicant’s approach to effluent control meets the applicable acceptance criteria found in 

Section 9.4.3.2.1 of NUREG-1520, Revision 0 (NRC, 2002). Therefore, the NRC staff finds 

these controls to be acceptable. The NRC staff also finds that the applicant has demonstrated 

that its liquid effluent controls reduce releases to provide adequate protection of the 

environment and of the health and safety of the public. 

9.3.3.8 As Low As Reasonably Achievable Program Reviews and Reports to Management 

In Section 4.2 of its LA (IIFP, 2012a), the applicant described its ALARA Program for the 

proposed facility. Section 4.3.2 of this SER also describes the ALARA Program. The ALARA 

Program provides for an annual review of the content and implementation of the RP Program, 

including the Effluent Control Program. The ALARA Committee, which is part of the Facility 

Safety Review Committee (FSRC), is responsible for conducting annual ALARA reviews. The 

FSRC is an independent advisory committee that reports directly to the COO/PM. 
9-8 

Membership of the ALARA Committee includes the following positions: 

• COO/PM 

• RP Manager 

• selected department managers 

• ESH Manager 

• selected supervisors and hourly personnel. 

The scope of the ALARA Committee’s activities include, at a minimum, annual review of the 

following: 

• site radiological operating performance; including trends in airborne concentrations, 

personnel exposures, and environmental monitoring results 

• operations and exposure records to identify opportunities to reduce exposures 

• employee training and methods for using information on the job to keep exposure 

ALARA 

• potential modifications of procedures and equipment when changes reduce exposures at 

reasonable cost. 

In addition, the ALARA Committee reviews major changes in authorized activities affecting RP 

practices and evaluates contamination minimization and removal activities. 

The proceedings, findings, and recommendations of the ALARA Committee are reported in 

writing to the COO/PM and appropriate line management and area managers responsible for 

operations reviewed by the Committee. Such reports are retained for a minimum of 3 years. 

Based on expected improvements, updated performance data, economics, and consideration of 

other site priorities, management decides which of the ALARA Committee recommendations 

should be pursued. If a specific recommendation is pursued, a task owner is assigned, and the 

action is tracked to completion. 

As discussed in Sections 4.1.1.3 and 4.3 of the LA (IIFP, 2012a), the RP Manager is 

responsible for the overall implementation of the RP Program and has direct access to the 

COO/PM who, in turn, has the overall responsibility for safety and activities conducted at the 

proposed facility. 

As described in Section 4.3.1 of this SER, the applicant implements the ALARA Program in 

accordance with 10 CFR 20.1101 through approved written procedures and policies. The 

approach described meets the above-referenced requirements in 10 CFR Part 20 and the 

applicable acceptance criteria found in Section 9.4.3.2.1 of NUREG-1520, Revision 0 

(NRC, 2002), and is, therefore, acceptable to the NRC staff because the applicant describes 

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

contamination of the facility and environment and minimize the generation of radioactive waste. 
9-9 

9.3.3.9 Waste Minimization 

As required by 10 CFR 20.1406, “Minimization of Contamination,” an applicant for a license 

must describe how the facility design and procedures for operation 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. 

Section 9.4.3.2.1 of NUREG-1520, Revision 0 (NRC, 2002), addresses applicable acceptance 

criteria. The applicant’s program for waste minimization is acceptable if the applicant describes 

how the facility’s design procedures for operation minimize, to the extent practicable, 

contamination of the facility and the environment; facilitate decommissioning; and minimize, to 

the extent practicable, the generation of radioactive waste. In addition, the acceptance criteria 

in NUREG-1520 Section 9.4.3.2.1 requires for the program to have senior management 

support, provide methods to characterize waste generation and waste management costs, 

provide for periodic waste minimization assessments, provide for technology transfer to seek 

and exchange technical information on waste minimization, and provide methods to implement 

and evaluate waste minimization recommendations. 

In Section 4.7.8 of its LA (IIFP, 2012a) and in Section 4.13.4 of its ER (IIFP, 2011), the applicant 

committed to design and operate its facility to meet the requirements in 10 CFR 20.1406 to 

minimize contamination; facilitate eventual decommissioning; and minimize, to the extent 

practicable, the generation of radioactive waste. The applicant has assigned high priority to 

minimizing the generation of waste through reduction, reuse, or recycling. For example, the 

applicant incorporated closed-loop cooling systems in the design to reduce water use. In 

addition, the applicant maintains ALARA controls during facility operation to account for 

standard waste minimization practices, as directed in 10 CFR Part 20. As part of its waste 

management program described in Section 3.12 of the applicant’s ER (IIFP, 2011), minimization 

practices include reclamation, recycle, reuse, compaction, and design features or procedures to 

avoid or reduce the generation of wastes. 

In Section 1.1.6 of its LA (IIFP, 2012a), the applicant described its solid waste management 

program at the proposed facility for industrial (nonhazardous), radioactive, and hazardous 

wastes. Solid waste is grouped into one of these waste categories. The applicant may send 

wastes that are candidates for volume reduction, recycling, or treatment to licensed treatment 

facilities that have the ability to reduce the volume of most Class A low-level radioactive waste. 

The applicant also described its program to reuse or recycle process effluents. 

The NRC staff evaluated and determined that the applicant meets the requirements specified in 

10 CFR 20.1406, which provides that IIFP adopt procedures to minimize the generation of 

radioactive waste. Furthermore, the applicant’s waste minimization practices are consistent 

with the guidance provided in Regulatory Guide 4.21, “Minimization of Contamination and 

Radioactive Waste Generation: Life-Cycle Planning,” issued June 2008 (NRC, 2008), and meet 

applicable acceptance criteria in Section 9.4.3.2.1of NUREG-1520, Revision 0 (NRC, 2002). 

Therefore, the NRC staff finds the applicant’s proposed waste minimization program acceptable. 

The NRC staff finds that the applicant’s implementation of its program for managing solid 

radiological and nonradiological wastes related to facility operation, including its volume 

reduction and recycling programs, reduce unnecessary exposures to these wastes and ensure 

adequate protection of public health and safety and the environment. 
9-10 

9.3.4 Effluent and Environmental Monitoring 

The regulations in 10 CFR 20.1302 require a licensee to make or cause to be made, as 

appropriate, surveys of radiation levels in unrestricted and controlled areas and radioactive 

material in effluents released to unrestricted and controlled areas to demonstrate compliance 

with the dose limits for individual members of the public in 10 CFR 20.1301. 

Section 9.4.3.2.2 of NUREG-1520, Revision 0 (NRC, 2002), addresses the acceptance criteria 

for effluent monitoring. NUREG-1520 Revision 0 (NRC, 2002) states that an applicant’s effluent 

monitoring program is acceptable if radioactive materials concentrations in airborne and liquid 

effluents are ALARA and are below the limits specified in Table 2 of Appendix B to 10 CFR 

Part 20, or if site-specific limits are established in accordance with 10 CFR 20.1302(c). In 

addition, if the applicant proposes to demonstrate compliance with 10 CFR 20.1301 by using 

calculations of the TEDE, the applicant needs to perform pathway analyses using appropriate 

models, codes, and assumptions according to the acceptance criteria in NUREG-1520, 

Revision 0 (NRC, 2002), Section 9.4.3.2.2. 

Consistent with NUREG-1520, Revision 0 (NRC, 2002), each of the following criteria should to 

be addressed. The applicant should identify and monitor all liquid and airborne effluent 

discharge locations and continuously sample airborne effluents from all routine and nonroutine 

operations, unless periodic sampling or other means has been justified. Sample collection and 

analysis methods and frequencies need to be appropriate for the effluent medium and the 

radionuclides being sampled. Radionuclide specific analyses need to be performed on 

appropriate samples using justified methods. In addition, the minimum detectable concentration 

for sample analysis needs to be adequate (sufficiently sensitive) for comparison to the 

concentration limits in Appendix B to 10 CFR Part 20, and laboratory quality control procedures 

need to be adequate to validate the analytical results. The applicant also needs to establish 

action levels and proposed action if action levels are exceeded. In addition, the applicant needs 

to completely and accurately describe all applicable Federal or State discharge limits for 

gaseous and liquid effluents applicable to the proposed facility. Leakage detection systems also 

need to be in place to detect leaks from tanks, ponds, or lagoons that could affect groundwater, 

surface water, and soils. The applicant needs to control and maintain releases to sewer 

systems to meet the requirements of 10 CFR 20.2003, “Disposal by Release into Sanitary 

Sewerage.” The applicant also needs to have reporting procedures that meet 10 CFR 40.65, 

“Effluent Monitoring Reporting Requirements.” In addition, the applicant’s procedures and 

facilities for solid and liquid waste handling, storage, and monitoring need to result in safe 

storage and timely disposition of the material. 

Section 9.4.3.2.2 in NUREG-1520, Revision 0 (NRC, 2002), contains the acceptance criteria for 

environmental monitoring. An applicant’s environmental monitoring program should include the 

establishment of background and baseline radionuclide concentrations in environmental media. 

Monitoring should include sampling and analyses for air, surface water, groundwater, soil, 

sediments, and vegetation, as appropriate; and identify adequate and appropriate sampling 

locations and frequencies for each environmental medium and the analyses to be performed for 

each medium. Monitoring procedures should employ acceptable analytical methods and 

instrumentation, and instrumentation should be appropriately maintained and calibrated. If the 

applicant proposes to use its own laboratory for environmental sample analyses, the applicant 

should commit to providing third-party verification of its methods. In addition, the applicant 

should identify appropriate action levels and actions to be taken if action levels are exceeded for 

each environmental medium and radionuclide. The applicant should select action levels based 

on pathway analyses that demonstrate that, below those concentrations, doses meet the 
9-11 

ALARA and 10 CFR Part 20, Subpart B, limits. The applicant also should specify minimum 

detectable concentrations for sample analyses at least as low as those selected for effluent 

monitoring in air and water consistent with the selected action levels. Data analysis methods 

and criteria should be provided for evaluating and reporting the results of environmental 

sampling and indicate when an action level is being approached in time to take corrective 

actions. The applicant also should provide a description of the status of all licenses, permits, 

and other approvals for facility operation that is complete and accurate. In addition, the program 

should be adequate to assess environmental impacts from potential radioactive and 

nonradioactive material releases as identified in high- and intermediate-consequence events in 

the ISA Summary (IIFP, 2012b). 

The IIFP Radiation Safety Program incorporates the applicant’s effluent monitoring program. 

The applicant described its effluent and environmental monitoring programs for radiological and 

nonradiological effluents released from the proposed facility in Section 9.2.2 of its LA 

(IIFP, 2012a) and in Chapter 6 of its ER (IIFP, 2011). In ER Chapter 6 (IIFP, 2011), the 

applicant indicated that it performs measurements and monitoring necessary to demonstrate 

that the amount of radioactive material present in effluent from the proposed facility is kept 

ALARA, in compliance with 10 CFR Part 20. The applicant also identified guidance in 

Regulatory Guide 4.16, Revision 2, “Monitoring and Reporting Radioactivity in Releases of 

Radioactive Materials in Liquid and Gaseous Effluents from Nuclear Fuel Processing and 

Fabrication Plants and Uranium Hexafluoride Production Plants,” issued December 2010 

(NRC, 2010), to ensure that it adheres to the ALARA principle such that there is no undue risk 

to the public health or safety at or beyond the IIFP site boundary. 

9.3.4.1 Air Effluent Monitoring 

The IIFP Radiation Safety Program includes the air effluent monitoring program. Radioactive air 

effluent from the facility is discharged only through monitored pathways (release points). The 

NRC staff stated that uranium isotopes and daughter products are expected to be the prominent 

radionuclides in the gaseous effluent. 

9.3.4.2 Expected Concentrations 

Expected concentrations of radioactive materials in airborne and liquid effluents are addressed 

in Section 9.2.2 of the applicant’s LA (IIFP, 2012a); and Section 4.12.2 of its ER (IIFP, 2011) 

address expected concentrations of radioactive materials in airborne and liquid effluents, which 

were estimated using conservative assumptions. As described in Section 9.3.3.2 of this SER, 

the NRC staff finds that the expected concentrations of radioactive materials in airborne 

effluents is well below the regulatory limits specified in 10 CFR 20.1302(c). The applicant 

demonstrated compliance with air effluent limits by calculating the TEDE to the individual who is 

likely to receive the highest dose, in accordance with 10 CFR 20.1302(b)(1). 

The NRC staff finds that the applicant’s control of these concentrations is ALARA and below 

limits specified in Table 2 of Appendix B to 10 CFR Part 20. In addition, the applicant’s 

approach to controlling these concentrations meets the applicable acceptance criteria found in 

Section 9.4.3.2.2 of NUREG-1520, Revision 0 (NRC, 2002), and is, therefore, acceptable to the 

NRC staff. 
9-12 

9.3.4.3 Total Effective Dose Equivalent 

The applicant established constraints on atmospheric releases for the proposed facility such 

that no member of the public is expected to receive a TEDE in excess of 0.1 mSv/yr 

(10 mrem/yr) from these releases. Written procedures approved by IIFP management dictate 

that atmospheric releases be monitored and measured. In Section 4.12.2 of the ER 

(IIFP, 2011), the applicant provided a detailed description of the calculation of the TEDE to the 

individual who is likely to receive the highest dose using GENII (version 2.08), which 

implements dosimetry models recommended by the International Commission on Radiological 

Protection (ICRP). For both the inhalation and ingestion exposure pathways, the applicant used 

the exposure-to-dose conversion factors in Federal Guidance Report 11 (EPA, 1988). For 

direct dose from material deposited on the ground plane or from the passing plume, the 

applicant used the conversion factors in Federal Guidance Report 12 (EPA, 1993). The 

ingestion pathway models performed by the applicant were taken from Regulatory Guide 1.109, 

Revision 1, “Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for 

the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I,” issued October 1977 

(NRC, 1977), and then multiplied by the relative age-dependent dose factor found in ICRP 

Publication 72, “Age dependent Doses to the Members of the Public from Intake of 

Radionuclides—Part 5, Compilation of Ingestions and Inhalation Coefficients” (ICRP, 1996). 

The applicant calculated the direct dose equivalent from the 30-year life expectation of the 

proposed facility by using the MCNP4C2 computer code (ORNL, 2000). In Tables 4-25 and 

4-26 of its ER (IIFP, 2011), the applicant provided calculated dose equivalents for the MEI and 

nearest resident (adult, teen, child, and infant) to the proposed facility, which the NRC staff 

noted are well below (by orders of magnitude) the regulatory limit in 10 CFR Part 20 and well 

within the EPA regulatory standard in 40 CFR Part 190. The NRC staff determined that the 

TEDE calculation to the individuals likely to receive the highest dose in accordance with 

10 CFR 20.1302(b)(1) was appropriate because it accurately represent the facility, site and 

surrounding area. The NRC staff finds that the applicant’s approach acceptable because it 

meets the requirements in 10 CFR 20.1302(b)(1) and meets the applicable acceptance criteria 

found in Section 9.4.3.2.2 of NUREG-1520, Revision 0 (NRC, 2002). 

9.3.4.4 Locations, Continuous Sampling, and Analysis Methods of Airborne Effluents 

Sections 9.2.2 and 1.1.6.3 of its LA (IIFP, 2012a) and Section 6.1.1 of its ER (IIFP, 2011) 

describe the applicant’s Gaseous Effluent Monitoring Program. Discharges of gaseous effluents 

have the highest potential of introducing uranium into the environment. Public exposure from 

routine operations at the proposed facility may occur as a result of gaseous effluents, including 

controlled releases from the uranium deconversion process lines during decontamination and 

maintenance of equipment. The applicant’s Gaseous Effluent Monitoring Program is designed 

to determine the quantities and concentrations of gaseous discharges to the environment. The 

sources of air emissions from the proposed facility are the process building stacks and dust 

collector stacks, which are described in Chapter 9 of the applicant’s LA (IIFP, 2012a) and in 

Chapter 6 of the applicant’s ER (IIFP, 2011). Table 6-1 of the applicant’s ER (IIFP, 2011) 

identifies the type and frequency of sample analyses for the Gaseous Effluent Sampling 

Program. 

As noted above in Section 9.3.3.4 of this SER, the ventilation system air stream passes through 

a series of high efficiency particulate air and high efficiency gas absorption filters before being 

vented to the atmosphere through the process building stacks and monitored for uranium and 

hydrogen fluoride (HF). Continuous air sampler filters are analyzed each week for gross alpha 

and beta radiation. 
9-13 

The applicant stated that it annually reviews the trends in gaseous emissions and liquid effluent 

monitoring data to determine whether changes are needed in systems or practices to achieve 

the ALARA effluent goals. The applicant stated it will revise the program as appropriate to 

maintain its effectiveness as changes are noted, such as those related to operations or other 

factors identified in Chapter 6 of the ER (IIFP, 2011). 

In Section 6.2 of its ER (IIFP, 2011), the applicant stated that it collects HF continuously on 

particulate filters that are in vent stacks and analyze these samples weekly. As noted in 

Section 6.1.2 of its ER (IIFP, 2011) and in Section 9.2.2.1 of the LA (IIFP, 2012a), the applicant 

also collects uranium isotopes continuously on particulate filters in air samplers located around 

the proposed facility and initially analyze these samples daily. 

The NRC staff finds that the applicant has identified and monitors all airborne effluent discharge 

locations to determine contributions to dose limits, in accordance with requirements in 

10 CFR Part 20. The NRC staff also finds that the applicant’s effluent monitoring meets the 

acceptance criteria found in Section 9.4.3.2.2 of NUREG-1520, Revision 0 (NRC, 2002), which 

is applicable to both airborne and liquid effluents. The NRC staff further finds that the applicant 

continuously monitors samples of airborne effluents from all routine and nonroutine operations 

in a manner that meets the acceptance criteria found in Section 9.4.3.2.2 of NUREG-1520, 

Revision 0 (NRC, 2002); and ensures that the air sample collection and analysis methods and 

frequencies described in the IIFP Radiological Monitoring Program are appropriate to meet the 

acceptance criteria found in Section 9.4.3.2.2 of NUREG-1520, Revision 0 (NRC, 2002), which 

is applicable to both airborne and liquid effluent discharges. Therefore, the NRC staff finds the 

applicant’s airborne effluent monitoring and analysis program to be acceptable. 

9.3.4.5 Radionuclide-Specific Analyses 

In Section 9.2.2.1 of its LA (IIFP, 2012a), the applicant indicates they perform radionuclide 

specific analyses on selected composite samples, which Section 6.1.1 of its ER (IIFP, 2011) 

identifies. Because uranium in gaseous effluent may exist in a variety of compounds (e.g., 

depleted uranium hexafluoride, uranium oxide, depleted uranium tetrafluoride, and depleted 

uranyl fluoride [UO2F2]), the radiation program staff maintain, review, and assess the effluent 

data to ensure that the gaseous effluents meet with regulatory release criteria for uranium. 

Monitoring reports, in which the quantities of individual radionuclides are estimated on the basis 

of methods other than direct measurement, include an explanation and justification of how the 

results were obtained. In Section 9.2.2.1 of its LA (IIFP, 2012a), the applicant committed to 

follow guidance specified in Regulatory Guide 4.16, Revision 2 (NRC, 2010). The applicant will 

submit to the NRC a semiannual report that includes the concentrations of the principal 

radionuclides released in the unrestricted area, as well as other information necessary to 

evaluate the radiation doses from effluent releases to the public in liquid and gaseous effluents. 

The report also includes the minimum detectable concentration (MDC) for analysis and the error 

for each data point. 

The NRC staff finds that the applicant’s proposed radionuclide-specific analyses and reporting 

meet the requirements of 10 CFR 40.65 and the guidance specified in Regulatory Guide 4.16, 

Revision 2 (NRC, 2010) because the applicant has committed to submit a report which specifies 

the quantity of each radionuclide released to unrestricted areas. The proposed radionuclidespecific analyses and reporting also meet the applicable acceptance criteria found in 

Section 9.4.3.2.2 of NUREG-1520, Revision 0 (NRC, 2002) because the effluents are below the 
9-14 

limits specified in Table 2 of Appendix B to 10 CFR Part 20; and they are sampled according to 

procedures. The radionuclide specific analysis is acceptable to the NRC staff. 

9.3.4.6 Minimum Detectable Concentrations 

The applicant’s Gaseous Effluent Monitoring Program is designed to determine the quantities 

and concentrations of gaseous effluents discharged into the environment. Uranium isotopes 

and “daughter” decay products are expected to be the prominent radionuclides in the gaseous 

effluent. In Section 6.1.1 of its ER (IIFP, 2011), the applicant indicated that the MDC of 

3.7×10-11 becquerels per milliliter (Bq/ml) (1.0×10-15microcuries per milliliter (uCi/ml) is a 

detection requirement for the Gaseous Effluent Monitoring Program for all gross alpha analyses 

performed on gaseous effluent samples. The Gaseous Effluent Monitoring Program is not 

required to detect concentrations in samples that are below the MDC value. The NRC staff 

finds that the MDC value for effluents in air is below the concentration limits found in Table 2 of 

Appendix B to 10 CFR Part 20. 

The NRC staff finds that the MDCs for various listed media are sufficiently low to meet action 

level, regulatory, and permit requirements and are consistent with applicable acceptance criteria 

found in Section 9.4.3.2.2 of NUREG-1520, Revision 0 (NRC, 2002), and are, therefore, 

acceptable to the NRC staff. 

9.3.4.7 Laboratory Quality Control 

Section 9.2.2.1 of the applicant’s LA (IIFP, 2012a) and Section 6.1.2.2 of its ER (IIFP, 2011) 

address laboratory quality control (QC). The applicant’s laboratory QC procedures conform to 

the guidance in Regulatory Guide 4.15 “Quality Assurance for Radiological Monitoring Programs 

(Inception through Normal Operations to License Termination)—Effluent Streams and the 

Environment” (NRC, 2007). Laboratory QC procedures include the use of established 

standards, such as those used by the National Institute of Standards and Technology (NIST), as 

well as standard analytical procedures, such as those established by the National 

Environmental Laboratory Accreditation Conference (NELAC). 

The applicant committed to use written procedures to ensure that sampling and monitoring 

equipment is properly maintained, calibrated, and in good working condition. Onsite and 

contractor laboratories participate in third-party intercomparison programs to validate the 

laboratories’s performance. Examples of such programs include the Mixed Analyte 

Performance Evaluation Program and the U.S. Department of Energy’s Quality Assurance 

Program. The applicant also stated that it requires that all radiological and nonradiological 

laboratory vendors be certified by the NELAC or an equivalent State laboratory accreditation 

agency. 

The NRC staff finds that the applicant’s laboratory QC procedures are adequate to validate the 

analytical results to ensure compliance with the monitoring requirements in 10 CFR Part 20 and 

meet applicable acceptance criteria in Section 9.4.3.2.2 of NUREG-1520, Revision 0 

(NRC, 2002). The laboratory QC procedures use recognized established standards and 

standard analytical procedures. Therefore, the NRC staff finds these procedures acceptable. 

9.3.4.8 Action Levels 

An action level for environmental measurements is the concentration (or mass) of an analyte 

that indicates that some action needs to be taken, such as initiating an investigation or, if the 
9-15 

level is sufficiently high, shutting down operations. Action levels provide guidance for ensuring 

that concentrations of radioactivity are within the limits found in Appendix B to 10 CFR Part 20. 

The applicant has implemented a program of corrective actions to ensure that the cause for 

exceeding the action level can be identified and corrected; applicable regulatory agencies are 

notified, if required; lessons learned are communicated to appropriate personnel; and applicable 

operational procedures are revised accordingly, if needed. As noted in Section 9.2.2.1 of its LA 

(IIFP, 2012a), the applicant set administrative action levels at 50 percent of values found in 

Table 2 of Appendix B to 10 CFR Part 20, which is sufficiently below compliance levels needed 

to permit implementation of corrective actions before regulatory limits are exceeded. Approved 

written procedures specify these action levels according to the type of sample and the specific 

analysis. 

The applicant’s proposed development of action levels and related actions meets the regulatory 

requirements of Appendix B to 10 CFR Part 20 and the applicable acceptance criteria found in 

Section 9.4.3.2.2 of NUREG-1520, Revision 0 (NRC, 2002), because action levels are set to 

50 percent of values found in Table 2 of Appendix B to 10 CFR 20. This allows implementation 

actions to be taken when the limits are exceeded. The NRC staff finds the action levels 

acceptable. 

9.3.4.9 Federal and State Permits 

Section 1.4 of the ER (IIFP, 2011) identifies and describes statutory and regulatory requirements of the NRC and other Federal agencies and departments, as well as the State of New 

Mexico, which are applicable to the construction and operation of the proposed facility. In 

addition, Section 1.5, Table 1-4, lists required Federal and State permits, as well as their status. 

The applicant did not identify any applicable local government requirements. Section 1.3.2.2 of 

the ER (IIFP, 2011) lists the preliminary Federal and State codes and standards for the design 

and construction of the proposed facility. 

The NRC staff finds that the applicant accurately described applicable Federal and State 

standards for discharges—as well as any permits issued by Federal, and State for gaseous and 

liquid effluents, as documented in Sections 1.4 and 1.5 of the ER (IIFP, 2011). This meets the 

applicable acceptance criteria found in Section 9.4.3.2.2 of NUREG-1520, Revision 0 

(NRC, 2002), and is, therefore, acceptable to the NRC staff. 

9.3.4.10 Air Effluent Monitoring Summary Evaluation 

Based on the NRC staff’s review of the applicant’s LA (IIFP, 2012a) and ER (IIFP, 2011), 

described in Section 9.3.3.1 of this SER, the NRC staff finds that the air effluent monitoring 

program maintains air effluent concentrations below the regulatory limits in Appendix B to 

10 CFR Part 20 by detecting and measuring concentrations of radioactivity in air effluents during 

operation of the proposed facility. Furthermore, the NRC staff finds that the applicant’s air 

effluent monitoring program meets the applicable acceptance criteria found in Section 9.4.3.2.2 

of NUREG-1520, Revision 0 (NRC, 2002), and is, therefore, acceptable. 

9.3.4.11 Liquid Effluent Monitoring 

The IIFP Radiation Safety Program encompasses the liquid effluent monitoring program. As 

noted in Section 9.3.1.1 of this SER and in Sections 4.2 and 9.2.1 of the applicant’s LA 

(IIFP, 2012a) and Chapter 4.12.2 of the ER (IIFP, 2011), no liquid effluent discharges are 

expected by the applicant from the facility because of the anticipated low volume of 
9-16 

contaminated liquid waste and the effectiveness of the treatment process. Liquid plant effluents 

are maintained on the IIFP site, and there is no discharge of process wastewater radioactive 

wastes. There is no offsite release of liquid effluents to unrestricted areas. Thus, the NRC staff 

determined that liquid effluents result in negligible increases to the environmental or public 

radiological exposures. 

9.3.4.12 Detection of Leaks to Groundwater, Surface Water, or Soil 

As noted in Section 9.2.2.1 of its LA (IIFP, 2012a), the applicant’s conceptual design includes 

appropriate spill and leak control pads and containment dikes. The applicant stated that liquid is 

monitored to ensure compliance with the limits in Appendix B to 10 CFR Part 20, and with the 

National Pollutant Discharge Elimination System permit levels for fluoride and other constituents 

specified in the permit. The applicant stated that compliance is demonstrated through effluent 

and environmental sampling data. The applicant conducts sampling at the site’s storm water 

retention (evaporation) basin and at the DUF6 cylinder storage pad. Groundwater sampling is 

conducted at four locations onsite. As described in Section 10.3.2.3 of this SER, the NRC staff 

expects any discharges of liquid effluents are to be negligible and within the regulatory limits of 

10 CFR Part 20, Appendix B. Therefore, the NRC staff finds the applicant’s detection of leaks 

to groundwater, surface water or soil acceptable. 

9.3.4.13 Releases to Sewer Systems 

As noted by the applicant in Section 9.2.2.1 of its LA (IIFP, 2012a) and in Section 6.1.1.2 of its 

ER (IIFP, 2011), all liquid effluents are to be maintained on the facility’s site. The proposed 

facility does not discharge any process effluents into natural surface waters or grounds. No 

liquid effluents are released to the sewer system, and there is no tie into the publicly owned 

treatment works. Sanitary waste is treated onsite and rendered suitable for horticultural 

purposes. 

Because there are no radiological liquid effluents releases to the sewer system and no 

radiological releases from the sewer system to the environment, the NRC staff finds that the 

releases to the sewer systems are controlled and maintained in a manner sufficient to meet the 

regulatory requirements of 10 CFR 20.2003 and to meet the acceptance criteria found in 

Section 9.4.3.2.2 of NUREG-1520, Revision 0 (NRC, 2002). Therefore, the NRC staff finds the 

applicant’s method for controlling liquid effluent releases to be acceptable. 

9.3.4.14 Reporting Procedures 

In Section 9.2.2.1 of its LA (IIFP, 2012a), the applicant committed to implementing the recording 

guidance specified in Regulatory Guide 4.16, Revision 2 (NRC, 2010). As noted in the 

discussion of radionuclide-specific analyses in Section 10.3.3.1 of this SER, the NRC expects 

the applicant to submit a single semiannual effluent report that includes the concentrations of 

principal radionuclides released in the unrestricted area, as well as other information necessary 

to evaluate the radiation doses to the public from liquid and gaseous effluent releases. The 

report also would include the MDC for analysis and the error for each data point. 

 

The NRC staff finds that the applicant has committed to reporting procedures that meet the 

requirements of 10 CFR 40.65 and are consistent with the guidance specified in Regulatory 

Guide 4.16, Revision 2 (NRC, 2010). The procedures also meet the applicable acceptance 

criteria found in Section 9.4.3.2.2 of NUREG-1520, Revision 0 (NRC, 2002), and are, therefore, 

acceptable to the NRC staff. 
9-17 

9.3.4.15 Liquid and Solid Waste Handling 

The applicant addressed liquid and solid waste management in Section 3.12 of its ER 

(IIFP, 2011) and Section 9.2.2.1of its LA (IIFP, 2012a). As discussed in Sections 10.3.2.3 and 

10.3.2.5 of this SER, as well as in Section 9.2.2.1 of the applicant’s LA (IIFP, 2012a), fluoride 

bearing waste liquors are treated and recycled in the plant’s scrubber system. Uranium 

removed from liquid streams is collected and sent to a licensed, low-level radioactive waste 

disposal site, along with waste uranium oxides produced by the deconversion process. 

Nonprocess waste liquids that are determined to contain regulated or hazardous contaminants 

are collected and disposed at an offsite licensed facility. 

Solid waste is grouped into three categories: industrial (nonhazardous), radioactive, and 

hazardous waste. The applicant may send wastes that are candidates for volume reduction, 

recycling, or treatment to licensed treatment facilities that have the ability to reduce the volume 

of most Class A low-level radioactive waste. (See the LA, Section 1.1.6, Table 1-3, “Estimated 

Annual Waste Generated at the IIFP Facility” [IIFP, 2012a]). 

The NRC staff finds that the applicant’s program for management of liquid and solid wastes 

reduces unnecessary exposures in accordance with the 10 CFR Part 20 ALARA requirements; 

meets the waste handling acceptance criteria found in Section 9.4.3.2.2 of NUREG-1520; 

Revision 0 (NRC, 2002), and is, therefore, acceptable. 

9.3.4.16 Liquid Effluent Monitoring Summary Evaluation 

Based on the NRC staff’s review of the applicant’s LA (IIFP, 2012a) and ER (IIFP, 2011) 

described in Section 9.3.2.2 of this SER, the NRC staff finds that the Liquid Effluent Monitoring 

Program during operation of the proposed facility detects and measures concentrations of 

radioactivity in liquid effluent in a manner sufficient to demonstrate that liquid effluent 

concentrations remain below the regulatory limits in 10 CFR Part 20 and meet the applicable 

acceptance criteria found in Section 9.4.3.2.2 of NUREG-1520, Revision 0 (NRC, 2002). 

Therefore, the NRC staff finds the IIFP Liquid Effluent Monitoring Program acceptable. 

9.3.4.17 Environmental Monitoring 

Section 9.2.2.2 of the applicant’s LA (IIFP, 2012a) and Chapter 6 of its ER (IIFP, 2011) describe 

the applicant’s environmental monitoring program. Table 6.2 of the ER (IIFP, 2011) 

summarizes the program, including the medium (e.g., air, groundwater, soil, vegetation), sample 

locations, sample type, and parameter (e.g., gross alpha, gamma, beta) and frequency. 

Section 4.12.2 of the ER (IIFP, 2011) also discusses environmental monitoring during facility 

operation. 

9.3.4.18 Background and Baseline Radionuclide Concentrations 

The primary objective of the applicant’s Radiological Environmental Monitoring Program 

(REMP) is to provide verification that the operations at the facility do not result in detrimental 

radiological impacts on the environment. To meet this objective, the applicant collects and 

analyzes representative samples from various environmental media for the presence of 

facility-related radioactivity. The REMP includes the collection of data during preoperational 

years to establish baseline radiological information against which operational radiological 

information can be compared. Such comparisons provide a means of assessing the magnitude 
9-18 

of potential radiological impacts on members of the public and of demonstrating compliance with 

applicable RP regulatory requirements. 

The NRC staff finds that the applicant has established an acceptable program to collect 

background and baseline concentrations of radionuclides in environmental media through 

sampling and analyses under the REMP for the proposed facility. These data are used as part 

of its RP Program and trending analyses to ensure that doses to individual members of the 

public are within the dose limits identified in 10 CFR Part 20, Subpart D. In addition, the REMP 

meets the applicable acceptance criteria in Section 9.4.3.2.2 of NUREG-1520, Revision 0 

(NRC, 2002), and is, therefore, acceptable to the NRC staff. 

9.3.4.19 Sampling and Analyses for Monitoring 

Table 6.2 of the applicant’s ER (IIFP, 2011) summarizes the REMP, including the medium 

sampled, sampling locations, sample type (e.g., thermal luminescent dosimeters, grab 

samples), and analyte/parameter frequency (e.g., total uranium sampled quarterly). 

As noted in Section 9.3.3.1 of this SER, the applicant stated that air monitoring includes 

continuous sampling of the facility release points, as well as weekly composite samples for 

gross alpha activity and concentrations of uranium isotopes from active air monitors placed 

around the restricted area fence line. The applicant conducts periodic surveys in and around 

outdoor storage areas and use dosimeters at the fence line to ensure that direct radiation doses 

are maintained ALARA. 

The groundwater monitoring program includes analysis of four onsite wells. Well samples are 

collected semiannually. Sampling begins before commencement of operations to establish 

baseline groundwater conditions and continue throughout operations and decommissioning. 

The soil and vegetation monitoring includes onsite and offsite sample locations. Samples are 

collected quarterly and analyzed for uranium concentrations. Sampling continues throughout 

operations and decommissioning. Sediment samples are collected semiannually from the storm 

water runoff retention basins onsite and analyzed for uranium concentrations. 

The NRC staff finds that the applicant’s monitoring program, which includes sampling and 

analyses for monitoring air, surface water, groundwater, soil, sediments, and vegetation, meets 

the applicable acceptance criteria in Section 9.4.3.2.2 of NUREG-1520, Revision 0 

(NRC, 2002), and is, therefore, acceptable to the NRC staff. The NRC staff also finds that the 

applicant’s descriptions in Section 9.2.2 of its LA (IIFP, 2012a) and Section 6.1 of its ER 

(IIFP, 2011) adequately identify sample media, locations, types, and sampling frequencies and 

are consistent with applicable guidance found in Section 9.4.3.2.2 of NUREG-1520, Revision 0 

(NRC, 2002), and are, therefore, acceptable to the NRC staff. 

9.3.4.20 Monitoring Procedures 

As noted in the NRC staff’s discussion in Section 9.3.3.1 of this SER and Section 6.1.2.2 of the 

applicant’s ER (IIFP,2011), IIFP uses written procedures to ensure that sampling and monitoring 

equipment is properly maintained; calibrated at regular intervals, including functional testing and 

routine checks; and is in good working condition. These procedures conform to guidance in 

Regulatory Guide 4.15 (NRC, 2007) and key parts of the REMP. The QC procedures include 

the use of established standards, such as those provided by NIST, and standard analytical 

procedures, such as those established by the NELAC. Samples are analyzed onsite for facility 
9-19 

related radiological constituents and may be shipped to a qualified independent laboratory for 

analyses. Laboratories will participate in third-party comparison studies to validate their 

performance. Examples of third-party programs include the Mixed Analyte Performance 

Evaluation Program and the DOE Quality Assurance Program. Section 11.8 of the applicant’s 

LA (IIFP, 2012a) also addresses, among other things, aspects of the Quality Assurance 

Program related to maintaining monitoring related records, as well as to developing and 

implementing monitoring-related procedures. 

The NRC staff finds that the applicant’s monitoring procedures conform to guidance found in 

Regulatory Guide 4.15 (NRC, 2007), because it employs nationally recognized analytical 

methods, instrumentation, laboratory validation methods, and laboratory procedures that are 

adequate to validate analytical results; and meet the applicable acceptance criteria in 

Section 9.4.3.2.2 of NUREG-1520, Revision 0 (NRC, 2002). Therefore, the NRC staff finds 

these procedures to be acceptable. 

9.3.4.21 Action Levels 

An action level is the concentration (or mass) of an analyte that indicates that some action 

needs to be taken, such as an investigation or, if the level is high enough, shutting down 

operations. Procedures specify action levels according to the type of samples and the specific 

analysis. The applicant stated that documented procedures include action levels for monitored 

environmental parameters, as appropriate, to provide guidance to ensure compliance with 

appropriate regulatory limits specified in 10 CFR Part 20, Subpart B. Documented procedures 

set response actions for elevated measurements at increasing levels of priority, ranging from 

(1) increasing monitoring frequency, (2) adjusting operations, and (3) performing corrective 

actions to prevent regulatory compliance levels from being exceeded. The applicant has 

designed administrative action levels to include, when practical, provision for automatic 

shutdown of the facility in the event action levels are exceeded. 

Administrative action levels for physiochemical monitoring, sampling protocols, and emissions 

and effluent monitoring are performed for routine operations with provisions for additional 

evaluation in response to potential accidental release. The applicant will implement action 

levels before facility operation to ensure that chemical discharges remain below the limits 

specified in the facility discharge permits. The applicant’s ESH organization staff selected the 

sampling and monitoring locations in accordance with facility permits and good sampling 

practices. 

The NRC staff finds that the applicant’s use of parameter action levels ensure that 

concentrations of radioactivity remain below the regulatory limits in 10 CFR Part 20 and meet 

the applicable acceptance criteria in Section 9.4.3.2.2 of NUREG-1520, Revision 0 

(NRC, 2002). Therefore, the NRC staff finds the use of these actions levels to be acceptable. 

9.3.4.22 Minimum Detectable Concentrations 

Table 6-5 of the applicant’s ER (IIFP, 2011) lists the MDCs for both effluent and environmental 

samples and identifies the medium (e.g., groundwater), activity (e.g., gross alpha, isotopic 

uranium), and typical MDCs (e.g., μCi/ml). The applicant will conduct physiochemical 

monitoring for required MDCs for fluoride and metals, among other things, via sampling of storm 

water, soil, sediment, vegetation, and groundwater, as noted in Table 6-3 of the applicant’s ER 

(IIFP, 2011). 
9-20 

The NRC staff finds that the MDCs are sufficient to meet action level, regulatory, and permit 

requirements, as well as the requirements of environmental media monitoring programs. The 

MDCs also meet the applicable acceptance criteria in Section 9.4.3.2.2 of NUREG-1520, 

Revision 0 (NRC, 2002), and are, therefore, acceptable to the NRC staff. 

9.3.4.23 Data Analysis 

Field and laboratory analytical procedures, discussed above, address the collection of 

representative radiologic and physiochemical samples, use of appropriate sampling methods 

and equipment, proper sampling locations, and proper handling and analysis of samples. 

The NRC staff finds that the data analysis methods and criteria that the applicant uses to 

evaluate and report environmental sampling results indicate when an action level is being 

approached in time to take corrective actions to ensure that concentrations remain within the 

regulatory limits in 10 CFR Part 20, Appendix B. The NRC staff finds that the applicant’s data 

analysis methods and criteria meet the applicable acceptance criteria in Section 9.4.3.2.2 of 

NUREG-1520, Revision 0 (NRC, 2002), and are, therefore, acceptable. 

9.3.4.24 Federal, State, and Local Requirements 

Section 1.4 of the applicant’s ER (IIFP, 2011) contains a complete description of required 

licenses, permits, and other approvals that are required by the Federal Government, as well as 

agencies of the State of New Mexico; Table 1-4 provides the status of each of these 

requirements. The applicant committed to follow applicable requirements for effluent monitoring 

activities described in the LA, Section 9.2.2 (IIFP, 2012a). 

Based on the information in Table 1-4 of the ER (IIFP, 2011), the NRC staff finds that the 

information provided satisfies the applicable acceptance criteria in Section 9.4.3.2.2 of 

NUREG-1520, Revision 0 (NRC, 2002), and is, therefore, acceptable. 

Measurement of Accidental Radioactive and Nonradioactive Releases to the Environment

In Chapter 3 of its LA (IIFP, 2012a), the applicant described its ISA, which evaluates potential 

risks and radiological and nonradiological (e.g., chemical) hazards from postulated unmitigated 

accident scenarios that could result in injuries to workers and the public or in significant 

environmental impacts. The applicant’s ISA Summary (IIFP, 2012b) documents assessments of 

these accidents. The ISA Summary (IIFP, 2012b) also identifies active- and passiveengineered IROFS or administrative IROFS that prevent or mitigate the likelihood and 

consequences of those accident scenarios to acceptable levels. The applicant also addresses 

environmental effects of accidents in Section 4.12.3 of its ER (IIFP, 2011). 

Because the IIFP facility only processes depleted uranium, the applicant mentioned that no 

nuclear criticality potential exists, and no materials that contain fission products or transuranic 

elements will be located onsite. Thus, the applicant stated that offsite radiological 

consequences associated with the facility are limited. Large inventories of depleted uranium 

material are present onsite. Despite these large inventories, the applicant has identified no 

credible accident to pose intermediate- or high-radiological consequences to the public. 

However, the applicant recognizes credible, intermediate-chemical consequence events could 

result from potential uranium oxide releases because of the acute chemical, as opposed to 

radiological, exposure of the uranium material. In addition, the applicant identified credible 

release scenarios that could result in both intermediate- and high-offsite consequences 
9-21 

associated with acute chemical exposure. Thus, the only high- or intermediate consequence 

events postulated by the applicant are chemical in nature and involve the release of HF, either 

as a direct release of HF or as a release of a fluoride-bearing compound and subsequent 

release of HF as a reaction byproduct. However, these intermediate- or high-consequence 

events have a low probability of occurring. 

The postulated credible accidents identified in the ISA Summary result from process upset 

conditions (e.g., loss of process and safety controls), natural phenomena (e.g., seismic events 

and tornados), and fire (IIFP, 2012b). Because of a combination of safety prevention limits and 

controls and mitigation measures, a significant process upset condition is not expected to result 

in the mitigated release of radiological or hazardous chemical material that would result in 

intermediate or high consequences to the public. Also, given the bounding of seismic events 

and tornados at the facility, as well as the preventive and mitigation methods employed if a 

significant fire occurs at the facility, a mitigated release of radiological or hazardous chemical 

material is not expected to result in intermediate or high consequences to the public. 

Table 3-1 of the applicant’s LA (IIFP, 2012a) identifies, among other things, consequence 

descriptions to the offsite public and the environment for high- and intermediate-severity 

accidents. Table 3-2 identifies EPA acute exposure guideline levels for uranium hexafluoride 

(UF6), HF, and soluble uranium. The applicant also addressed radiological and non-radiological 

accident analyses, including mitigation measures to attenuate releases to the environment, in 

Section 4.12.3 of its ER (IIFP, 2011). As noted above, Chapter 3 of this SER presents the NRC 

staff’s evaluation of these hazards. 

As noted above in Sections 9.3.3.1, 9.3.3.2, and 9.3.3.3 of this SER, the applicant’s monitoring 

program provides for effluent and environmental monitoring of airborne and liquid releases of 

uranium, UF6, and HF. For example, air emissions are monitored continuously for uranium and 

HF. In addition, ambient air is monitored for activities from the UF6 cylinder pads. Active air 

monitors (dosimeters) are also placed around the restricted area fence line for analysis of gross 

alpha activity and concentrations of uranium isotopes. 

The NRC staff finds that the applicant’s environmental monitoring program is adequate to 

assess impacts to the environment from potential radioactive and nonradioactive releases, as 

identified in high- and intermediate consequence accident sequences discussed in the ISA; 

adequately addresses related performance requirements in 10 CFR 70.61, “Performance 

Requirements,” for individuals located outside the controlled area; and meets the applicable 

acceptance criteria in Section 9.4.3.2.2 of NUREG-1520, Revision 0 (NRC, 2002). Therefore, 

the NRC staff finds the applicant’s approach to environmental monitoring to be acceptable. 

9.3.5 Integrated Safety Analysis Summary 

In Chapter 3 of this SER, the NRC staff provides its evaluation of the ISA Summary and 

documents its conclusion that the ISA Summary (IIFP, 2012b) is complete, provides reasonable 

estimates of the likelihood and consequences of each accident sequence, and provides 

sufficient information to determine whether adequate engineering or administrative controls are 

identified for each accident sequence. Chapter 11 of this SER contains the NRC staff’s 

evaluation of management measures used to ensure that IROFS perform their intended safety 

functions. The NRC staff verified that environmental release limits would be met using existing 

IROFS. Therefore, no additional IROFS are identified for the proposed facility for reducing the 

environmental risks of natural phenomena and potential accidents. 
9-22 

Under 10 CFR Part 70, Subpart H, “Additional Requirements for Certain Licensees Authorized 

To Possess a Critical Mass of Special Nuclear Materials (10 CFR 70.60 through 10 CFR 70.76), 

an applicant is to ensure, among other things, compliance with various performance 

requirements. Title 10 CFR 70.61(c)(3) requires that the applicant apply controls such that a 

credible intermediate-consequence event is unlikely to occur or that the consequence of such 

an event does not exceed a 24-hour averaged release of radioactive material outside the 

restricted area in concentrations 5,000 times the values in Table 2 of Appendix B to 

10 CFR Part 20. 

In its ISA Summary (IIFP, 2012b), the applicant identified various sequences for radiological and 

non-radiological accidents which it evaluated to ensure adequate protection of worker health 

and safety. By ensuring that all credible, high-consequence events are rendered highly unlikely 

and that all intermediate-consequence events are unlikely, the applicant also ensured that the 

environmental performance requirements of 10 CFR 70.61(c)(3) are met. The NRC staff 

determined that offsite environmental consequences could occur only if uncontrolled, 

intermediate or high consequences to workers were also present. The NRC staff did not identify 

any accident sequence that would fail to meet the environmental performance requirements of 

10 CFR 70.61. 

9.3.6 10 CFR 70.61(c)(3) 

The applicant stated that risk reduction is accomplished through a combination of preventive 

and mitigative measures, with an emphasis on preventive measures. Chapter 3 of this SER, 

which addresses accident sequences for intermediate- and high-consequence events, provides 

a more complete discussion. Chapter 3 also addresses preventive and mitigative measures. 

The NRC staff finds that the applicant’s ISA Summary complies with 10 CFR Part 70 because it 

meets the performance requirements of 10 CFR 70.61. The applicant’s ISA Summary meets 

the applicable acceptance criteria in Section 9.4.3.2.3 of NUREG-1520, Revision 0 

(NRC, 2002), as well and is, therefore, acceptable. 

9.4 Evaluation Findings

The applicant has developed a program to implement adequate environmental protection 

measures during operation. These measures include (1) environmental and effluent monitoring 

and (2) effluent controls to maintain public doses ALARA as part of the radiation protection 

program. The NRC staff concludes that the applicant’s program, as described in its application, 

is adequate to protect the environment and the health and safety of the public and complies with 

regulatory requirements imposed by the Commission in 10 CFR Part 20, 10 CFR Part 40, 

10 CFR Part 51, and 10 CFR Part 70. 

The NRC staff will issue a final Environmental Impact Statement (EIS) as part of this licensing 

action, as required by 10 CFR 51.20, “Criteria for and Identification of Licensing and Regulatory 

Actions Requiring Environmental Impact Statements.” The final EIS will consider the 

environmental impacts of the construction, operation, and decommissioning of the proposed 

facility and compare alternatives, which will inform the NRC staff’s recommendation concerning 

the LA for the proposed facility. 
9-23 

9.5 References

(EPA, 1988) U.S. Environmental Protection Agency, “Limiting Values of Radionuclide Intake and 

Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion,” 

Federal Guidance Report No. 11, EPA- 520/1-88-020, Washington, D.C., September 1988. 

(EPA, 1993) U.S. Environmental Protection Agency, Federal Guidance Report No. 12, External 

Exposure to Radionuclides in Air, Water, and Soil, Washington, D.C., 1993. 

(ICRP, 1996) International Commission on Radiological Protection, “Age-dependent Doses to 

the Members of the Public from Intake of Radionuclides - Part 5 Compilation of Ingestion and 

Inhalation Coefficients,” Publication 72, 1996. 

(IIFP, 2012a) International Isotopes Fluorine Products, Inc., “Fluorine Extraction Process and 

Depleted Uranium Deconversion Plant (FEP/DUP) License Application, Revision B,” May 2012, 

Agencywide Documents Access and Management System (ADAMS) Accession 

No. ML12123A245. 

(IIFP, 2012b) International Isotopes Fluorine Products, Inc., “ISA Summary Rev. B for IIFP,” 

May 2012, Agencywide Documents Access and Management System (ADAMS) Accession 

No. ML12123A245. 

(IIFP, 2011) International Isotopes Fluorine Products, Inc., “Fluorine Extraction Process and 

Depleted Uranium Deconversion Plant (FEP/DUP) Environmental Report,” Rev. A, 2011, 

Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML100120758. 

(NRC, 2010) U.S. Nuclear Regulatory Commission, “Monitoring and Reporting Radioactivity in 

Releases of Radioactive Materials in Liquid and Gaseous Effluents from Nuclear Fuel 

Processing and Fabrication Plants and Uranium Hexafluoride Production Plants,” Regulatory 

Guide 4.16, Revision 2, December 2010. 

(NRC, 2008) U.S. Nuclear Regulatory Commission, “Minimization of Contamination and 

Radioactive Waste Generation: Life-Cycle Planning,” Regulatory Guide 4.21, June 2008. 

(NRC, 2007) U.S. Nuclear Regulatory Commission, “Quality Assurance for Radiological 

Monitoring Programs (Inception through Normal Operations to License Termination) - Effluent 

Streams and the Environment,” Regulatory Guide 4.15, 2007. 

(NRC, 2002) U.S. Nuclear Regulatory Commission, “Standard Review Plan for the Review of a 

License Application for a Fuel Cycle Facility,” NUREG-1520, Rev. 0, March, 2002. 

(NRC, 1996) U.S. Nuclear Regulatory Commission, “Constraint on Releases of Airborne 

Radioactive Material to the Environment for Licensees Other Than Power Reactors,” Regulatory 

Guide 4.20, December 1996. 

(NRC, 1993) U.S. Nuclear Regulatory Commission, “ALARA Levels for Effluents from Materials 

Facilities,” Regulatory Guide 8.37, July 1993. 
9-24 

(NRC, 1977) U.S. Nuclear Regulatory Commission, “Calculation of Annual Doses to Man from 

Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 

10 CFR Part 50, Appendix I,” Regulatory Guide 1.109, Revision 1, October 1977. 

(ORNL, 2000) Oak Ridge National Laboratory, “MCNP4C Monte Carlo N-Particle Transport 

Code System,” CCC-700 MCNP4C2, RSICC Computer Code Collection, Oak Ridge, TN, 2000. 

http://www.nea.fr/abs/html/ccc-0740.html 
10-1 

10.0 DECOMMISSIONING 

The purpose of the U.S. Nuclear Regulatory Commission’s (NRC’s) review of the International 

Isotopes Fluorine Products, Inc. (IIFP or the applicant), Decommissioning Plan (DP) is to 

evaluate whether the application provides for decommissioning the facility safely and in 

accordance with NRC requirements. 

At the time of the initial license application (LA) for a uranium hexafluoride deconversion facility, 

the applicant is required to submit a Decommissioning Funding Plan (DFP). The purpose of 

NRC’s review of the DFP is to determine whether the applicant has considered 

decommissioning activities that may be needed in the future, has performed a credible 

site-specific cost estimate for those activities, and has presented the NRC with financial 

assurance to cover the cost of those activities in the future. The DFP, therefore, should contain 

an overview of the applicant’s proposed decommissioning activities, the methods used to 

determine the cost estimate, and the financial assurance mechanism. This overview should 

contain sufficient detail to enable the reviewer to determine whether the Decommissioning Cost 

Estimate (DCE) is reasonably accurate. 

10.1 Regulatory Requirements 

The following NRC regulations require planning, financial assurance, and recordkeeping for 

decommissioning, as well as procedures and activities to minimize waste and contamination: 

• Title 10 of the Code of Federal Regulations (10 CFR) Part 20, “Standards for Protection 

against Radiation,” Subpart E, “Radiological Criteria for License Termination” 

• 10 CFR 40.36, “Financial Assurance and Recordkeeping for Decommissioning” 

• 10 CFR 40.42, “Expiration and Termination of Licenses and Decommissioning of Sites 

and Separate Buildings or Outdoor Areas” 

10.2 Regulatory Guidance and Acceptance Criteria 

Volume 3, ”Financial Assurance, Recordkeeping, and Timeliness” of NUREG-1757, 

“Consolidated NMSS Decommissioning Guidance,” issued September 2003 (NRC, 2003), 

contains the guidance applicable to NRC’s review of the decommissioning section of the LA 

(IIFP, 2012a); and provides guidance for developing final DPs required under 10 CFR Part 20, 

Subpart E, and 10 CFR 40.42(g). The applicant will submit a final DP to the NRC at the time of 

decommissioning. At the time of initial licensing, as well as for license renewals, only an 

overview of the proposed decommissioning activities is needed to develop the DFP. This 

overview is a more generalized discussion of the detailed information that would be needed for 

the final DP described in NUREG-1757, Volume 3 (NRC, 2003). 

Consistent with the guidance in NUREG-1757, Volume 3 (NRC, 2003), the application should 

include the following information: 

• an overview of the proposed decommissioning activities that contains sufficient detail to 

enable the reviewer to determine whether the DCE is reasonably accurate 

• proposed plans for minimizing contamination 
10-2 

• proposed plans for meeting the recordkeeping requirements of 10 CFR 40.36(f) 

• proposed procedures to protect the health and safety of workers, the public, and the 

environment during decommissioning 

• proposed environmental protection measures, specifically, the commitments to minimize 

waste associated with decommissioning 

• proposed radiation protection program as it applies to radiological decontamination and 

the management of radiological effluents. 

The overview of the decommissioning activities should cover the conceptual decontamination 

and DP, including the decommissioning program and steps, management and organization, 

health and safety, radiological decommissioning criteria, waste management, security and 

nuclear material control, recordkeeping, the decontamination process, and the minimization of 

contamination. 

The following sections discuss these criteria and the applicant’s information that addresses 

them. 

10.3 Staff Review and Analysis

The NRC staff’s review of the DFP focused on the applicant’s conceptual decontamination and 

DP for the deconversion facility, the DCEs, and the financial assurance for decommissioning 

activities. The applicant identified the decommissioning activities that may be needed in the 

future for decommissioning and presented site-specific estimates of decommissioning costs for 

those activities. Using the cost data as a basis, the applicant stated that it has presented 

financial assurance to cover the costs required to release the deconversion facility site for 

unrestricted use. The following subsections discuss these decommissioning aspects, as 

described by the applicant, and the NRC staff’s assessment of the applicant’s proposed DP, 

cost estimates, and funding plan. Before license termination, the applicant will provide a 

detailed DP that will include specific activities which it will use to protect workers, the public, and 

the environment. 

Chapter 10 of the LA presents the IIFP DFP for the proposed facility. The applicant developed 

the DFP following the guidance provided in NUREG-1757, Volume 1, Revision 2, 

“Decommissioning Process for Materials Licensees,” issued September 2006 (NRC, 2006). 

IIFP committed to decontaminate and decommission the facility at the end of its operation so 

that the facility and grounds can be released for unrestricted use. In accordance with 

10 CFR 40.42(d), the IIFP will review and update the DFP as necessary every year for the first 

3 years of operation and at least once every 3 years thereafter. Before facility decommissioning, the applicant will prepare a DP in accordance with 10 CFR 40.42, “Expiration and 

Termination of Licenses and Decommissioning of Sites and Separate Buildings or Outdoor 

Areas,” and submit it to the NRC for approval. 
10-3 

10.3.1 Conceptual Decontamination and Decommissioning Plan 

The DFP contains an overview of the proposed decommissioning activities. The applicant will 

provide a detailed DP to the NRC at the time of decommissioning, in accordance with 

10 CFR Part 20, Subpart E; 10 CFR 40.36, “Financial Assurance and Recordkeeping for 

Decommissioning”; and 10 CFR 40.42. This section also addresses the recordkeeping 

requirements in 10 CFR 40.36(f). 

The NRC staff reviewed the proposed conceptual decontamination and DP and determined that 

it contains commitments to meet the radiological criteria for unrestricted use contained in 

10 CFR 20.1402, “Radiological Criteria for Unrestricted Use,” and to meet the timeliness in 

decommissioning requirements in 10 CFR 40.42(h). Therefore, as discussed below and based 

upon the applicant’s commitments, the NRC staff finds the applicant’s conceptual 

decontamination and DP acceptable. 

10.3.2 Decommissioning Strategy 

The overall strategy for decommissioning is to decontaminate or remove all materials from the 

site so that the facility can be released and the site made available for unrestricted use. At the 

end of useful plant life, the Fluorine Extraction Process and Depleted Uranium Deconversion 

Plant (FEP/DUP) will be decommissioned such that the site and remaining facilities may be 

released for unrestricted use, as defined in 10 CFR 20.1402. The NRC staff finds that IIFP 

committed to the use of guidance provided in NUREG-1505, Revision 1, “A Nonparametric 

Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys - 

Interim Draft Report for Comment and Use,” issued in 1998 (NRC, 1998); NUREG-1575, 

“Multi-Agency Radiation Survey and Site Investigation Manual (MARSIMM),” issued August 

2000 (NRC, 2000); and NUREG-1757, Revision 2 (NRC, 2006), in developing initial and final 

site survey plans. These three guidance documents describe methodologies that NRC staff 

have found acceptable for implementing the Commission’s decommissioning regulations in 

10 CFR 20 Subpart E. The NRC staff finds that IIFP’s commitment to follow them is sufficient to 

provide background and post decontaminationsite conditions to enable the free release of the 

site. 

The applicant stated that it will decontaminate all remaining facilities where needed to 

acceptable levels for unrestricted use. Hazardous wastes will be treated or disposed of in 

licensed hazardous waste facilities. Disposal of radioactive or hazardous material will not occur 

at the plant site, but at licensed facilities located elsewhere. Following decommissioning, the 

facilities and site will be available for reuse. 

Financial arrangements are made to cover costs required for returning the IIFP facility to 

unrestricted use. The applicant stated it will provide updates on cost and funding to the NRC 

every year for the first 3 years of operations, and then every 3 years thereafter, as described 

above. In addition, IIFP will submit a detailed, updated DP at a date near the end of plant life, in 

accordance with 10 CFR 40.42. 
10-4 

10.3.3 IIFP Facility Description 

Chapter 1 of the LA (IIFP, 2012a) describes the IIFP FEP/DUP and site. The Integrated Safety 

Analysis (ISA) Summary (IIFP, 2012b) presents a detailed description of the plant site and 

facility and the safety aspects of the plant processes. Chapter 1 of the LA also describes the 

specific quantities and types of licensed materials used at the facility (see section entitled, 

“Institutional Information”). Also in Chapter 1 of the LA is a general description of how licensed 

materials are used at the facility (see section entitled, “General Information”). 

10.3.4 Design Features 

Section 10.1.2 of the LA describes design features incorporated into the plant’s initial design 

that will simplify eventual dismantling and decontamination. The plans are implemented through 

proper management and health and safety programs. Decommissioning policies address 

radioactive waste management, radioactive contamination control, physical security, and 

material control. 

Section 10.1.2.2 of the LA describes major features incorporated into the facility design to 

facilitate decontamination and decommissioning. These features include the following: 

• Building areas where uranium is processed and handled are separated physically from other 

building rooms and areas in which there is no need to have uranium present. These areas 

have separate ventilation and filtration systems to preclude contamination spread. 

Boundary control stations and hand/foot and portable monitors are used at applicable 

locations to verify that personnel and items exiting uranium process areas are not spreading 

radiological materials into nonuranium areas. The depleted uranium tetrafluoride (DUF4) 

process building, fluoride extraction process (FEP) oxide staging building, plant operations 

decontamination building, DUF4 container storage building, DUF4 container staging building, 

and the FEP process building (in areas where licensed material is processed) meet these 

specific design features. 

• All areas of the plant are sectioned into unrestricted and restricted areas. All procedures for 

these areas fall under the radiation protection program, and serve to minimize the spread of 

contamination and simplify the eventual decommissioning. 

• The applicant will conduct routine radiological surveys throughout the operating lifetime of 

the facility to minimize the likelihood that radioactive contamination goes undetected and to 

provide a historical record which will simplify the site characterization process. The 

historical data collected by the Radiological Environmental Monitoring Program (REMP), 

described in Section 6.1.2 of the Environmental Report (IIFP, 2011d), will also be used to 

provide guidance for the final site survey. Samples will be collected to verify the historical 

data obtained via the REMP. Additional samples will be collected if areas of contamination 

are found. 

• Non-radioactive process equipment and systems are minimized in locations subject to 

potential contamination. This limits the size of the restricted areas and limits the activities 

occurring inside these areas. 

• Local air filtration is provided for areas with potential airborne contamination to preclude its 

spread. Containment equipment with hoods that exhaust through dust collectors, which are 
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designed with high-removal efficiencies, are used where uranium materials are being 

packaged or withdrawn from process systems. 

• The hazardous material processes include designs for purge and evacuation (P&E) systems 

and dust collection equipment as a means to provide effective cleanout of residual 

chemicals or dust from equipment or piping before opening systems for maintenance. The 

P&E and dust collector systems have multiple collection equipment in series 

(defense-in-depth) to ensure removal and treatment efficiency, redundancy, effectiveness, 

and reliability. 

• Storm water runoff via the plant storm sewer system flows to a retention basin for 

evaporation, landscape watering, or discharge. Before discharging, collected storm water 

can be sampled if needed. The applicant stated domestic sanitary waste water is tertiary 

treated to meet all discharge standards and is either evaporated or used as harvested water 

for facility trees, grass, and shrubs. The facility is designed for no liquid process water 

discharges. Engineered systems are used to provide for regeneration of scrubbing solutions 

and recycle within the process systems. 

10.3.5 Worker Exposure and Waste Volume Control 

In Sections 9.2.1.4 and 10.1.2.3 of the LA (IIFP, 2012a), the applicant described features that 

serve to minimize worker exposure to radiation and minimize radioactive waste volumes during 

decontamination activities. As a result, the spread of contamination is minimized as well. 

These features include the following: 

• During construction, the applicant will apply a washable coating to designated floors and 

walls in the restricted areas that have the higher potential to become radioactively 

contaminated during operation. The coating will serve to lower waste volumes during 

decontamination and simplify the decontamination process. 

• Sealed, nonporous pipe insulation will be used in areas with higher potential to become 

contaminated. This will facilitate cleaning in the event of a spill and will reduce waste 

volume during decommissioning. 

• Ample access will be provided for efficient equipment dismantling and removal of equipment 

that may be contaminated. This minimizes the time of worker exposure. 

• Tanks will have access for entry and decontamination. Design provisions will also be made 

to allow complete draining of the wastes contained in the tanks. 

• Connections in the process systems, provided for required operation and maintenance, will 

allow for thorough purging at plant shutdown. This will remove a significant portion of 

radioactive contamination before disassembly. 

• Design drawings, produced for all areas of the plant, will simplify planning and implementing 

decontamination procedures. This in turn will shorten the durations that workers are 

exposed to radiation. 

• Worker access to contaminated areas will be controlled to ensure that workers wear proper 

protective equipment and limit their time in the areas. 
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• Radioactive and hazardous wastes produced during decommissioning will be collected, 

handled, and disposed of in accordance with regulations applicable to the facility at the time 

of decommissioning. Generally, procedures will be similar to those described for wastes 

produced during normal operation. These wastes will ultimately be disposed in licensed 

radioactive or hazardous waste disposal facilities located elsewhere. Nonhazardous and 

nonradioactive wastes will be disposed in a manner consistent with good industrial practice 

and in accordance with applicable regulations. 

• To facilitate decommissioning, the information relating to the facility design, facility 

construction, design modifications, site conditions before and after construction, onsite 

contamination, and results of monitoring and radiological surveys will be readily recoverable 

through the IIFP document control and management process. 

10.3.6 Management Organization 

Section 10.1.2.4 of the LA (IIFP, 2012a) describes an appropriate organizational structure that 

the applicant will develop to support the decommissioning strategy. The applicant stated that 

the organizational structure will ensure that adequate numbers of experienced and 

knowledgeable personnel are available to perform the technical and administrative tasks 

required to decommission the facility. 

10.3.7 Health and Safety 

In Section 10.1.2.5 of the LA (IIFP, 2012a), the applicant stated that the policy during 

decommissioning will be to keep individual and collective occupational radiation exposure as 

low as reasonably achievable (ALARA). A health physics program will identify and control 

sources of radiation, establish worker protection requirements, and direct the use of survey and 

monitoring instruments. 

10.3.8 Waste Management 

In Section 10.1.2.6 of the LA (IIFP, 2012a), the applicant stated that radioactive and hazardous 

wastes produced during decommissioning will be collected, handled, and disposed of in 

accordance with all regulations applicable to the facility at the time of decommissioning. 

Generally, procedures will be similar to those described for wastes produced during normal 

operation. These wastes will ultimately be disposed of in licensed radioactive or hazardous 

waste disposal facilities located elsewhere. 

10.3.9 Physical Security and Material Control and Accounting 

In Section 10.1.2.7 of the LA (IIFP, 2012a), the applicant stated that it will maintain physical 

security and material control and accounting as required during decommissioning in a manner 

similar to the programs in force during operation. The IIFP plan for completion of 

decommissioning, submitted near the end of plant life, will describe any necessary revisions to 

these programs. 
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10.3.10 Recordkeeping 

In Section 10.1.2.8 of the LA (IIFP, 2012a), the applicant stated that records important for safe 

and effective decommissioning of the facility will be stored in the FEP/DUP records 

management system until the site is released for unrestricted use. Information maintained in 

these records includes the following: 

(1) Records of spills or other unusual occurrences involving the spread of contamination 

and cleanup around the facility, equipment, or site will be maintained. These records will 

include any known information identifying the involved nuclides; their quantities, forms, 

and concentrations; and survey results after cleanup of any spill area. 

(2) Routine radiological survey records of restricted and unrestricted areas will be retained 

indefinitely to support historical site assessment and facility characterization at the time 

of decommissioning. 

(3) As-built drawings and modifications of structures and equipment in restricted areas will 

be maintained where radioactive materials are used or stored. Required drawings will 

be referenced as necessary, although each relevant document will not be indexed 

individually. If drawings are not available, appropriate records of available information 

concerning these areas and locations will be substituted. 

(4) The following will be contained in a single records document, updated every 2 years, 

except for areas containing only sealed sources: 

• all areas designated and formerly designated as restricted areas as defined 

under 10 CFR 20.1003, “Definitions” 

• all areas outside of restricted areas that require documentation specified in 

item (1) above 

• all areas outside of restricted areas where current and previous wastes have 

been buried will be documented under 10 CFR 20.2108, “Records of Waste 

Disposal” 

• all areas outside of restricted areas that contain material such that, if the license 

expired, the licensee would be required to either decontaminate the area to meet 

the criteria for decommissioning in 10 CFR Part 20, Subpart E, or apply for 

approval for disposal under 10 CFR 20.2002, “Method for Obtaining Approval of 

Proposed Disposal Procedures.” 

(5) Records of the cost estimate performed for the DFP or of the amount certified for 

decommissioning and records of the funding method used for assuring funds, if either a 

funding plan or certification is used, will be maintained. 

10.3.11 Decommissioning Process 

Section 10.1.3 of the LA (IIFP, 2012a) describes the IIFP’s decommissioning process. 

Preparation for decommissioning is expected to begin for the facility upon a decision to cease 

operations permanently; and this preparation step is expected to be completed in approximately 
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1 year, including NRC’s review and approval of the final plan. Actual decontamination and 

decommissioning would follow shortly after approval of the plan and the award of any 

subcontracts. Figure 10-1 of the LA (IIFP, 2012a) illustrates the DP schedule for the IIFP 

facility. 

Before completely shutting down all processes, the bulk work-in-process (WIP) inventory of 

uranium materials would be processed as much as practical into depleted uranium oxide and 

the fluoride gas products, similar to normal operations. This activity would render the bulk 

materials into products for shipment to customers and into depleted uranium oxide approved for 

disposal, similar to normal plant operations. Based on the estimated maximum-average WIP 

inventories, the amount of time required to orderly process this material into its final form is 

between 12 and 15 days. After processing the bulk WIP, any residual inventory of uranium or 

contaminated materials would be included in the decommissioning steps that follow the 

decommissioning preparation and NRC approvals to proceed. As shown in Table 10-1 of the 

LA (IIFP, 2012a), the estimated residual amounts of uranium chemicals or 

uranium-contaminated chemicals expected to be disposed as low-level radioactive waste (LLW) 

are approximately 48 cubic m (1,700 cubic ft). 

Before the start of decommissioning operations, the applicant will perform a radiological survey 

of the facility in conjunction with a historical site assessment. The final DP submitted to the 

NRC will present the findings of the radiological survey and historical site assessment. The 

applicant will prepare this DP in accordance with 10 CFR 40.42 and the applicable guidance 

provided in NUREG-1757, Volume 3 (NRC, 2003). 

Decommissioning activities will include (1) outfitting of size reduction and packaging areas, 

(2) purging of process systems, (3) dismantling and removal of equipment, (4) sales of salvaged 

materials, (5) packaging and disposal of wastes, and (6) completion of a final radiation survey. 

Credit is not taken for any salvage value that might be realized from the sale of potential assets 

during or after decommissioning. 

Using the IIFP approach, 10 CFR 40.42(k) and 10 CFR 20.1402 requires residual radioactivity 

to be reduced below specified levels before the facility may be released for unrestricted use. 

Current NRC guidelines for release serve as the basis for the decontamination costs estimated 

in Chapter 10 of the LA (IIFP, 2012a). Portions of the facility that do not exceed contamination 

limits may remain as is without further decontamination measures applied. The intent of 

decommissioning the facility is to remove all uranium process-related equipment from the 

buildings, such that only the building shells and site infrastructure remain. The removed 

equipment includes all piping and components from systems providing, for example, uranium 

hexafluoride (UF6) or uranium tetrafluoride (UF4) containment; uranium oxide containment; 

systems in direct support of uranium processing (such as refrigerant and chilled water); 

radioactive and hazardous waste-handling systems; and contaminated heating, ventilation, and 

air-conditioning (HVAC) filtration systems. The site infrastructure remaining after 

decommissioning is complete will include the steam facilities, electrical power facilities, water 

supply systems, sanitary water treatment systems, fire protection systems, HVAC systems, 

cooling water systems, and communication systems. 

The applicant will outfit existing plant buildings, such as the decontamination building and 

material warehouse, to accommodate handling and packaging of components and materials for 

disposal. These areas will be the primary location for size reduction and packaging activities 

during the decommissioning process. Limited capabilities for decontamination will exist for 
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mildly contaminated items that may be decontaminated to free release criteria in a cost-effective 

manner. 

The applicant will decontaminate contaminated portions of the buildings as required. Potential 

contamination is limited to the structures in the restricted areas. Good housekeeping practices 

during normal operation will keep the other areas of the site clean, and routine radiological 

contamination surveys will ensure that radioactive contamination will not go undetected or be 

allowed to buildup to levels difficult to control. When decontamination is complete, the applicant 

will survey all areas and facilities on the site to verify that further decontamination is not 

required. Decontamination activities will continue until the entire site is demonstrated to be 

suitable for unrestricted use. 

10.3.12 Size Reduction and Packaging Facility Outfitting 

The NRC staff confirmed that the applicant’s facility description that IIFP’s proposed facilities 

can be adapted to accommodate the size reduction and packaging activities associated with 

decommissioning. IIFP identified the decontamination building and material warehouse as 

suitable for these purposes. IIFP estimated time for equipment installation as approximately 

2 months. This timeframe supports the dismantling of the equipment. Section 10.1.4.3 of the LA 

(IIFP, 2012a) describes these facilities. 

10.3.13 System Preparation 

At the end of the useful life of each process line, the uranium process will be shut down and 

UF6, UF4, and uranium oxides will be removed to the extent practicable by normal process 

operation. This will be followed by evacuation and purging with nitrogen and the application of a 

fixative, where applicable. The shutdown and preparation of the decommissioning process is 

estimated to take approximately 3 months. 

10.3.14 Dismantling 

Dismantling requires cutting and disconnecting all components requiring removal. Dismantling 

operations are labor intensive and generally require the use of protective clothing. The work 

process will be optimized, considering the following: 

• minimizing the spread of contamination and the level of protective clothing necessary using 

fixative coatings 

• balancing the number of cutting and removal operations with the resultant size reduction 

and disposal requirements 

• optimizing the rate of dismantling with the rate of size reduction of facility throughput 

• providing storage and laydown space required, as impacted by retrieving, security, and 

other activities 

• balancing the cost of salvage with the cost of disposal. 

The applicant will decide the details of the complex optimization process near the end of plant 

life, taking into account specific contamination levels, market conditions, and available waste 
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disposal sites. The DFP assumes that most items that were continuously in contact with UF6, 

UF4, or uranium oxide will be disposed of at a low-level radioactive waste disposal facility rather 

than employing rigorous decontamination techniques. Large contaminated components may be 

disassembled to separate contaminated and uncontaminated portions of the component. To 

avoid laydown space and contamination problems, dismantling will generally not be allowed to 

proceed faster than the downstream size reduction and packaging process. 

The timeframe to accomplish both dismantling and size reduction at FEP/DUP is estimated to 

be approximately 18 months. 

10.3.15 Decontamination and Size Reduction 

Section 10.1.4 of the LA (IIFP, 2012a) describes in detail the decontamination and size 

reduction process. The description encompasses the decontamination and size reduction 

methodology, size reduction and packaging facilities, decontamination and size reduction 

procedures, and results. 

10.3.16 Salvage of Equipment and Materials 

Items to be removed from the facilities can be categorized as potentially reusable equipment, 

recoverable scrap, and wastes. However, based on a 40-year license, or beyond, operating 

equipment is assumed to have no reuse value. Wastes will also have no salvage value. With 

respect to scrap, some amounts of uncontaminated metal (steel, copper, Monel) may be 

recovered and sold. Contaminated materials will be disposed of as low-level radioactive waste. 

No credit is taken for any salvage value that might be realized from the sale of potential assets 

during or after decommissioning. 

10.3.17 Disposal 

All wastes produced during decommissioning will be collected, handled, and disposed of in a 

manner similar to that described for wastes produced during normal operation. Wastes will 

consist of normal industrial trash, nonhazardous chemicals and fluids, small amounts of 

hazardous materials, and radioactive wastes. The radioactive waste will consist primarily of 

piping, tanks, hoppers, and compactable trash generated during the dismantling process. 

Radioactive wastes will ultimately be disposed of in licensed low-level radioactive waste 

disposal facilities. Hazardous wastes will be disposed of in hazardous waste disposal facilities. 

Nonhazardous and nonradioactive wastes will be disposed of in a manner consistent with good 

industrial practice and in accordance with all applicable regulations. The DP, which the 

applicant will submit to the NRC before initiating the decommissioning of the plant will provide a 

complete estimate of the wastes and effluent to be produced during decommissioning. 

10.3.18 Final Radiation Survey 

At the end of useful plant life, the FEP/DUP will be decommissioned such that the site and 

remaining facilities may be released for unrestricted use as defined in 10 CFR 20.1402. IIFP 

stated that it will use guidance provided in NUREG-1505 (NRC, 1998), NUREG-1575 

(NRC, 2000), and NUREG-1757, Volume 3 (NRC, 2003), in developing initial and final site 

survey plans capable of providing sufficient data on the background and post 

decontaminationsite conditions to enable the free release of the site. 
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A final radiation survey will be performed to verify proper decontamination to allow the site to be 

released for unrestricted use. The evaluation of the final radiation survey is based in part on an 

initial radiation survey performed before initial operation. Since the IIFP facility will only process 

depleted uranium, the initial and final site surveys required as part of the decommissioning 

process will include isotopic analysis for the uranium-238, uranium-235, and uranium-234 

isotopes of uranium. The final survey will systematically and representatively measure 

radioactivity for the entire site and will be designed to detect any unreported spills and any 

generalized contamination that might accumulate over the period of operation of the facility. 

The intensity of the survey will vary depending on the location (i.e., the buildings, their 

immediate areas, and the remainder of the site). 

Throughout the operating life of the facility, the applicant will conduct routine surveys of licensed 

material areas and maintain records of such surveys. These survey records will be used in 

conjunction with the REMP as part of the final survey evaluation and may reduce the amount of 

sampling in some areas for which the survey history indicates that no contamination has 

occurred. The applicant will document the survey procedures and results in a report which will 

include, among other things, a map of the survey site, measurement results, and the site’s 

relationship to the surrounding area. The results will be analyzed and shown to be below 

allowable residual radioactivity limits; otherwise, IIFP will perform further decontamination. 

For decommissioning funding purposes, the final site survey will consist of samples taken within 

the 40-acre IIFP restricted area, as well as at other locations outside the restricted area but 

within the 640-acre site boundary. Inside the restricted area, samples will be taken based on a 

sampling grid pattern of approximately 91 m by 91 m (100 yards by 100 yards). Additional 

samples will be collected within an area extending 3 m (10 ft) from process building walls on the 

basis of one sample per 9.3 square m (100 square ft) (i.e., one sample for every 3 m [10 ft] of 

building perimeter). It is unlikely that the area outside the restricted area, but within the site 

boundary, will be contaminated. Outside of the restricted area, but within the site boundary, the 

likelihood for contamination is extremely remote. Therefore, the grid will be expanded for this 

area such that samples will be taken on a grid pattern of approximately 610 m by 610 m 

(1,017 ft by 1,017 ft). A third part will provide the analysis of the samples since, at the time of 

performance of the final radiation survey, no analysis facilities will be available onsite. As part 

of the REMP, the applicant will perform a similar collection of samples for the initial site survey 

to provide a background value against which to compare the final site survey. 

10.3.19 Decommissioning Impact on Integrated Safety Analysis 

Although decommissioning steps are planned to be underway while some activities considered 

in the ISA continue to occur in the other areas of the plant, the current ISA does not fully 

evaluate these decommissioning risks. The applicant will perform an updated ISA before 

decommissioning activities begin to evaluate the risks from decommissioning operations on 

concurrent operations. 

10.3.20 NRC Staff Determination 

The NRC staff has reviewed the applicant’s conceptual decontamination and DP and 

determined that IIFP has considered site-specific activities necessary to decontaminate and 

decommission the FEP/DUP. Therefore, as discussed in Sections 10.3.3 through 10.3.19 

above, the NRC staff finds that the applicant’s proposed process provides reasonable 
10-12 

assurance that decommissioning can be performed in accordance with the guidance in 

NUREG-1757, Volume 3 (NRC, 2003), and the requirements in 10 CFR Part 20, Subpart E, and 

10 CFR 40.42(g). 

10.4 Site-Specific Cost Estimate

10.4.1 Regulatory Requirements 

IIFP is required to submit a DFP under 10 CFR 40.36(a), which states that an “applicant for a 

specific license authorizing the possession and use of more than 100 mCi of source material in 

a readily dispersible form shall submit a decommissioning funding plan....” 

Among other things, 10 CFR 40.36(d) states that the DFP must “contain a cost estimate for 

decommissioning and a description of the method of assuring funds for decommissioning ... 

including means for adjusting cost estimates and associated funding levels periodically over the 

life of the facility....” As part of its LA (IIFP, 2012a), IIFP submitted a DFP and cost estimate. 

10.4.2 Summary of Staff Review 

The NRC staff reviewed the information provided in the LA (IIFP, 2012a) and the responses to 

NRC’s requests for additional information (RAls) dated April 29, 2011 (IIFP, 2011a); 

June 29, 2011 (IIFP, 2011b); and August 12, 2011 (IIFP, 2011c) in order to determine whether 

the applicant’s DFP and cost estimate are acceptable. 

10.4.3 Staff Review of Decommissioning Funding Plan and Cost Estimate 

Applicants for licenses under 10 CFR Part 40, “Domestic Licensing of Source Material,” are 

subject to financial assurance requirements for decommissioning, decontamination, and 

reclamation in accordance with 10 CFR 40.36, “Financial Assurance and Recordkeeping for 

Decommissioning.” Chapter 10 of the LA contains a DFP and cost estimate, as required by 

10 CFR 40.36(a) and 10 CFR 40.36 (d), respectively. The NRC staff used NUREG-1757, 

Volume 3 (NRC, 2003), to guide its review. 

The NRC staff reviewed the DFP, cost estimate, and responses to NRC’s RAIs. In conjunction 

with IIFP’s responses to the RAIs, the NRC staff finds the cost estimate acceptable because it 

(1) is based on reasonable and documented assumptions, (2) incorporates the costs of a 

third-party contractor, (3) does not take credit for any salvage value that might be realized from 

the sale of potential assets during or after decommissioning, (4) includes the estimated costs of 

radioactive waste disposal, and (5) includes a 25-percent contingency factor. The cost estimate 

also includes the estimated cost of deconverting the estimated amount of depleted uranium 

remaining at the site at the time of decommissioning. IIFP estimates that the cost to 

decommission the site and facility in a manner that is consistent with the unrestricted release 

criteria set forth in 10 CFR 20.1402 would be $14,204,535. 

As set forth in 10 CFR 40.36(d), each DFP must also contain a cost estimate submitted 

concurrently with the LA, for decommissioning and a description of the method of assuring 

funds. In its August 12, 2011, submittal (IIFP, 2011c), IIFP requested an exemption to the 

requirement in 10 CFR 40.36(d) to submit the certificate of financial assurance along with the 

DFP. The certification of financial assurance requires a licensee to certify that financial 

assurance has been provided in the amount of the cost estimate for decommissioning. 

Specifically, the exemption would allow IIFP to provide the certification of financial assurance 
10-13 

separately from the DFP. The NRC approves this exemption, but imposes two license 

conditions, as described below. These license conditions require the certification of financial 

assurance to be provided to the NRC with the financial instrument it intends to execute 

6 months before commencing operations. The license conditions also require IIFP to provide 

the NRC with executed financial instruments at least 21 days before commencing operations. 

As described in detail in Section 1.2.3.5.2 of this SER, the NRC staff found IIFP’s exemption 

request acceptable. 1

As stated in Section 1.2.3.5 of this SER, the NRC has reviewed and approved this exemption 

request consistent with 40.36(d) and the categorical exclusion requirements in 51.22(c)(25). 

As part of its RAI response, IIFP provided a draft payment surety bond (PSB), a draft standby 

trust agreement (STA), and a draft certification of financial assurance (certification) for review. 

The NRC staff reviewed the draft PSB, STA, and certification and finds that the language of the 

documents are consistent with the guidance in NUREG-1757, Volume 3, Appendix A.9, 

Appendix A.17, and Appendix A.2.4, respectively (NRC, 2003). 

Consistent with the financial assurance exemption request described in Section 1.2.3.5.2 of this 

Safety Evaluation Report, the applicant will provide the NRC with its proposed final draft 

financial instruments for review in advance of IIFP obtaining NRC-licensed material at the site in 

quantities and form requiring decommissioning financial assurance (e.g., in quantities and form 

that exceed the regulatory thresholds of 10 CFR 40.22 and 10 CFR 40.36). In addition, IIFP will 

provide executed versions of the reviewed financial instruments before the receipt of 

NRC-licensed material; and, if they are acceptable, the NRC staff will provide its approval at that 

time. The NRC staff will impose the following license condition to which the applicant has 

agreed: 

 

Consistent with 10 CFR 40.36(a), prior to obtaining radioactive material under the 

NRC license, the licensee shall provide final copies of proposed financial 

assurance instruments and certification of financial assurance to the NRC for 

review at least 6 months before the anticipated date for obtaining radioactive 

material under the NRC license and shall provide final, executed copies of the 

NRC-reviewed financial assurance instruments and certification of financial 

assurance to NRC at least 21 days before the anticipated date for obtaining the 

radioactive material at the site. The amount of financial assurance provided shall 

be at least as great as the NRC-approved cost estimate. 

The licensee shall not obtain radioactive material under the NRC license, in any 

form, in an amount that would require financial assurance as set forth in 

10 CFR 40.36, until the NRC reviews and approves the executed financial 

assurance instrument and certification of financial assurance. 

 

1 As described in Section 1.2.3.5 of this SER, IIFP stated that it may use equipment obtained from 

the Sequoyah Fuels facility to build its facility. IIFP stated that if any of these components cannot 

be decontaminated, these components would be appropriately shipped to the site. To provide 

assurance that decommissioning funding is sufficient during the construction phase, the staff will 

impose another license condition that would require financial assurance if the total amount of 

contamination on equipment exceeds the regulatory limit set forth in 10 CFR 40.36.
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As required by 10 CFR 40.36(d), licensees must update their cost estimates for 

decommissioning at least triennially. In its application, IIFP committed to updating its cost 

estimate “approximately every three years.” The NRC staff finds that requiring IIFP to update 

its estimate more frequently than triennially would be prudent because the facility will not 

commence operations until sometime after license issuance and, consequently, the costs of 

decommissioning may change. Therefore, the NRC staff will impose the following license 

condition to which the applicant has agreed: 

The licensee shall provide an updated DFP and updated facility DCE to the NRC 

for review at least 6 months before the planned date for obtaining licensed 

material in any form, including contamination, in an amount that would require 

financial assurance as set forth in 10 CFR 40.36. 

On an annual basis, starting on the anniversary date of obtaining licensed 

material, and for 2 subsequent years the licensee shall provide an updated DFP 

and updated facility DCE to the NRC for review. With each annual update, if the 

cost estimate exceeds the amount of financial assurance provided, the licensee 

shall provide financial assurance in the amount of the updated cost estimate and 

an updated certification of financial assurance to the NRC for review and 

approval. 

All updates to the DFP and cost estimate for facility decommissioning and 

financial assurance instruments shall reflect current year U.S. dollars and shall 

encompass all current cost data—including but not limited to taking into account 

changes in inflation, possession limits, licensed material, labor rates, disposal 

and shipping rates, and site and facility factors. All costs shall be supported by a 

detailed basis, be based on the costs of a third-party contractor, and shall not 

take credit for any salvage value that might be realized from the sale of potential 

assets during or after decommissioning. The total cost estimate shall include a 

contingency factor of at least 25 percent. 

10.4.4 Evaluation Findings 

The NRC staff has evaluated the applicant’s plans and financial assurance for decommissioning 

in accordance with the “Consolidated NMSS Decommissioning Guidance,” NUREG-1757, 

(NRC, 2003). Relying on the information provided in the LA (IIFP, 2012a) and the responses to 

the NRC’s RAls dated April 29, 2011 (IIFP, 2011a); June 29, 2011 (IIFP, 2011b); and 

August 12, 2011 (IIFP, 2011c), the NRC staff finds that (1) the DFP and cost estimate for 

decommissioning and decontaminating the site and facility to unrestricted release criteria is 

consistent with NUREG-1757, Volume 3 (NRC, 2003); (2) the DFP and cost estimate is based 

on reasonable and documented assumptions; (3) the cost estimate is based on the costs of a 

third-party contractor; (4) the cost estimate includes a 25-percent contingency factor; (5) the 

cost estimate does not take credit for any salvage value that might be realized from the sale of 

potential assets during or after decommissioning; and (6) the PSB, STA, and certification are 

consistent with the applicable sections of NUREG-1757, Volume 3 (NRC, 2003). On the basis 

of this evaluation, the NRC staff has determined that the applicant’s plans and financial 

assurance for decommissioning comply with 10 CFR 40.42 and 10 CFR 40.36 and are 

acceptable. 
10-15 

10.5 References

(IIFP, 2012a) International Isotopes Fluorine Products, Inc., “Fluorine Extraction Process and 

Depleted Uranium Deconversion Plant (FEP/DUP) License Application, Revision B,” May 2012, 

Agencywide Documents Access and Management System (ADAMS) Accession 

No. ML12123A245. 

(IIFP, 2012b) International Isotopes Fluorine Products, Inc., “ISA Summary Rev. B for IIFP,” 

May 2012, Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML12123A245. 

(IIFP, 2011a) International Isotopes Fluorine Products, Inc., “Official Response to Financial 

Assurance RAIs,” April 29, 2011, Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML11129A083. 

(IIFP, 2011b) International Isotopes Fluorine Products, Inc., “Letter from INIS containing the 

Public Revision B of the Financial Assurance RAI Responses and Follow-up RAI Responses,” 

June 29, 2011, Agencywide Documents Access and Management System (ADAMS) Accession 

No. (Public) ML11195A176. 

(IIFP, 2011c) International Isotopes Fluorine Products, Inc., “Letter from INIS re: Affidavit and 

Financial Assurance Responses to the Second Follow-up RAIs for the IIFP License Application,” 

August 12, 2011, Agencywide Documents Access and Management System (ADAMS) 

Accession No. ML11234A074 

(IIFP, 2011d) International Isotopes Fluorine Products, Inc., “Fluorine Extraction Process and 

Depleted Uranium Deconversion Plant (FEP/DUP) Environmental Report,” Rev. A, 2011, 

Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML100120758. 

(NRC, 2006) U.S. Nuclear Regulatory Commission, “Consolidated NMSS Decommissioning 

Guidance,” NUREG-1757, Vol. 1, Rev. 2, “Decommissioning Process for Materials Licensees,” 

2006. 

(NRC, 2003) U.S. Nuclear Regulatory Commission, "Consolidated NMSS Decommissioning 

Guidance,” NUREG-1757, Vol. 3, “Financial Assurance, Recordkeeping, and Timeliness,” 2003. 

(NRC, 2000) U.S. Nuclear Regulatory Commission, “Multi-Agency Radiation Survey and Site 

Investigation Manual,” NUREG-1575, August 2000. 

(NRC, 1998) U.S. Nuclear Regulatory Commission, “A Nonparametric Statistical Methodology 

for the Design and Analysis of Final Status Decommissioning Surveys - Interim Draft Report for 

Comment and Use,” NUREG-1505, Rev. 1, 1998. 

 
10-16 

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11-1 

11.0 MANAGEMENT MEASURES 

Management measures are functions that International Isotopes Fluorine Products, Inc. (IIFP or 

the applicant), will perform, generally on a continuing basis, which are applied to items relied on 

for safety (IROFS) to provide reasonable assurance that the IROFS are available and reliable to 

perform their function when needed. The applicant will implement management measures to 

ensure compliance with performance requirements, and the degree to which they will be applied 

will be a function of the item’s importance in terms of meeting performance requirements, as 

evaluated in the Integrated Safety Analysis (ISA) (IIFP, 2012b). This chapter addresses each of 

the management measures included in the definition of management measures found in Title 10 

of the Code of Federal Regulations (10 CFR) Part 70, “Domestic Licensing of Special Nuclear 

Material,” including (1) configuration management (CM), (2) maintenance, (3) training and 

qualifications, (4) procedures, (5) audits and assessments, (6) incident investigations, 

(7) records management, and (8) other quality assurance (QA) elements. 

The purpose of this review is to verify whether IIFP’s license application (LA) for the Fluorine 

Extraction Process/Depleted Uranium Deconversion Process (FEP/DUP) plant provided 

conclusive information to ensure that the management measures applied to IROFS, as 

documented in the ISA Summary (IIFP, 2012b), offer adequate assurance that the IROFS will 

be available and reliable and consistent with the provisions of 10 CFR 70.61, “Performance 

Requirements.” This review also determines whether the measures are applied to the IROFS in 

a graded manner commensurate with their importance to safety. 

11.1 Regulatory Requirements

The following requirements for fuel cycle facility management measures, specified in 

10 CFR Part 70, apply to the IIFP facility: 

• 10 CFR 70.4, “Definitions,” which states that management measures include (1) CM, 

(2) maintenance, (3) training and qualifications, (4) procedures, (5) audits and 

assessments, (6) incident investigations, (7) records management, and (8) other QA 

elements 

• 10 CFR 70.62(a)(3), which states that records must be kept for all IROFS failures, 

describes required data to be reported, and sets time requirements for updating the 

records 

• 10 CFR 70.62(d), which requires a licensee to establish management measures for 

application to engineered and administrative controls and control systems that are 

identified as IROFS, under 10 CFR 70.61(e), to ensure they are available and reliable 

• 10 CFR 70.72, “Facility Changes and Change Process,” which requires a licensee to 

establish a CM program to evaluate, implement, and track changes to the facility; 

structures, systems, and components (SSCs); processes; and activities of personnel. 

11.2 Regulatory Acceptance Criteria

Section 11.4.3 of NUREG-1520, Revision 1, “Standard Review Plan for the Review of a License 

Application for a Fuel Cycle Facility,” issued March 2002 (NRC, 2002), contains the acceptance 
11-2 

criteria for the U.S. Nuclear Regulatory Commission’s (NRC’s) review of the FEP/DUP facility 

management measures program. 

11.3 Staff Review and Analysis

As stated by the applicant, the IIFP project is currently in the development, conceptual design, 

and licensing phase. The applicant will apply management measures throughout all phases of 

construction, operations, and maintenance of the facility. 

The applicant provided the IIFP design, construction, and operations organization description 

and organizational charts in the LA Chapter 2 (IIFP, 2012a). The IIFP President has overall 

responsibility for the establishment and implementation of the QA policies, goals, and objectives 

for IIFP. The application emphasizes that the responsibilities of line managers in the 

implementation and maintenance of the management measures policies and procedures are 

commensurate with delegated authority. 

As described in the LA (IIFP, 2012a), the IIFP Chief Operations Officer (COO) will be 

responsible for assuring that management measures in accordance with graded QA are being 

implemented by the Design and Build (DB) contractor chosen by IIFP to perform detailed design 

and construction. Also, the application clearly delineates in the LA, Chapter 2 (IIFP, 2012a), 

that the IIFP QA Manager and Environmental Safety and Health (ESH) Manager have 

responsibilities and authorities independent of production, engineering, and maintenance 

organizational functions. The ESH Manager provides authorized oversight and technical 

direction to ensure that production never takes priority over the safety and protection of 

employees, the public, and the environment. The ESH Manager reports to the COO directly. 

11.3.1 Configuration Management 

The IIFP QA Program requirements and associated procedures implement the IIFP CM 

Program. The application describes the CM policy, design requirements, document control, 

change control, and assessments. 

Configuration Management Policy

In Section 11.1.1, “Configuration Management Policy,” of the LA (IIFP, 2012a), the applicant 

stated that CM will provide a means to establish and maintain a technical baseline for the facility 

based on clearly defined requirements. In addition, the applicant described design documents 

as those that provide design input, analysis, and results specifically for IROFS with the 

appropriate QA level. In Section 11.8.2.2 of the LA (IIFP, 2012a), the applicant defined the QA 

levels, which will be discussed later in this Safety Evaluation Report (SER). The CM Program 

will provide a formal review process, including interdisciplinary reviews, for design changes. 

During design and construction, CM will be based on design control provisions and associated 

procedural control of the design documents. During construction, changes to drawings and 

specifications issued for construction, procurement, or fabrication will be systematically 

reviewed, verified, evaluated for impact, and approved before implementation. As the project 

progresses from design and construction to operation, CM responsibilities will be transferred 

from the Design Engineering Manager to the engineering organization. 

The scope of the CM Program includes all IROFS identified in the ISA and any items which may 

affect the function of the IROFS. The scope of documents included in the CM program will 
11-3 

expand throughout the design process. As described by the applicant, design documents that 

are subject to CM include calculations, safety analyses, design criteria, engineering drawings, 

system descriptions, technical documents, and specifications that establish design requirements 

for IROFS. During construction, the scope of CM includes documents such as vendor data, test 

data, inspection data, and applicable procedures. The CM Program will also address those 

documents generated through functional interfaces of the CM Program with other management 

measures, including QA, procedures, incident investigation audits and assessments, 

maintenance, records management, document control, training, and qualifications. The 

applicant will establish CM procedures that will provide for evaluation, implementation, and 

tracking of changes to IROFS, as well as processes, equipment, computer programs, and 

activities of personnel that impact IROFS. 

The applicant stated that the CM objectives are to ensure design and operation within the 

design basis of IROFS. As described by the applicant in the LA Chapter 11 (IIFP, 2012a), 

configuration control will be implemented through controlling procedures which address the 

preparation, review (including interdisciplinary review), verification, approvals, and distribution 

for use of the detailed design. Engineering documents will be assessed for QA level 

classification. Changes to the approved design are subject to review to ensure consistency with 

the design basis of IROFS. 

The applicant will perform periodic audits and assessments of the CM Program and of the 

design to ensure that the system meets its goals and that the design is consistent with the 

design basis. The result of audits and assessments, or incident investigations, may result in the 

development of prompt corrective actions. The applicant’s corrective action process will be in 

accordance with the IIFP Quality Assurance Program Description (QAPD) (IIFP, 2012c) and 

associated procedures. 

Activities included in CM are conducted during design control provisions, in accordance with a 

systematic process of preparation, review, and approval. The process will ensure consistency 

between the design and design bases of IROFS. In addition, CM includes the activities for 

IROFS during operations to ensure that the activities are within the limits and constraints 

established in the ISA. The CM Program also ensures that changes to the facility are controlled 

and are in accordance with the requirements of 10 CFR 70.72. Finally, CM includes the training 

records of personnel to ensure that only qualified personnel will be performing activities 

associated with the operation of IROFS. 

The applicant adequately described how CM will be implemented for management measures 

throughout the organizational structure and staffing interfaces. The DB Engineering Manager 

will administer CM during design and construction, with input from the engineers responsible for 

each discipline. The Configuration Manager will have the responsibility to ensure the 

appropriate conduct of interdisciplinary reviews. 

The applicant stated that the primary IIFP contractors will be responsible for development of 

their respective QA programs and CM elements. The contractor’s programs must be consistent 

with the requirements of the IIFP QA Program. Section 11.1.1.4 of the LA (IIFP, 2012a) 

documents and describes the interfaces among IIFP and contractors performing quality-related 

activities. IIFP and contractors are responsible for identifying problems with quality and 

elevating their concerns to the appropriate management, if there is a potential problem with a 

system component. 
11-4 

Design Requirements

The Design Engineering organization establishes and maintains the design requirements during 

design and construction; the Engineering Manager establishes and maintains the design 

requirements during operations. The applicant stated that the CM Program would document 

and control design requirements. 

The application stated that IROFS and any items that affect the function of the IROFS will be 

designated as QA Level 1 or QA Level 2. The ISA Summary lists the IROFS (IIFP, 2012b). The 

applicant committed in the LA, Section 11.1.2, “Design Requirements,” to augment and maintain 

the IROFS list, as appropriate, so that it remains current during detailed design of the facility 

(IIFP, 2012a). The design documents associated with IROFS will be subject to interdisciplinary 

reviews and design verification. In addition, the applicant will evaluate changes to the design to 

ensure consistency with the design basis. Any computer codes used in the design of IROFS 

will be subject to the same design control measures, with additional requirements as 

appropriate for software control, verification, and validation. 

The applicant stated that qualified individuals will prepare design documents—including the 

appropriate codes, standards, and licensing commitments—and identify any deviations or 

changes in the design documentation package. The application states that an additional 

qualified individual will review the design documentation package for concept and conformity to 

design inputs. The manager having overall responsibility for the design function will approve the 

document, while the Configuration Manager will ensure that the designated engineering 

organization documents the entire review process in accordance with approved procedures. 

This will ensure that the design documents specify appropriate quality standards, including 

quantitative or qualitative acceptance criteria. The QA Manager will audit the design control 

process through augmented audit teams. 

Qualified individuals, other than those who performed the design but who may be from the same 

organization, will perform the design verification. Any verification performed by supervisors of 

the individuals doing the design will be documented and approved in advance by the 

supervisor’s management. The applicant stated that any supervisor approving a design may 

not specify a singular design approach or rule out certain design considerations. The supervisor 

may not establish the inputs used in the design or be the only individual in the organization to 

perform the verification. The independent design verification will be accomplished before the 

design document is used by other organizations for design work or to support other activities, 

such as procurement, construction, or installation. However, when the independent verification 

is not practical because of time constraints, the applicant will identify and control the unverified 

portion of the document. The applicant stated that, in all cases, the design verification will be 

completed before relying on the item to perform its function or before the installation becomes 

irreversible. The applicant will review, check, and approve changes to the design and 

procurement documents commensurate with the original approval requirements. 

The Corrective Action Program and its procedures will address design deficiencies affecting 

IROFS. The responsible manager will receive the nonconformance report in order to complete 

a review of the problem and assess the extent of the negative conditions. As required, the 

engineering organization will be informed for coordination of any necessary revisions. 

The applicant described the design interface and stated that it will be maintained by 

communication among the principal users. During the operational phase, any design changes 

will be provided to pertinent personnel to ensure the correct performance of their duties. 
11-5 

Document Control

The applicant stated that it had established procedures to control the preparation and issuance 

of documents, such as the ISA, and all procedures that pertain to IROFS. IIFP will develop 

procedures for training, QA, maintenance, audits and assessments, emergency operating 

procedures, emergency operating plans, system modification documents, assessment reports, 

and others that seem applicable as part of CM. The applicant will put in place measures to 

ensure that documents—including revisions—are adequately reviewed, approved, and released 

for use by authorized personnel. The document control procedures establish the distribution 

requirements and controls to ensure that documents are transmitted and received in a timely 

manner at the appropriate locations. The controlled copies are distributed to the persons 

performing the activity for their use. 

The applicant uses an electronic document management system to file project records and to 

make available the official (controlled) copy of the current documents. The system indexes the 

controlled documents through unique numbering of the documents, including revision numbers. 

If hard copies are needed, the applicant will provide them in accordance with the applicable 

procedures. Superseded or cancelled documents are appropriately labeled and maintained as 

records for the life of the project or the termination of the license, whichever occurs later. 

Change Control

The applicant stated that procedures will control changes to the technical baseline. During the 

design stage of the project, the applicant will control the changes through the IIFP QAPD design 

control process (IIFP, 2012c). The applicant committed to an appropriate level of technical, 

management, and safety review and approval before implementation. The process will include 

interdisciplinary reviews as the applicant’s primary mechanism to ensure consistency of the 

design with the design basis. During construction and operations, the applicant will perform the 

appropriate reviews to ensure (1) consistency with the design basis of IROFS and (2) consistent 

with the ISA, the design is constructed and operated or modified within the limits of the design 

basis. 

The applicant will control changes to the design to ensure consistency through a systematic 

review of the design basis for the ISA and any other affected documents. During the design 

stage, the applicant committed to a systematic interdisciplinary review to ensure consistency 

between documents, including the design changes and safety assessment. Interdisciplinary 

reviews will also ensure that the design changes either (1) do not impact the ISA, (2) are 

accounted for in subsequent changes to the ISA, or (3) are not approved or implemented. 

During the construction stage, the applicant committed to document, review, approve, and post 

any changes to documents issued for construction, fabrication, and procurement against each 

affected design document. The changes will continue to be evaluated through interdisciplinary 

review to ensure compliance with procurement specifications and drawings. As required by 

10 CFR 70.72, the applicant must implement a change control process that includes the 

reporting of changes to the ISA made without prior NRC approval and must submit a license 

amendment for changes that require Commission approval before implementation. 

During the operations stage, the applicant committed to document, review, and approve any 

design changes before their implementation. For changes made to the ISA, in accordance with 

10 CFR 70.72, the applicant must have measures in place to ensure that changes to the onsite 
11-6 

documentation are made promptly to avoid inadvertent access by facility personnel to outdated 

information that may affect the performance of their duties. The applicant stated that it had 

established measures to ensure that the quality of the facility’s SSCs is not compromised by 

planned changes and that the system will maintain its quality through the modification. 

Administrative procedures that are approved by the ESH Manager or designee describe the 

modification process. The modification procedures will include the technical and quality 

requirements that must be met in order to implement the modification, as well as the 

requirements for initiating, approving, monitoring, designing, verifying, and documenting the 

modification. 

The applicant will evaluate the modifications to ensure consistency among the facility’s 

procedure, personnel training, testing program, and regulatory documents. Other areas that the 

applicant will consider during the evaluation of changes or modifications may include, but are 

not limited to, radiation exposure, lessons learned, QA aspects, potential operability or 

maintainability concerns, constructability concerns, post modification testing, environmental 

considerations, human factors, and modification costs. After the completion of the modification 

or change, the system is tested and personnel are trained to ensure correct operations. When 

the system becomes operational, all of the required documentation will be distributed to 

operations and maintenance staff, including formal notice to all appropriate managers. 

Assessments

The applicant committed in the LA, Section 11.1.6, “Assessments” (IIFP, 2012a), to conduct 

periodic assessments of the CM Program to determine the program’s effectiveness and to 

correct deficiencies. Assessments and system walkdowns will be planned, conducted, and 

documented in accordance with procedures. Incident investigations are conducted in 

accordance with the QA Program and associated corrective action procedures. The applicant 

stated that it will develop prompt, corrective actions as a result of incident investigation or in 

response to adverse audit/assessment results. 

11.3.2 Maintenance 

The applicant described the maintenance and functional testing programs for the operations 

phase of the facility. In the LA, Section 11.2 (IIFP, 2012a), the applicant committed to 

conducting planned and scheduled maintenance to ensure that equipment and controls are 

maintained in a condition of readiness to perform as intended when required. 

The applicant will implement measures to ensure that the quality of these IROFS is not 

compromised by planned changes or maintenance activities in order to provide for the 

continued safe and reliable operation of the facility IROFS. The applicant will use a systems 

based program for planning, scheduling, tracking, and maintaining records for maintenance 

activities that affect IROFS. The maintenance function will be implemented in accordance with 

approved procedures for IROFS. The applicant will require, when applicable, that contractors 

working on or near IROFS identified in the ISA Summary (IIFP, 2012b) follow the same 

maintenance guidelines described for maintenance function activities. 

In terms of written procedures for maintenance involving IROFS, the applicant committed to 

(1) review the work performed for premaintenance activities; (2) notify affected parties before 

performing work and completion of maintenance work; and (3) ensure maintenance technicians 

follow comprehensive procedures, among others listed in Section 11.2.1 of the LA 

(IIFP, 2012a). 
11-7 

Surveillance and Monitoring

The applicant stated that it uses surveillance and monitoring to detect degradation and adverse 

trends of IROFS so that action may be taken before component failure. The parameters to be 

monitored are selected based upon the applicant’s ability to detect the predominant failure 

modes of the critical components. The applicant will establish performance criteria to monitor 

plant operations, IROFS function, and component parameters. Performance criteria will also be 

used to demonstrate that IROFS are being effectively controlled through appropriate predictive 

and repetitive maintenance strategies. 

IIFP will perform surveillance of IROFS at specified intervals. The surveillance frequency will be 

established in accordance with the IROFS degree of safety importance. The surveillance 

activity supports the determination of performance trends for IROFS, indicating when potential 

performance degradation exists, adjusting the preventive maintenance frequencies, or taking 

any other corrective action. Moreover, the applicant will evaluate surveillance and monitoring 

results to determine any impact on the ISA or any updates needed. For surveillance tests that 

can only be done while the equipment is out of service, the maintenance procedures prescribe 

the proper compensatory measures. 

The applicant described how incident investigations may identify root causes of failures. The 

Surveillance/Monitoring and Preventive Maintenance Programs will use incident investigation 

results as lessons learned when appropriate. IIFP will maintain records showing the current 

surveillance schedule, performance criteria, and test results in accordance with the records 

management system. 

Corrective Maintenance

The applicant’s corrective maintenance function will involve repairs or replacement of equipment 

that has unexpectedly degraded or failed. The corrective maintenance function provides a 

planned, systematic, integrated, and controlled approach for the repair and replacement 

activities associated with IROFS. After conducting corrective maintenance and before returning 

IROFS to operational status, the applicant will conduct a functional test to ensure that an IROFS 

will perform its intended safety function. If the performance of a replaced or repaired 

component differs from the original, the CM Program will review the change and 

preoperationally test it to ensure that the item will perform its safety function. The applicant will 

evaluate corrective maintenance results to determine any impact on the ISA or any updates 

needed. 

Preventive Maintenance

The applicant stated that preventive maintenance (PM) includes preplanned and scheduled 

periodic refurbishment; partial or complete overhaul; or replacement of IROFS, if necessary, to 

ensure their continued safety function. The applicant will use the results of surveillance and 

monitoring, including failure history and the review of the records of IROFS failures, to plan PM 

activities. 

As described in the application, the determination of PM frequency will seek to balance the 

objective of preventing failures through maintenance against the objective of minimizing the 

availability of IROFS because of PM. Feedback will be collected from PM, corrective 

maintenance, results from incident investigation, and identified root causes. The applicant will 
11-8 

use the feedback, as appropriate, to modify the frequency or scope of PM activities. IIFP will 

document the rationale for any PM deviations from industry standards or vendor 

recommendations. 

The applicant addressed calibration and testing within the PM function. Facility personnel will 

use PM Program procedures and calibration standards to calibrate equipment and monitor 

devices to plant safety and safeguards. The applicant will provide compensatory measures 

during testing of nonredundant IROFS to ensure that their function is performed until the item is 

back in service. 

After conducting PM and before returning QA Level 1 or QA Level 2 items to operational status, 

the applicant will conduct a functional test, if necessary, to ensure that the item will perform its 

intended safety function. The applicant will evaluate PM activity results to determine their 

impact on the ISA or if any updates needed. IIFP will maintain records pertaining to PM of 

IROFS, and items affecting IROFS, in accordance with the records management system. 

Functional Testing

The applicant stated that it will perform functional testing of IROFS, as appropriate following 

initial installation, as part of periodic surveillance testing and after corrective or preventive 

maintenance or calibration. The purpose of this testing will be to ensure that the item is capable 

of performing its safety function when required. The applicant committed in the LA, 

Section 11.2.2.4 (IIFP, 2012a), to performing functional tests in accordance with appropriate 

administrative controls and approved written procedures. The procedures define the method for 

the test and the required acceptable results that will be recorded and maintained as quality 

records. 

The applicant’s overall testing program is divided into two major testing programs: 

preoperational testing and post maintenance testing. The applicant described the 

preoperational testing program as testing conducted to initially determine various facility 

parameters and to initially verify the capability of SSCs to meet performance requirements. As 

stated by the applicant, the major objective of preoperational testing is to verify that IROFS are 

capable of performing their intended function. 

Initial startup testing is part of the preoperational program and will be performed beginning with 

the introduction of deplete uranium hexafluoride (DUF6) and ending with the startup of depleted 

uranium tetrafluoride (DUF4), silicon tetrafluoride (SiF4), and boron trifluoride (BF3) operations. 

The applicant stated that the purpose of initial startup testing is to ensure safe and orderly DUF6

feeding and to verify parameters assumed in the ISA. The applicant will maintain records of the 

preoperational and startup tests required before operation. 

The operational testing program will consist of periodic and special testing. The periodic testing 

will be conducted to monitor various facility parameters and verify the continuing integrity and 

capability of IROFS. IIFP may conduct special testing at the facility in a nonrecurring nature, 

which does not fall under any other category of testing program. 

The applicant will also establish a post maintenance testing program to provide assurance that 

IROFS will perform their intended function following maintenance activities. These tests will 

confirm that the maintenance performed was satisfactory, the identified deficiency was 

corrected, and the maintenance activity did not adversely affect the reliability of the IROFS. 

Before returning the equipment to service, the applicant will perform post maintenance testing to 
11-9 

ensure that the equipment is ready for operations. If no acceptable results are obtained, the 

applicant will conduct and document corrective actions through the nonconformance report 

process. 

11.3.3 Training and Qualifications 

Section 11.3 of the application describes the training program for operations of the facility, 

including training for the preoperational functional testing and initial startup testing. The 

applicant will provide training to each individual commensurate with his or her role and 

responsibilities during the design, construction, and operations phases. The QA training will be 

for personnel performing QA Level 1 and QA Level 2 work activities; personnel performing 

nondestructive examination, inspection, and testing; and QA auditors. 

The applicant stated that the principal objective of the training program is to ensure job 

proficiency of all facility personnel through effective training and qualification. The training 

program will be designed to accommodate future growth and meet commitments with applicable 

established regulations and standards. In addition, employees will be provided with training and 

continuing training, as applicable, to establish and maintain knowledge foundations and develop 

work performance skills. 

As described by IIFP, qualification will be indicated by successful completion of prescribed 

training, demonstration of the ability to perform assigned tasks, and maintenance of 

requirements established by regulation. The applicant will use a graded approach to systematic 

training incorporating methods to accomplish the analysis, design, development, 

implementation, and evaluation of training. Exceptions from training requirements might be 

granted when justified, documented and approved in accordance written procedures. 

11.3.3.1 Organization and Management of the Training Function 

The applicant stated that line managers are responsible and have commensurate authority to 

develop and effectively conduct training. The training function will support line managers by 

facilitating the planning, directing, analyzing, developing, conducting, evaluating, and controlling 

of the performance-based training process. 

The applicant will have facility procedures which specify the requirements for training of 

personnel performing activities related to IROFS. Procedures will allow for exceptions from 

training only when properly justified, documented, and approved. The applicant will update any 

lesson plans affected through the change control process of the CM Program when design 

changes or facility modifications are implemented. 

The applicant will maintain programmatic and individual training records to support management 

information needs associated with personnel training, job performance, and qualifications. 

Individual records will include general employee training, technical training, and employee 

development training. This training will be conducted at the facility and may include special 

company-sponsored training conducted by others. The applicant will maintain training records 

in an accurate and retrievable method; such records will be retained in accordance with the 

records management procedures. 
11-10 

11.3.3.2 Analysis and Identification of Functional Areas Requiring Training 

The applicant will perform a needs/job analysis and will identify tasks to ensure that it will 

provide appropriate training to personnel working on tasks related to IROFS. For each task, the 

needs/job analysis will include job hazards that are referred to as precaution and limitations. 

The applicant will evaluate the effectiveness of the training and materials and will update the list 

of activities selected for training, as necessitated by changes in procedures, processes, plant 

systems, equipment, or job scope. 

11.3.3.3 Position Training Requirements 

The applicant stated that it will develop minimum training requirements for positions whose 

activities are related to IROFS. IIFP will provide training for personnel of various abilities and 

experience backgrounds. Continuing training courses will be established to ensure that the 

appropriate personnel remain proficient. 

The objective of the training is to ensure safe and efficient operation of the facility and 

compliance with applicable established regulations and requirements. Training requirements, 

as described by the applicant, will be applicable to, but not necessarily restricted to, those 

personnel within the plant organization who have a direct relationship to the operation, 

maintenance, testing, or other technical aspect of the facility IROFS. 

11.3.3.4 Training Basis and Objectives 

The applicant will design its training program to prepare initial and replacement personnel for 

safe, reliable, and efficient operation of the IIFP facility. The applicant will emphasize the safety 

requirements where human actions are important to safety. Established learning objectives will 

identify the training content and define satisfactory trainee performance for the task, or a group 

of tasks, selected for training from the job analysis. Learning objectives will be sequenced 

within training materials based on the relationship to one another and documented in lesson 

plans and training guides. 

11.3.3.5 Organization of Instruction, Using Lesson Plans, and Other Training Guides 

The applicant will develop lesson plans or other approved process-controlling documents from 

the learning objectives based on job performance requirements and under the guidance of the 

training functional organization. These documents will be reviewed and approved before they 

are issued or used by the training function and, generally, by the organization cognizant of the 

subject matter. Lesson plans or other approved process-controlling documents will be used for 

classroom training and on-the-job training, as required, and will include standards for evaluating 

acceptable trainee performance. 

11.3.3.6 Evaluation of Trainee Learning 

The applicant will use observation, demonstration, or oral or written tests to evaluate the 

trainee’s understanding and command of learning activities. Individuals qualified in the subject 

matter will perform the evaluations. 
11-11 

11.3.3.7 Categories of Required Training 

The applicant described the different categories of required training in the LA, Section 11.3.7.1 

through the LA, Section 11.3.7.5 (IIFP, 2012a), which include the following: 

• General Employee Training (GET): GET includes QA, radiation protection, environment, 

health and safety, emergency response, and administrative policies established by 

facility management and applicable regulations, among others. Unescorted access to 

the controlled access area is permitted only after completion of the appropriate level of 

GET. 

• Radiological Safety Training: The radiation safety training program includes topics such 

as radiation and radioactive materials; risks involved in receiving low-level radiation 

exposure in accordance with 10 CFR 19.12, “Instruction to Workers”; notices, reports, 

and instruction to workers; inspection and investigation; and the basic criteria and 

practices for radiation protection. 

• Industrial Safety/Industrial Hygiene: The applicant will conduct industrial safety/industrial 

hygiene training (other than those general aspects that are addressed in GET), including 

various aspects of industrial and chemical safety protection, as necessary, to train new 

employees in specific job duties and to provide refresher training topics to workers 

depending on employee job responsibilities. 

• Emergency Preparedness/Security and Fire Brigade: This category includes the details 

on the development, maintenance, and implementation of the IIFP Emergency Plan (EP) 

and EP implementing procedures. It also includes the initial training for all new fire 

brigade members, classroom refresher training and drills, annual practical training, and 

leadership training for fire brigade and incident commanders. 

• Technical Training: The applicant will design, develop, and implement technical training 

to assist facility employees in gaining an understanding of applicable fundamentals, 

procedures, and practices related to IROFS and in developing skills necessary to 

perform assigned work related to IROFS. It consists of four segments: initial training, 

on-the-job training and qualifications, continuing training, and special training. 

11.3.3.8 Evaluation of Training Effectiveness 

The applicant will evaluate the training program periodically to measure program effectiveness. 

The evaluation will consider feedback provided from trainees after completion of the classroom 

training sessions for program improvement. It will determine strengths and weaknesses, 

whether the program content matches current job needs, and whether corrective actions are 

needed to improve the program’s effectiveness. 

The training function will be leading any training program evaluations and implementing the 

corrective actions. Evaluation results will be documented highlighting the program’s strengths 

and weaknesses. The applicant will review identified weaknesses; recommend improvements; 

and make changes to the affected procedures, practices, or training materials. 

The QA department will audit the facility system and training program. Trainees and vendors 

can provide input related to the facility training program through surveys, questionnaires; 
11-12 

performance appraisals, staff evaluations, and overall training program effectiveness evaluation 

techniques. 

11.3.3.9 Personnel Qualification 

The applicant will provide the training and qualification requirements for QA personnel and key 

management positions. The applicant will establish and implement qualification and training 

requirements for process operator candidates in plant procedures. 

11.3.3.10 Provisions for Continuing Assurance 

The applicant will establish applicable continuing education or periodic retraining to ensure 

personnel proficiency and retention of knowledge and skills which are important to Operations. 

Personnel who perform activities relied on for safety will be evaluated periodically by written 

test, oral test, or on-the-job performance to determine whether they are capable of continuing 

performing these activities. The applicant will provide retraining or other appropriate action 

when results of the evaluations dictate the need. In addition, the applicant will provide retraining 

to personnel if any new or revised information results in plant modifications, procedure changes, 

and QA program changes. Unacceptable, individual performance will be transmitted to the 

appropriate line manager. 

11.3.4 Procedures Development and Implementation 

The applicant will conduct all activities involving IROFS or QA Level 1 and QA Level 2 items in 

accordance with approved procedures. Compliance with procedures is mandatory. The 

applicant will use procedures to control IROFS activities. 

11.3.4.1 Type of Procedure 

As stated by the applicant, procedures will be categorized as management control procedures 

or operating procedures/instructions. Management control procedures are those that will 

describe administrative and general practices approved and issued by management at a level 

appropriate to the scope of the practice, directing and controlling activities across the various 

organizational functions. Operating procedures will provide specific direction for task-based 

work and will be used to directly control process operations at the work place. 

The applicant described four types of plant procedures that it will use to control QA Level 1 and 

QA Level 2 plant and administrative activities: (1) operating procedures, (2) management 

control procedures, (3) maintenance procedures, and (4) emergency procedures. It is clearly 

stated in the application that if any aspect of a procedure is unclear or incorrect as written, 

personnel shall safely stop the operation and/or activity, contact management and shall not 

restart until corrective action has been taken. 
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11.3.4.1.1 Operating Procedures 

As described by the applicant, operating procedures will be used to directly control process 

operations. The operating procedures will include, among other things, the (1) purpose of the 

activity; (2) governing regulations; polices; and guidelines; (3) type of procedure; (4) steps for 

each operating process stage; (5) hazards and safety considerations; (6) operating limits; 

(7) measures to prevent exposure; (8) associated IROFS and their functions; and (9) the 

timeframe for which the procedure is valid. 

The applicant states that it will implement a methodology for identifying, developing, approving, 

implementing, and controlling operating procedures. The method will ensure, as a minimum, 

that, (1) ISA results are considered when identifying needed procedures; (2) operating limits and 

IROFS are specified in the procedure; (3) procedures include required actions for off-normal 

conditions of operation; as well as normal operations; (4) if needed; safety checkpoints are 

identified at appropriate steps in the procedure; (5) procedures are validated through field tests; 

(6) procedures are approved by management personnel responsible and accountable for the 

operation; (7) mechanisms are specified for revising and reissuing procedures in a controlled 

manner; (8) QA elements and CM Program at the facility provide reasonable assurance that 

current procedures are available and used at all work locations; and (9) the facility training 

program trains the required persons in the use of the latest procedures available. 

11.3.4.1.2 Management Control (Administrative) Procedures 

As described by the applicant, management control procedures will provide programmatic 

requirements for policy or programs and administrative systems. The applicant will use 

management control procedures to perform activities that support production and control 

processes with IROFS or hazardous chemicals, or both, incident to the processing of licensed 

material. Management control procedures will include management measures such as (1) CM; 

(2) industrial safety/industrial hygiene; radiation safety; chemical safety; and fire safety; (3) QA; 

(4) design control; (5) facility personnel training and qualification; (6) audits and assessments; 

(7) incident investigations; (8) recordkeeping and document control; (9) reporting; and 

(10) procurement. 

The procedures for construction will be an integral part of the procedures of the DB contractor 

that IIFP will choose. In case the DB contractor uses its own QA program, instead of IIFP’s QA 

Program, the applicant will review and approve the QA program that will be used by the DB 

contractor. 

11.3.4.1.3 Maintenance Procedures 

Maintenance procedures will address the preventive and corrective maintenance, surveillances, 

functional or post maintenance testing, and the requirements for premaintenance activities 

involving reviews of the work to be performed and reviews of procedures for accuracy and 

completeness. The LA, Section 11.4.1.3 (IIFP, 2012a), mentions testing, calibration, and 

selection and qualification of maintenance personnel as just some of the activities that will be 

performed in accordance with written procedures. 

Various safety disciplines, including fire, radiation, and industrial and chemical process safety, 

will review maintenance procedures. These procedures will describe, at a minimum, the 

following: 
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• premaintenance activities 

• steps that require notification of all affected parties before performing work on 

completion of maintenance work 

• controls on and specification of any replacement components or materials to be used to 

ensure like-kind replacement 

• functional testing to verify operability of the equipment 

• tracking and records management of maintenance activities 

• safe work practices (e.g., lockout/tagout, confined space entry, control of exclusion area, 

radiation or hot work permits, and fire, chemical, and environmental requirements). 

11.3.4.1.4 Emergency Procedures 

The applicant will develop and maintain a documented controlled set of EP Implementation 

Procedures (EPIPs) applicable to the IIFP facility. The EPIPs will include emergency 

instructions pertinent to specific accident scenarios and other categorized nonroutine 

operational events clearly stating the duties, responsibilities, action levels, and actions to be 

taken by responders. 

The ESH Manager will review periodically the EP and will update it as needed. The applicant 

will not implement any changes to the plan that decrease its effectiveness without prior NRC 

approval. Changes that will not decrease the effectiveness of the EP may be implemented 

without NRC prior approval, provided the applicant submits the changes to the NRC and 

appropriate organizations within 3 months of making the changes. 

11.3.4.2 Procedures Process 

The applicant described the process of development and modification of procedures. The 

process will utilize the following nine basic elements to accomplish procedure development, 

review, approval, and control: (1) identification, (2) development, (3) verification, (4) review and 

comment resolution, (5) approval, (6) validation, (7) issuance, (8) change control, and 

(9) periodic review. 

11.3.4.2.1 Identification 

As described by the applicant, site managers are responsible for identifying the tasks to be 

included in procedures within their areas of control. Procedures are required when actions are 

taken necessary to prevent or mitigate the consequences of accidents described in the ISA. At 

a minimum, the applicant will require a procedure for any task or activity that affects QA Level 1 

and QA Level 2 SSCs. The applicant will evaluate new or revised NRC certification 

requirements to determine their impact on existing implementing procedures or to identify the 

need for new implementing procedures. 
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11.3.4.2.2 Development 

The applicant will categorize procedures in different designations depending on the 

administrative or nonadministrative use and the safety or financial consequences of failing to 

adhere to the requirements. The applicant will use the following designations: “In Hand,” which 

means that the procedure is of continuous use; “General Intent,” which means that the 

procedure is for reference use; or “Information Use,” which means that the procedure is for 

information only. 

The user organization will be responsible for procedure development, preparation, and quality. 

The applicant will require input and review by affected parties and other selected parties, such 

as safety and quality, to ensure that safety and QA requirements are identified and included 

when QA Level 1 or QA Level 2 SSCs are involved. 

The applicant utilized interviews with procedure users and process walkdowns to ensure that 

procedures are usable, reflect as-built conditions, guide production operations, and maintain 

management controls for safety and quality. The applicant will ensure that each procedure 

identifies and incorporates regulatory commitments, ISA, and QA Program requirements. 

11.3.4.2.3 Verification 

The applicant will review two basic attributes during the verification process: (1) technical 

accuracy, including all technical information, and (2) administrative verification to verify that the 

procedure has the correct format and style. Procedure verification will consist of a walkdown of 

the procedure in the field or a tabletop walkthrough to ensure its performance as written. 

11.3.4.2.4 Review and Comment Resolution 

The applicant will distribute draft new procedures and procedure changes, as needed, to 

technical and cross-discipline individuals not having direct responsibility for processing the new 

procedure or procedure change for their review. The applicant also described the 

comments/questions process that will ensure that any comments will be resolved with the 

originating organizations. If the original draft changes extensively, the revised draft procedure 

will be reverified, and the validation will be checked. The QA function will review QA 

implementing procedures for compliance and consistency with the QA Program. 

11.3.4.2.5 Approval 

The responsible manager will approve the procedures following the resolution of any review 

comments. Managers will also be responsible for ensuring that necessary training or required 

reading is completed before procedure implementation. 

11.3.4.2.6 Validation 

The applicant described the purpose of procedure validation as a process to ensure that 

technical errors or human factor issues will not be inadvertently introduced during the procedure 

review process. Qualified personnel will perform and document the validation, which may be 

accomplished by detailed evaluation of the procedure as part of a walkthrough exercise or as 

part of a walkthrough drill. If the particular system or process is not available for a walkthrough 

validation, talk through may be performed in the particular shop or training environment. 
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11.3.4.2.7 Issuance and Distribution 

The applicant will issue and control procedures in accordance with the applicant’s records 

management and document control programs. The applicant states that line managers, or 

designees, will be responsible for ensuring personnel doing work that requires the use of 

procedures have access to controlled copies of the required procedures. 

11.3.4.2.8 Procedures Process Change Control 

The applicant described the process for changes to procedures. The preparer of the change 

will document the change and the reason for it. The applicant will perform an evaluation in 

accordance with 10 CFR 70.72, as appropriate, to determine whether the proposed changes will 

require an amendment to the license. If an amendment to the license is needed, the change will 

not be implemented until the applicant receives prior approval from the NRC. A designated 

reviewer will then review the procedure. The functional area manager will be responsible for 

approving procedure changes and for determining whether a cross-disciplinary review is 

necessary and by which departments. The applicant will perform interdisciplinary reviews, at a 

minimum, for changes involving chemical safety, radiation safety, facility safety, engineering, 

and QA or ESH. The applicant will maintain the records for cross-functional reviews for all 

changes to procedures involving licensed materials of QA Level 1 and QA Level 2 items and 

activities, in accordance with the LA, Section 11.7 (IIFP, 2012a), “Records Management.” 

The applicant described the process for temporary changes to procedures. The temporary 

changes will be issued for production activities that are of a nonrecurring nature or when 

revision of a production or other permanent procedure is not practical. Temporary changes to 

procedures will not involve a change to IROFS and will not alter the intent of the original 

procedure. Two members of the facility management staff, at least one of whom is a shift 

superintendent or designee, will approve the temporary changes to procedures. Temporary 

changes to procedures will have a designated expiration date and may be made permanent 

once the change is reviewed and approved through the normal procedure change and approval 

process. 

11.3.4.2.9 Periodic Review 

The procedure owner or subject matter expert will perform a complete administrative and 

technical review periodically to ensure that information is complete and accurate and that the 

procedure is usable as written. Specifically, operating procedures will be reviewed every three 

years while EPIPs will be reviewed annually. In addition to periodic reviews, procedures will be 

reviewed after unusual incidents to determine if changes are appropriate. 

11.3.4.3 Use and Control of Procedures 

 

The applicant described uses of the different designations of the procedures. IIFP staff will 

perform in-hand or continuous use procedures step by step without deviation, unless the 

procedure allows deviation. General Intent or Reference Use procedures will be followed as 

written, unless the procedure allows deviation. Information Use procedures will be followed to 

implement programmatic requirements. 

The applicant will mark the controlled copies of procedures as “Controlled Copy” and 

information copies, which are not used to perform any work, as “Information Only.” The 

applicant stated that working copies will be verified as the latest version by managing and 
11-17 

limiting access to the most current revision of the document. Procedures will include provisions 

for operations to stop and place the process in safe condition, if a step of a procedure involving 

QA Level 1 and QA Level 2 SSCs cannot be performed as written. 

11.3.4.4 Topics To Be Covered in Procedures 

The LA, Section 11.4.4 (IIFP, 2012a), lists the topics covered by the different types of 

procedures including those used for: management controls, operations, maintenance, and 

emergencies. The list is not all inclusive but provides many of the activities that will be covered 

by operating, management control, maintenance, and emergency procedures. 

11.3.4.5 Temporary Procedures 

The applicant will have formal requirements governing the use of temporary procedures. The 

applicant will issue temporary procedures only when permanent procedures do not exist to 

(1) direct operations during testing, maintenance, and modifications, (2) provide guidance in 

unusual situations not within the scope of permanent procedures, and (3) ensure orderly and 

uniform production for short periods of time when the plant, a system, or component of a system 

is performing in a manner not covered by existing permanent procedures or has been modified 

or extended in such a manner that portions of existing procedures do not apply. The applicant 

stated that these temporary procedures may be used for a period of time which should not 

exceed 90 days or a period for which the temporary condition must exist, whichever is greater. 

If use of the temporary procedure exceeds the 90-day timeframe, the applicant will assess the 

procedure to ensure that it is appropriate to extend its use. 

11.3.4.6 Records 

The applicant will generate records that will be identified in the governing procedure and 

controlled according to the plant records management and document control program practices. 

11.3.5 Audits and Assessments 

The applicant stated that audits will be focused on verifying compliance with regulatory and 

procedural requirements and licensing commitments. Assessments will be focused on 

determining the effectiveness of activities and ensuring that IROFS are reliable and are 

available to perform their intended safety functions. The applicant will perform audits and 

assessments on critical work activities associated with facility safety, environmental protection, 

and other areas that it identifies. 

11.3.5.1 Audits 

The applicant stated that will conduct audits of the QA Level 1 and QA Level 2 work activities 

associated with IROFS, and any items that affect the function of the IROFS, in accordance with 

the QA Program, written plans, and checklists. Audits will be performed under the direction of a 

lead auditor. Lead auditors and staff auditors will be functionally and organizationally 

independent of the programs and activities audited or assessed. Audit teams may be 

supplemented with onsite or offsite technical specialists. 

Plant procedures will specify the method in which audit results should be documented and 

reported. The applicant will have provisions in place for immediate reporting and 

implementation of corrective action, where warranted. 
11-18 

11.3.5.2 Assessments 

As described by the applicant, management responsible for implementing the respective 

portions of the QA Program will perform and document assessments. They will assess the 

adequacy of the part of the program for which they are responsible and ensure its effective 

implementation. The application states that personnel may perform the assessment for their 

area, provided that they do not have direct responsibility for the specific area being evaluated. 

The assessment program contains management assessments and independent assessments. 

Management assessments review occupational safety and health, radiological protection, 

environmental compliance, fire safety, emergency preparedness and security, safety 

requirements, conduct of operations, conduct of maintenance, training, QA, maintenance, and 

CM, among other programs. Qualified staff personnel that are not directly responsible for 

production activities will routinely perform independent assessments. Any deficiencies identified 

during the assessment will be forwarded to the corrective action program and addressed by the 

responsible manager of the applicable area. 

11.3.5.3 Conduct of Audits and Assessments 

 

The frequency of audits and assessments will be based on the safety importance of the 

activities, the status of the activity, and on the work history. IIFP independent qualified auditors 

from the QA and technical organizations within ESH will conduct audits in accordance with 

written plans, procedures, and checklists. Audit frequency will be determined by the status and 

safety importance of the activities being performed. As stated by the applicant, audit team 

members will be indoctrinated in audit techniques, in addition to being independent to the 

function and area being audited. The applicant will also require auditors to have technical 

expertise or experience in the area being audited. 

Any deficiencies identified during the audit or assessment will be forwarded to the corrective 

action program and be addressed by the responsible manager of the applicable area or function 

in accordance with the Corrective Action Program (CAP) procedure. In addition, future audits 

and assessments must include a review to evaluate the effectiveness of corrective actions taken 

to address any deficiencies found. 

The results of the audits will be provided in a written report to the Plant Manager and provided 

to the managers responsible for the activities audited or assessed. The applicant will maintain 

records of the instructions and procedures, persons conducting the audits or assessments, and 

corrective actions taken. 

11.3.5.4 Activities Subject to Audit and Assessments 

The applicant will conduct audits and assessments in the areas of radiation safety, chemical 

safety, industrial safety (including fire protection), environmental protection, emergency 

management, QA, CM, maintenance, training and qualification, procedures, CAP/incident 

investigation, and records management. 

11.3.5.5 Scheduling of Audits and Assessments 

The applicant will establish a schedule identifying the audits and assessments to be performed 

and the responsible organization assigned to conduct the activity. All major activities will be 
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audited or assessed on a periodic basis. The applicant will review the schedule periodically and 

revise it as necessary to ensure coverage commensurate with current and planned activities. 

11.3.5.6 Procedures for Audits and Assessments 

The applicant stated that it will conduct internal and external audits and assessments using 

approved procedures that meet QA Program requirements. The procedures will provide 

requirements for audit and assessment activities, such as scheduling and planning, certification 

requirements of audit personnel, development of audit plans, any applicable checklists, 

performance of the audit and assessment, reporting and tracking of findings to closure, and 

closure of the audit and assessment. The applicable procedures will emphasize reporting and 

correction of findings to prevent recurrence. 

The applicant will require the team leader to develop the audit or assessment report to 

document the findings, observations, and recommendations for program improvement. The 

applicant will require that responsible managers review the reports and provide any required 

responses based on the reported findings. The CAP will track the audit and assessment results 

data, which the applicant will periodically analyze for potential trends. 

11.3.5.7 Qualification and Responsibilities for Audits and Assessments 

The applicant stated that the QA Manager will initiate the audits and, in coordination with the 

responsible lead auditor, will determine the scope of each audit. The QA Manager may also 

initiate special audits or expand the scope of audits, if appropriate. 

The lead auditor, as described by the applicant, will direct the audit team in developing 

checklists, instructions, or plans and performing the audit. Lead auditors will have to participate 

in a minimum of three QA audits or audit equivalent within a period of time not to exceed 

3 years before the date of certification. Audit equivalents include assessments, pre-award 

evaluations, or comprehensive surveillances (provided the prospective lead auditor took part in 

the planning, checklist development, performance, and reporting of the audit equivalent 

activities). One of the audits must be a nuclear-related QA audit or audit equivalent within the 

year preceding certification. 

Audits will be conducted in accordance with the checklists, but the audit team may expand the 

scope during the audit. The audit team will consist of one or more auditors who are responsible 

for performing audits in accordance with the applicable QA procedures. The auditors and lead 

auditors must hold certifications as required by the QA Program. Before certification under the 

IIFP QA Program, auditors will have to complete training on the IIFP QA Program; audit 

fundamentals, including audit scheduling, planning, performance, reporting, and follow-up action 

involved in conducting audits; objectives and techniques of performing audits; and on-the-job 

training. Auditors and lead auditors will be certified based on the QA Manager's evaluation of 

education, experience, professional qualifications, leadership, sound judgment, maturity, 

analytical ability, tenacity, and past performance and completion of QA training courses. 

The applicant will not require personnel performing assessments to be certified; but they will be 

required to complete QA orientation training, as well as training on the assessment process. 
11-20 

11.3.6 Incident Investigations and Corrective Action Process 

The applicant described the incident investigations and corrective action process in Section 11.6 

of the application (IIFP, 2012a). 

The applicant described the incident investigation process in the LA, Section 11.6.1 

(IIFP, 2012a), as a simple mechanism available for reporting deficiencies, abnormal events, and 

potentially unsafe conditions or activities to assure that the upset condition is understood and 

appropriate corrective actions are implemented to prevent recurrence. The applicant will 

consider each event in terms of its requirements for reporting in accordance with regulations 

and will evaluate the event to determine the level of investigation required. Written incident 

investigation and CAP procedures will address the applicant’s process of incident identification, 

investigation, root-cause analysis, recording, reporting, and followup. Incident investigation 

procedures will include guidance for classifying occurrences, including examples of threshold 

off-normal occurrences. The depth of the investigation will depend, as explained by the 

applicant, upon the severity of the classified incident in terms of the levels of uranium or 

chemicals released or the degree of potential for exposure of workers, the public, or the 

environment. 

The QA Manager will maintain a record and track to completion corrective actions to be 

implemented as a result of off-normal occurrence investigations in accordance with CAP 

procedures. 

The incident investigation process, as described by the applicant, will establish a process to 

investigate abnormal events that may occur during operation of the facility to determine their 

specific or generic root causes and generic implications, to recommend corrective actions, and 

to report to the NRC as required by 10 CFR 70.50, “Reporting Requirements,” and 

10 CFR 70.74, “Additional Reporting Requirements.” The investigation process will include a 

prompt, risk-based evaluation; and depending on the complexity and severity of the event, one 

individual may suffice to conduct the evaluation. The investigators will be independent from the 

line functions involved with the incident under investigation. Investigations will begin within 

48 hours of the abnormal event, or sooner, depending on the event’s safety significance. The 

record of IROFS failures required by 10 CFR 70.62(a)(3) will be reviewed as part of the 

investigation. The applicant will make record revisions necessitated by post failure investigation 

conclusions, accordingly. 

The applicant will appoint qualified internal or external investigators to serve on investigating 

teams when required. The teams will include at least one process expert and at least one team 

member trained in root-cause analysis. 

The applicant will monitor and document corrective actions through completion. The applicant 

will maintain auditable records and documentation related to abnormal events, investigations, 

and root-cause analyses so that “lessons learned” may be applied to future operations of the 

facility. For each abnormal event, the incident report will include a description, contributing 

factors, a root-cause analysis, findings, and recommendations. Details of the event sequence 

will be compared with accident sequences already considered in the ISA. The applicant will 

modify the ISA Summary (IIFP, 2012b) to include evaluation of the risk associated with 

accidents of the type actually experienced. 

The applicant will develop procedures for conducting incident investigations such as the 

following: 
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• a documented plan for investigating an abnormal event 

• a description of the functions, qualifications, and responsibilities of the manager who 

would lead the investigative team and those of the other team members; the scope of 

the team’s authority and responsibilities; and assurance of management cooperation 

• assurance of the team’s authority to obtain all of the information considered necessary 

and its independence from responsibility to the functional area involved in the incident 

under investigation 

• retention of documentation relating to abnormal events for 2 years or for the life of the 

operation, whichever is longer 

• guidance for personnel conducting the investigation on how to apply a reasonable 

systematic, structured approach to determine the specific or generic root causes and 

generic implications of the problem 

• requirements to make available original investigation reports to the NRC upon request 

• a system for monitoring the completion of appropriate corrective actions. 

In the LA, Section 11.6.2 (IIFP, 2012a), the applicant described the corrective action program. 

IIFP identified the responsibilities and provided the authority for those individuals involved in 

quality activities to identify any condition adverse to quality. These individuals will identify and 

document conditions adverse to quality, analyze and determine how the conditions can be 

corrected or resolved, and take the necessary steps to implement corrective actions in 

accordance with documented procedures. 

Employees will have the authority and responsibility to initiate the corrective action process if 

they discover deficiencies. The QA Program will contain procedures for identifying, reporting, 

resolving, documenting, and analyzing conditions adverse to quality. The applicant will analyze 

reports of conditions adverse to quality to identify trends in quality performance. Significant 

conditions adverse to quality and significant trends will be reported to senior management, in 

accordance with CAP procedures. The QA Manager will take follow-up actions to verify proper 

and timely implementation of the corrective actions. 

11.3.7 Records Management and Document Control 

The applicant will establish records management and document control programs to ensure that 

records and documents required by the IIFP QA Program are appropriately managed and 

controlled through standard methods and requirements for collecting, maintaining, and 

disposing of records. The programs will include administrative controls for the generation and 

revision of records and documents. The principal elements and a brief description of each are 

provided within the application and will be discussed below. 

The baseline design requirements in 70.64 require applicants to provide for adequate records of 

management measures and IROFS to facilitate their remaining available and reliable. The 

records management program described below are consistent with 70.64(a)(1). 
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11.3.7.1 Records Management Program 

The applicant will apply the Records Management Program to QA Level 1 and QA Level 2 

SSCs and activities. Records management may also apply to records related to ESH, financial, 

quality, emergency response, or investigation records, as required by regulations or approved 

procedures. The Records Management Program direct the handling, transmittal, storage, and 

retrieval of records. Records will be categorized and handled in accordance with their relative 

importance to safety and storage needs. A records organization will be responsible for the 

administration of the Records Management Program. However, the managers and functional 

organizations that generate the records will be responsible for ensuring compliance with the 

Records Management Program. 

The applicant described the program elements that will be implemented through procedures in 

Sections 11.7.1.1 through 11.7.1.14 of the application (IIFP, 2012a), including provisions for the 

following: 

• legibility, accuracy, and completeness 

• identification of items and activities 

• authentication 

• indexing and filing 

• retention and disposition 

• corrections 

• protection of records 

• storage requirements 

• receipt of records 

• access to records and accountability for removed records 

• records requirements for procured goods or services 

• types of records 

• usage and control of computer codes and data 

• assessment.

11.3.8 Quality Assurance Program Elements 

The provisions provided by the applicant in the QAPD (IIFP, 2012c) apply during the design and 

construction and will remain in effect during operations. The QAPD will be mandatory for items 

identified as IROFS. The LA, Section 11.8, “Quality Assurance Program Elements” 

(IIFP, 2012a), includes a brief description of the QA Program elements that are discussed in 
11-23 

detail in the LA, Appendix A, “Quality Assurance Program Description” (IIFP, 2012c). The 

introduction of Appendix A includes the IIFP commitment to ensure safety for employees and 

the public relative to its facility operations. The NRC staff conducted a review of the QAPD 

(IIFP, 2012c), and found it acceptable. The NRC staff’s report of its review is included in 

Chapter 12 to this SER. 

11.4 Evaluation Findings

11.4.1 Configuration Management 

The NRC staff has reviewed the CM function for IIFP Facility according to Chapter 11 of 

NUREG-1520. The NRC staff found that the applicant’s description of the overall CM program 

appropriately covered CM policy, design requirements, document control, change control, and 

assessments. Based on this evaluation, the NRC staff finds the applicant’s CM Program 

acceptable. 

The applicant has suitably and acceptably described its commitment to a proposed CM system, 

including the method for managing changes in procedures, facilities, activities, and equipment 

for IROFS. Management-level policies and procedures, including an analysis and independent 

safety review of any proposed activity involving IROFS, are described and will provide 

reasonable assurance that consistency among design requirements, physical configuration, and 

facility documentation is maintained as part of a new activity or change in an existing activity 

involving licensed material. The management measures will include the following elements of 

CM: 

• CM Management: The applicant will develop the organizational structure, procedures, 

and responsibilities necessary to effectively implement CM. 

• Design Requirements: The applicant will document the design requirements and bases, 

which will be supported by analyses. The applicant will maintain all design 

documentation current. 

• Document Control: The applicant will appropriately store its documents, including 

drawings, so that they are accessible. Drawings and related documents captured by the 

system will include those necessary and sufficient to adequately describe IROFS. 

• Change Control: Responsibilities and procedures will adequately describe how the 

applicant will achieve and maintain strict consistency among the design requirements, 

the physical configuration, and the facility documentation. Methods will be in place for 

suitable analysis, review, approval, and implementation of identified changes to IROFS, 

including appropriate CM controls. 

• Assessments: The applicant has committed to an adequate function that includes both 

initial and periodic assessments. The assessments will verify and ensure the adequacy 

of the CM function. 

11.4.2 Maintenance 

The applicant committed to maintaining IROFS. The applicant’s maintenance commitments 

contain the basic elements to ensure availability and reliability: surveillance/monitoring, 
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corrective maintenance, PM, functional testing, equipment calibration, and work control for 

maintenance of IROFS. The applicant’s maintenance function is proactive, using maintenance 

records, PM records, and surveillance tests to analyze equipment performance and to seek the 

root causes of repetitive failures. 

The surveillance and monitoring, PM, and functional testing activities described in the LA 

provide reasonable assurance that the IROFS identified in the ISA summary will be available 

and reliable to prevent or mitigate accident consequences. 

The maintenance function (1) will be based on approved procedures; (2) will employ work 

control methods that properly consider personnel safety; awareness of facility operating groups, 

QA, and the rules of CM; (3) will use the ISA summary to identify IROFS that require 

maintenance and determine the level of maintenance needed; (4) will justify the PM intervals in 

terms of the equipment reliability goals; (5) will provide for training that emphasizes the 

importance of IROFS identified in the ISA summary, regulations, codes, and personnel safety; 

and (6) will create documentation that includes records of all surveillance, inspections, 

equipment failures, repairs, and replacements of IROFS. 

The NRC staff concludes that the applicant’s maintenance functions meet the requirements of 

10 CFR Part 70, provide reasonable assurance of public health and safety and the protection of 

the environment, and is acceptable. 

11.4.3 Training and Qualification 

Based on its review of the applicant’s strategy for developing and implementing the training and 

qualification program and a comparison of this strategy to the acceptance criteria guidance in 

NUREG-1520, the NRC staff concludes that the applicant has adequately described and 

assessed its personnel training and qualification in a manner that (1) satisfies regulatory 

requirements and (2) is consistent with the guidance in NUREG-1520. The NRC staff’s 

evaluation found that the applicant’s description of its training program appropriately covered 

(1) organization and management of the training function, (2) analysis and identification of 

functional areas requiring training, (3) position training requirements, (4) training basis and 

objectives, (5) organization of instruction, (6) evaluation of trainee learning, (7) conduct of 

on-the-job training, (8) evaluation of training effectiveness, (9) personnel qualification, and 

(10) periodic personnel evaluation. Based on this evaluation, the NRC staff concludes that the 

applicant has adequately described and assessed its personnel training and qualification in a 

manner that satisfies the regulatory requirements and is consistent with the guidance in 

NUREG-1520. 

There is reasonable assurance that implementation of the described training and qualification 

will result in personnel who are qualified and competent to design, construct, start up, operate, 

maintain, modify, and decommission the facility safely. The NRC staff concludes that the 

applicant’s plan for personnel training and qualification meets the requirements of 

10 CFR Part 70 and are acceptable. 

11.4.4 Procedures 

The application described a suitably detailed process for the development, approval, and 

implementation of procedures. The applicant addresses IROFS, as well as items important to 

the health of facility workers and the public and protection of the environment. The NRC staff 
11-25 

concludes that the applicant’s plan for developing and implementing procedures meets the 

requirements of 10 CFR Part 70 and are acceptable. 

11.4.5 Audits and Assessments 

Based on its review of the safety analysis report and a comparison of the applicant’s plan and 

processes to the review acceptance criteria in NUREG-1520, the NRC staff concludes that the 

applicant has adequately described its audits and assessments. The NRC staff has reviewed 

the applicant’s implementation process for audits and assessments and its description of IIFP’s 

policy directives, plans, and procedural requirements, considering the following: (1) the general 

structure of the audits and assessments program, (2) the activities to be audited or assessed, 

(3) the scheduling of audits and assessments, (4) the procedures for audits and assessments, 

and (5) the qualifications and responsibilities for audits and assessments. 

The NRC staff concludes that the applicant’s plan for audits and assessments meets the 

requirements of 10 CFR Part 70 and provides reasonable assurance of protection of the health 

and safety of the public, workers, and the environment. 

11.4.6 Incident Investigations 

The applicant has committed to and established an organization responsible for (1) performing 

incident investigations of abnormal events that may occur during operation of the facility; 

(2) determining the root causes and generic implications of the event; and (3) recommending 

corrective actions for ensuring a safe facility and safe facility operations, in accordance with the 

acceptance criteria of Section 11.4 of NUREG-1520. 

The applicant has committed to monitoring and documenting corrective actions through to 

completion. The applicant has also committed to taking follow-up actions to verify the proper 

and timely implementation of the corrective actions. 

The applicant has committed to the maintenance of documentation so that “lessons learned” 

may be applied to future operations of the facility. 

Accordingly, the NRC staff concludes that the applicant’s description of the incident 

investigation process complies with applicable NRC regulations and is adequate. 

11.4.7 Records Management 

The NRC staff has reviewed the applicant’s records management system against the 

acceptance criteria and concluded that the system is adequate because it (1) will be effective in 

collecting, verifying, protecting, and storing information about the facility and its design, 

operations, and maintenance; (2) will enable retrieval of the information in readable form for the 

designated lifetimes of the records; (3) will provide a records storage area with the capability to 

protect and preserve health and safety records that are stored there during the mandated 

periods, including protection of the stored records against loss, theft, tampering, or damage 

during and after emergencies; and (4) will provide reasonable assurance that any deficiencies in 

the records management system or its implementation will be detected and corrected promptly. 
11-26 

11.4.8 Other Quality Assurance Elements 

In Appendix A, “Quality Assurance Program Description” to this SER, the NRC staff reviewed 

the QA elements that the applicant presented in the form of a QA Program. The elements 

included were organization; QA program; design control; procurement document control; 

instructions procedures and drawings; document control; control of purchased items and 

services; identification and control of material parts and component; control of special 

processes; inspection; test control; control of measuring and test equipment; inspection, test, 

and operating status; control of nonconforming materials; corrective action; QA records; and 

audits. Based on its review of the LA, the NRC staff concludes that the applicant has 

adequately described the application of the QA elements described in the QA Program (and the 

applicable QA elements of its principal contractors). Appendix A to this SER presents a detailed 

description of the NRC staff’s findings. The NRC staff concludes that the applicant’s use of 

other QA elements described in the QA Program meets the requirements of 10 CFR Part 70 

and provides reasonable assurance of protection of public health and safety and of the 

environment. Specifically, the NRC staff concludes the following: 

• The applicant has adequately established and documented a commitment to an 

organization responsible for developing, implementing, and assessing the management 

measures for providing reasonable assurance of safe facility operations, in accordance 

with the criteria in Section 11.4 of NUREG-1520. 

• 

The applicant has adequately established and documented a commitment to QA 

elements, and the administrative measures for staffing, performance, assessment of 

findings, and implementation of corrective action are in place. 

• The applicant has adequately developed a process for preparation and control of written 

administrative plant procedures, including procedures for evaluating changes to 

procedures, IROFS, and tests. The applicant will implement and maintain a process for 

review, approval, and documentation of procedures. 

• The applicant has adequately established and documented surveillances, tests, and 

inspections to provide reasonable assurance of satisfactory in-service performance of 

IROFS. 

• The applicant will conduct periodic, independent audits to determine the effectiveness of 

the management measures. Management measures will provide for documentation of 

audit findings and implementation of corrective actions. 

• The applicant has adequately established and documented training requirements to 

provide employees with the skills to perform their jobs safely. Management measures 

will provide for the evaluation of the effectiveness of training against predetermined 

objectives and criteria. 

• The organizations and persons performing QA element functions will have the required 

independence and authority to effectively carry out their QA element functions without 

undue influence from those directly responsible for process operations. 
11-27 

• QA elements cover the IROFS, as identified in the ISA summary, and the applicant 

adequately established measures to prevent hazards from becoming pathways to higher 

risks and accidents. 

Accordingly, the NRC staff concludes that the applicant’s use of other QA elements is adequate 

because it meets the requirements of 10 CFR Part 70, the baseline design criteria in 

70.61(a)(1), and provides reasonable assurance of protection of public health and safety and 

the environment. 

11.5 References

(IIFP, 2012a) International Isotopes Fluorine Products, Inc., “Fluorine Extraction Process and 

Depleted Uranium Deconversion Plant (FEP/DUP) License Application, Revision B,” May 2012, 

Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML12123A245. 

(IIFP, 2012b) International Isotopes Fluorine Products, Inc., “ISA Summary Rev. B for IIFP,” 

May 2012, Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML12123A245. 

International Isotopes Fluorine Products, Inc., “Appendix A QAPD Rev B of IIFP License 

Application,” May 2012, Agencywide Documents Access and Management System (ADAMS) 

Accession No. ML12123A245. 

(NRC, 2002) U.S. Nuclear Regulatory Commission, “Standard Review Plan for the Review of a 

License Application for a Fuel Cycle Facility,” NUREG-1520, Rev. 0, March, 2002. 

 
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12-1 

12.0 QUALITY ASSURANCE PROGRAM DESCRIPTION 

12.1 Organization

The International Isotopes Fluorine Products, Inc.’s (IIFP), President is the responsible 

executive for quality assurance (QA), including policies and quality-related goals and objectives. 

However, the application clearly states that all IIFP employees will be responsible for ensuring 

safe facility design and operation and that its products and services meet all necessary 

requirements. During the licensing, design, and construction of the facility, the President of 

International Isotopes, Inc., (INIS) will be acting as the IIFP President. 

The application states that a design and build (DB) contractor will be selected to be responsible 

for detailed design, engineering, procurement, and construction of the facility and will report to 

the Chief Operations Officer (COO) while the Engineering Manager position is being filled. 

Once the Engineering Manager position is filled, the DB Contractor will report directly to the 

Engineering Manager. The DB contractor will be responsible by contract for ensuring that the 

design meets all applicable regulations, codes, and standards. The work under the DB 

contractor will be in accordance with the IIFP’s Quality Assurance Program Description (QAPD) 

(IIFP, 2012c) or an equivalent IIFP-approved program. 

The IIFP President will report to the INIS Chief Executive Officer. The COO will report to the 

IIFP President and will be responsible for managing design, engineering, construction, QA, 

environmental, safety, and health (ESH), and training. The IIFP’s QA Manager will report to and 

support the COO. The QA Manager’s responsibilities will include ensuring compliance with the 

IIFP’s QAPD and procedures; verifying compliance with the QAPD during design and 

construction of the IIFP,s facility; and reviewing the DB contractor’s QA programs in accordance 

with the IIFP QAPD (IIFP, 2012c). 

The applicant described the responsibilities during the project transitions in Section A.1.2 of the 

license application (LA) (IIFP, 2012a). The applicant also established that qualified contractors 

will perform the work in accordance with the IIFP’s QAPD (IIFP, 2012c) or equivalent QA 

programs approved in accordance with written procedures. The IIFP’s QA function will ensure 

that evaluation and preapproval of vendor qualification is performed in accordance with the 

IIFP’s QAPD (IIFP, 2012c) and procedures or that increased inspection requirements are 

implemented where the procurement involves items relied on for safety (IROFS). 

Section A.1.3 of the LA (IIFP, 2012a), also describes responsibilities during the operations 

phase. During operations, the COO will have the overall responsibility for implementation of the 

QA policies and programs by delegated authority from the IIFP President. The IIFP’s Plant QA 

Manager will be responsible for the following: 

• maintaining and complying with the corporate quality management system 

• ensuring that identified project deficiencies are addressed by the appropriate 

management personnel and that containment actions, including interim corrective 

actions, are implemented until permanent corrective action are completed and verified 

• ensuring that periodic audits are conducted to ensure compliance with the QA Program 

and to assess its effectiveness. 
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However, the applicant stated that while providing oversight for the corrective action, the QA 

Manager will not be directly involved in the affected project or production performance 

assignments. 

The IIFP’s President will be ultimately responsible for approval of company policies that could 

impact the quality system but may delegate approval authority to the IIFP’s COO. 

The applicant also described the responsibilities of the Plant Manager and QA lead positions. 

The QA lead position may be designated by the QA Manager to direct the QA effort within the 

production line organization and to exercise the necessary authority to fulfill all organization 

quality requirements. 

12.1.1 Program 

The applicant’s QA Program will be applicable to all IIFP workers at all levels and to all IIFP 

products and services, in accordance with the applicable contract, and at the earliest time 

consistent with the project schedule. The IIFP’s QA Program as described by the applicant sets 

forth the minimum requirements for those items, activities, and services within the scope of the 

IIFP QAPD (IIFP, 2012c) and will be established, maintained, and executed as described by the 

applicant in the QAPD(IIFP, 2012c). Supplementary manuals, procedures, and instructions may 

be implemented for project-specific quality standards and requirements. 

12.1.1.1 Program Basis 

The applicant committed, in Section A.2.1.2 of the LA (IIFP, 2012a), to incorporate into the QA 

Program a graded approach to QA (based on an item’s importance to safety) that will ensure 

compliance with the regulatory requirements in Title 10 of the Code of Federal Regulations

(10 CFR) Part 40. The corporate processes and procedures for the IIFP will be incorporated 

into the IIFP’s QA program. 

12.1.1.2 Quality Assurance Program Implementation 

The applicant commits to implementing the QA Program through policies, procedures, 

instructions, specifications, drawings, procurement documents, contractual documents, among 

other appropriate documents, to ensure that activities within the scope of the QA Program are 

planned, accomplished, and monitored under conditions that ensure the achievement of project 

goals. 

Quality-related activities will be controlled and conducted using documented procedures. The 

documents will provide details needed to accomplish quality-related activities under suitably 

controlled conditions. If work cannot be accomplished as specified in the implementing 

procedures, the applicant requires the work to be stopped until proper corrective action is taken. 

12.1.1.3 Graded Application 

The applicant will perform risk analysis reviews to determine the appropriate elements and 

principles to ensure that the necessary quality-related aspects of the facility are implemented. 

The applicant will take a risk-based, quality approach to determine to what extent the 

requirements of 10 CFR Part 70, Subpart H, apply. 
12-3 

QA levels will be assigned to the different components that are determined to be IROFS. The 

applicant described the three QA levels that will be used in its graded approach as discussed 

below. 

QA Level 1 Requirements

The QA Level 1 requirements will conform to the criteria established in 10 CFR Part 70, 

Subpart H, and will apply to single IROFS preventing or mitigating a high-consequence event. 

All QA requirements will be applied to QA Level 1 IROFS. 

QA Level 2 Requirements

The QA Level 2 requirements will be applied where two or more IROFS are credited to prevent 

or mitigate a high-consequence event, or where any single IROFS (sole IROFS) is credited to 

prevent or mitigate an intermediate-consequence event. The applicant will apply management 

measures and QA Program requirements to QA Level 2 IROFS, as appropriate, to ensure that 

IROFS remain reliable and available. 

As described in Section A.2.1.3.2 of the LA (IIFP, 2012a), the following QA elements will be 

applied to QA Level 1 and QA Level 2 IROFS: design control; procurement control; document 

control; control of purchased items and services; identification and control of materials, parts, 

and components; control of measuring and test equipment; control of nonconforming items; 

corrective actions; and QA records. The remaining QA elements will be applied through a 

graded approach evaluating the factors contributing to the IROFS reliability and considering the 

following: 

• risk significance 

• relative importance to ESH, safeguards, and security 

• applicable regulations, industry codes, and standards 

• complexity or uniqueness of an item/activity and the environment in which it must 

function 

• quality or safety history of the item in service or the activity 

• degree to which functional compliance can be demonstrated or assessed by test, 

inspection, or maintenance methods 

• anticipated life span 

• degree of standardization 

• importance of data generated 

• reproducibility of results. 

The applicant will incorporate results of the graded approach to quality into design requirement 

documents, specifications, procedures, instructions, drawings, inspection plans, test plans, 
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procurement documents, and other documents that establish the requirements for items or 

activities. 

For contractors, the applicant will describe QA Level 1 and QA Level 2 requirements in IIFP 

approved documents. 

QA Level 3 Requirements

The applicant described the QA Level 3 requirements as commercial standard practice. 

12.1.1.4 Indoctrination and Training 

The applicant will provide and document training and indoctrination of personnel performing or 

managing activities affecting quality to ensure that they are knowledgeable about the 

applicability, purpose, scope, and implementation of the quality management system (QMS) and 

the appropriate implementing procedures. Methods for indoctrination may include formal 

classes, supervised on-the-job training and evaluation, or completion of required reading or 

self-study. Personnel will be trained as needed to achieve initial proficiency, maintain 

proficiency, and adapt to changes in the technology, methods, or job responsibility. 

The QA Manager will be responsible for indoctrinating appropriate IIFP management and 

supervisory personnel in the basis for, objective of, and methods for ensuring the quality of 

IIFP’s work. However, the IIFP’s management team will be responsible for working together to 

determine the appropriate methods of indoctrinating and training company personnel. 

12.1.1.5 Quality Improvement 

The applicant stated that managers at all levels will be responsible for creating an atmosphere 

of continuous improvement as an integral part of the work activities. Managers will be expected 

to encourage the development and exploration of new ideas, increase all employees’ 

awareness of the importance of quality, and emphasize the need for enhancing product and 

process safety and reliability. 

The applicant will have measures in place for the detection and prevention of quality problems 

and the verification of the implementation of quality improvements. Sections A.11.1 and A.12.1 

of the LA (IIFP, 2012a) establish the controls for the products, services, and processes that do 

not meet the specified requirements and the corrective actions to be implemented, as 

necessary. 

The QA Manager will establish procedures to periodically perform trend analysis of 

nonconformances and corrective actions. Process trends and lessons learned, which are 

incorporated as a result of audit reports, surveillance reports, corrective action reports, 

management assessments, and other sources, will be reported regularly to management. 

Internal IIFP audits and management reviews will be used to identify opportunities for 

improvement. The Plant Manager will be responsible for managing process quality, identifying 

potential improvements, and emphasizing the responsibility of each project team member to 

understand how their processes contribute to the success of the overall project. 
12-5 

12.1.1.6 Review and Assessment 

Managers of organizations that implement QMS elements will regularly assess the adequacy of 

that part of the program or any other sections assigned to them and will address any audit 

findings in their respective area by implementing appropriate corrective actions and ensuring 

their effective implementation in accordance with applicable procedures. Internal and external 

audits will be the responsibility of the QA Manager and will be performed in accordance with 

Section A.14.1 of the LA (IIFP, 2012a). 

12.1.1.7 Qualification and Certification of Personnel 

The principal objective of the applicant’s training program system is to ensure job proficiency of 

all personnel through effective training and qualification. As described by the applicant, 

qualification will be achieved by successful completion of prescribed training, demonstration of 

the ability to perform assigned tasks, and maintenance of requirements established by the 

regulations and IIFP. The applicant will use a graded approach to systematic training when 

applied to the level of detail needed relative to safety. 

The applicant stated that managers, with the support of the training organization, will have 

responsibility for, and authority to develop and conduct, training for their personnel. The QA 

Manager will be responsible for ensuring that personnel performing quality-related activities are 

adequately trained in activities associated with their work assignment and will periodically 

assess activities to ensure compliance with these QA Program requirements. 

As discussed by the applicant, indoctrination will include the technical objective and 

requirements of the applicable codes and standards and the QA program elements that are to 

be employed. On-the-job training will be a systematic method of providing the required 

job-related skills and knowledge for a position in an environment as close to the work 

environment as feasible. Continuing training, defined as any training not provided as initial 

qualification or basic training and that maintains and improves job-related knowledge and skills, 

will be conducted as required. 

QA/quality control inspections, examinations, surveillances, nondestructive examinations, and 

special processes will be performed by individuals qualified and certified in the discipline and/or 

method in which the activity is being performed. For special processes not covered by existing 

codes and standards or where quality requirements specified for an item exceed those of 

existing codes or standards, the necessary requirements for qualifications of personnel, 

procedures, and equipment will be specified or referenced in the procedures or instructions. 

12.1.1.8 Work Control 

The applicant stated that the QAPD will establish the requirements and define the procedures 

for controlling work activities for maintenance and future projects after the facility operations 

begin. This will ensure that the work activities will comply with the requirements of both the 

applicable contract and the QAPD itself (IIFP, 2012c). As described by the applicant, the 

degree of complexity and detail in instructions and procedures will be commensurate with the 

risk associated with the work being performed and the specific customer requirements. 

Managers will have the responsibility to ensure that adequate controls are established over 

activities affecting quality and that personnel are properly trained and qualified and have the 

proper tools available before performing the work. Managers will also be involved in the work 
12-6 

and the work processes to stay current and to create an environment that encourages 

employees to improve the quality of the work and work processes. The applicant stated that 

each individual must focus on his or her specific tasks and take responsibility for the quality of 

the work performed in order to meet work performance objectives and expectations. The Plant, 

QA, and ESH Managers are responsible for review and approval of the new procedures and 

procedural revisions before their implementation. 

The applicant will conduct activities involving licensed materials or IROFS in accordance with 

approved procedures that are reviewed by the line managers, QA Manager, production staff, 

and the radiation protection staff to ensure that the activities are carried out safely and in 

accordance with regulatory requirements. Integrated safety analysis (ISA) results will also be 

considered in the identification of needed procedures. The applicant will use standardized 

methods for the identification, development, approval, implementation, and control of operating 

procedures. The applicant will also develop procedures for specific products in accordance with 

the requirements of the applicable portions of the QAPD (IIFP, 2012c). 

The U.S. Nuclear Regulatory Commission (NRC) staff finds the applicant’s QAPD (IIFP, 2012c) 

program acceptable because it complies with 10 CFR Pat 40, and it is implemented through 

procedures and policies. The NRC staff also determined that the personnel will be properly 

trained and qualified to engage in work activities. 

12.1.2 Design Control 

12.1.2.1 General 

The design control provisions contained in the applicant’s QAPD (IIFP, 2012c) will be applicable 

to activities during design, construction, and operation of the facility. The applicant stated that 

reconstitution of any prior conceptual design will not be required, but any deviation discovered 

from the design will be resolved by the Engineering Department. 

The applicant will provide requirements and controls to ensure that (1) new design and design 

change activities will be carried out in a planned, controlled, and orderly manner; (2) design 

basis, regulatory requirements, and appropriate quality standards are correctly translated into 

design output, procurement requirements, and procedural documents; and (3) provisions will be 

established for verifying or checking the technical adequacy of design documents including 

computer codes. 

12.1.2.2 Responsibilities 

The Engineering Manager will be the design authority having responsibility for the 

implementation and execution of the design control system. The COO, with appropriate 

delegation of authorities to the DB Design Engineering Manager, will be the design authority 

with responsibility for implementation and execution of the design control system. Management 

will be responsible for ensuring that completed plant changes are tested and that personnel 

affected by the change are adequately trained as described in procedures. 
12-7 

12.1.2.3 Implementation 

12.1.2.3.1 Design Inputs 

Design inputs are those criteria, parameters, or other design requirements upon which the final 

design is established. Design inputs will be identified, documented, reviewed and approved by 

the responsible design organization. As described by the applicant, applicable design inputs will 

be appropriately specified, correctly translated into design documents, and approved on a timely 

basis and to the level of detail necessary to permit the design activity to be carried out correctly 

and to provide a consistent basis for making design decisions, accomplishing design verification 

measures, and evaluating design changes. 

Changes to approved design inputs will include the reason for the changes, and they will be 

identified, approved, documented, and controlled before implementation. 

12.1.2.3.2 Design Process 

Design methods, materials, parts, equipment, and processes essential to the function of the 

IROFS will be selected and reviewed for suitability of the application. The applicant committed 

to having appropriate design documents to support facility design, construction, and operation. 

Appropriate quality standards and changes to design documents will be identified and 

documented and their selection reviewed, approved, and controlled before implementation. 

The outputs of design and development will be provided in a format that enables verification 

against the design and development input requirements. Design and development outputs shall 

meet the input requirements for design and development; provide appropriate information for 

purchasing, production, and service provision; contain or reference product acceptance criteria; 

specify the characteristics of the product that are essential for its safe and proper use; and be 

reviewed and approved before design release. 

12.1.2.3.3 Design Analysis 

The applicant will perform design analyses in a planned, controlled, and documented manner. 

The design analyses documents will contain detailed description that will be sufficient to define 

the purpose, method, assumptions, design input, references, and units, such that a person 

technically qualified in the subject can review and understand the analyses and verify the 

adequacy of the results without recourse to the originator. Calculations within the design 

analysis documents will be identifiable by subject, originator, reviewer, and date; or by other 

data so that the calculations are retrievable. 

12.1.2.3.4 Design Verification 

The applicant stated that during the design/construction phase, the design requirements and 

associated design basis will be established and maintained by the design engineering 

organization (designated by the Configuration Management Lead and approved by the COO). 

After operations begin, the Plant Engineer will be responsible for designation and approval of 

design requirements and documents. The design requirements and basis of design documents 

will be controlled under the design control provisions of the configuration management program. 

Design documents associated with QA Level 1 or QA Level 2 IROFS will be subject to 

interdisciplinary reviews and design verification. The applicant will evaluate changes to the 

design to ensure consistency with the design basis. In addition, the applicant stated that the 
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computer codes that will be used in the design of IROFS will also be subject to these design 

control measures, with additional requirements as appropriate for software control, verification, 

and validation. 

Qualified individuals will prepare the design documents and will specify and include the 

appropriate codes, standards, and licensing commitments within the design documents. They 

will also note any deviations or changes from such standards within the design documentation 

package. Another individual qualified in the same discipline will review each design document 

for concept and conformity with the design inputs. The manager with overall responsibility for 

the design function will approve the document, and the Configuration Manager will ensure that 

the designated engineering organization documents the entire review process in accordance 

with approved procedures. 

The applicant described the check and review process consistent with the description given in 

the application. The applicant discussed the independent design verification process and its 

requirements. The applicant requires independent design verification whenever appropriate; 

however, if the verification cannot be performed, the unverified portion of the document will be 

controlled and the design verification will be completed before the item is relied on to perform its 

functions. Any changes to the design and procurement documents will be reviewed, checked, 

and approved commensurate with original requirements. 

Any deficiencies identified by the applicant affecting the design of IROFS will be documented 

and resolved in accordance with approved corrective action program procedures. Design 

interfaces will be maintained by communication among principals. During the operational 

phase, the applicant will provide measures to ensure that responsible facility personnel are 

made aware of design changes and modifications that may affect the performance of their 

duties. 

12.1.2.3.5 Design Changes 

The applicant will use procedures for controlling design, including the processes of preparation, 

review, design verification, approval, and release and distribution for use. Engineering 

documents will be assessed based on the QA level classification of the item being reviewed. In 

addition, changes to the approved design will be subject to a review to ensure consistency with 

the design basis of IROFS. 

The procedures, as required by the QA Program, will specify that work will be performed in 

accordance with the requirements and guidelines imposed by applicable specifications, 

drawings, codes, standards, regulations, QA criteria, and site characteristics. These 

procedures, instructions, and drawings will also include any acceptance criteria established by 

the applicant. 

The applicant will maintain documentation demonstrating that the work has been properly 

performed. 
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12.1.2.3.6 Design Interfaces 

The applicant will identify, document, and control internal and external design interfaces. Any 

design change efforts will be coordinated among participating organizations. The 

responsibilities for the preparation, review, approval, release, distribution, and revision of 

documents involving design changes will require cross-functional team evaluation and must 

follow the standard document and configuration control requirements. 

12.1.2.3.7 Design Documentation and Records 

As stated by the applicant, design documentation and records will provide evidence that the 

design and design verification processes were performed in accordance with the program 

requirements. The documentation and records will be collected, stored, and maintained in 

accordance with documented retention policies and procedures. Before modification or 

extensive repair of equipment or systems, configuration management requirements must be 

followed to ensure that the planned, as-built condition is documented correctly and completely 

shown on the drawings, specifications, engineering change notices, and other descriptions of 

equipment and systems before implementation of the change. 

The NRC staff finds that the design control of the proposed facility is adequate because 

implementation of design inputs, design process, design analysis and design verification will be 

reviewed, approved and documented The NRC staff determined that changes during the 

design will be consistent with approved procedures. 

12.1.3 Procurement Document Control 

12.1.3.1 General 

The applicant stated that the procurement system requirements (1) ensure that applicable 

regulatory requirements, drawing and technical requirements, along with QA Program 

requirements, are included or referenced in procurement documents for the procurement of 

items and services for QA Level 1 and QA Level 2 control items; and (2) establish provisions for 

the preparation, review, approval, and control of procurement documents, including changes. 

12.1.3.1.1 Responsibilities 

The Engineering Department will be responsible for (1) the preparation and maintenance of 

design specifications and for identifying the technical and quality requirements necessary to 

ensure item acceptability; and (2) development of procedures that define these activities, 

including the criteria for developing the necessary technical and quality requirements for 

procurement. 

12.1.3.1.2 Procurement Document Contents 

The following are requirements for procurement documents as described by the applicant in 

Section A.4.1.3.1 of the LA (IIFP, 2012a): 

• statement of work for procurement of services, or an engineering specification for the 

procurement of items for QA Level 1 and QA Level 2 items 
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• technical requirements by specific reference to drawings, specifications, codes, 

standards, regulations, procedures, or instructions, including revisions, that describe the 

items or services to be furnished for QA Level 1 and QA Level 2 items 

• any special instructions and requirements for designing, fabricating, erecting, packaging, 

shipping, handling, storing, testing, inspecting, and accepting, if required 

• requirements for suppliers to have a documented QA program consistent with the 

applicable requirements of the applicant’s QA Program and other applicable codes and 

standards and commensurate with the item being procured and its importance to safety 

• requirements for suppliers of noncommercial grade items and services to evaluate their 

lower tier suppliers that provide IROFS items or services within the scope of the 

statement of work or engineering specification 

• documentation required to be submitted for information, review, or approval, as well as 

the time of submittal, where applicable 

• requirements for reporting and obtaining disposition of nonconforming items and 

services, as appropriate. 

12.1.3.1.3 Procurement Document Review 

The applicant committed to performing document reviews to provide objective evidence of 

satisfactory accomplishment before contract award. If any changes are made as a result of the 

bid evaluation or precontract negotiations, those will be incorporated into procurement 

documents before contract award. Appropriate personnel who have access to pertinent 

information, and have an adequate understanding of the requirements and procurement 

documents, will perform the reviews and approvals. 

The NRC staff finds the applicant’s procurement and document control acceptable because a 

statement of work is required for procurement services and technical requirements will be 

referenced to drawings or regulations. Moreover, suppliers to IIFP will have a documented QA 

program that meets the applicant’s QA Program. 

12.1.4 Instructions, Procedures, and Drawings 

12.1.4.1 General 

The applicant described the requirements for instructions, procedures, and drawings as those 

applied to quality- and process-related activities and services. The applicant will have 

measures in place to ensure that activities affecting quality are prescribed by documented 

procedures, drawings, and instructions, appropriate to the circumstances, and are accomplished 

in accordance with these documents referencing quantitative and qualitative acceptance criteria. 

12.1.4.2 Responsibilities 

The IIFP’s President will be responsible for approving all policy-level documents. The Training 

Manager or QA Manager will be responsible for reviewing the approval and use of procedures 

and instructions in accordance with the management oversight requirements defined in 
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Section A.5..2 of the LA (IIFP, 2012a). The Engineering Department will be responsible for the 

system of preparation, review, and approval of drawings. 

Line and functional managers will be responsible for the development and approval of 

procedures controlling functions or activities within their area of responsibility. It will be the 

responsibility of all personnel to use and adhere to the requirements of applicable procedures, 

instructions, and drawings for activities. 

12.1.4.3 Implementation 

The NRC staff finds the applicant’s implementation of instructions, and procedures acceptable 

because activities requiring skills normally possessed by qualified personnel are performed in 

accordance with work instructions, procedures, or drawings of a type appropriate to the 

circumstances for the control of maintenance and modification work. For activities known as 

“skill-of-the-craft,” the applicant will not require detailed, step-by-step procedures. Written 

procedures for all of these activities will be prepared, reviewed, approved, implemented, and 

maintained in accordance with the IIFP’s document control process. 

12.1.5 Document Control 

12.1.5.1 General 

The applicant will establish a document control system to maintain policies, procedures, work 

instructions and any other documentation-related activities and services provided by the 

applicant to ensure that documents defining the performance of processes and quality-related 

activities will be controlled in order to maintain only current and correct information available at 

the location where the activity will be performed before the start of the work. 

12.1.5.2 Responsibilities 

The records/document lead will have the overall responsibility for the development and 

implementation of the records management and document control system. Managers will be 

responsible for (1) identifying documents to be included in the controlled document system; 

(2) ensuring that instructions, procedures, drawings, and other specified documents are 

reviewed for adequacy and approved for release; (3) complying with document distribution 

requirements; and (4) ensuring that personnel performing the prescribed activities maintain and 

use these documents. 

12.1.5.3 Implementation for Document Control 

The applicant will establish procedures for the control of document preparation, review, 

approval, and issuance to ensure that identification of documents will be controlled. In addition, 

the procedures will include provisions for (1) specific distribution; (2) identification of 

responsibilities for preparing, reviewing, approving, and issuing documents; and (3) review of 

documents for adequacy, completeness, and correctness before approval and issuance. The 

applicant will remove and/or replace obsolete or superseded documents in a timely manner. 

The applicant will prepare drawings depicting as-built conditions, including changes, and related 

documentation in a timely manner reflecting the actual design. The applicant will identify 

controls to specify the current revision and any changes to instructions, procedures, 
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specifications, drawings, and procurement documents. The applicant’s document control 

system will also have provisions for update and distribution to predetermined personnel. 

Changes to documents, other than minor changes, will be reviewed and approved by the same 

organization that performed the initial review and approval or will be delegated to other qualified 

organizations. Minor changes to documents will not require that the revised documents receive 

the same review and approval as the original documents. Procedures will specify the review 

and approval process for minor changes. Therefore, the NRC staff finds the document control 

program adequate. 

12.1.6 Control of Purchased Items and Services 

12.1.6.1 General 

In Section A.7.1.1 of the application, the applicant committed to establishing a system for the 

control of purchased items and services. 

12.1.6.2 Responsibilities 

The QA Manager will be responsible for providing the necessary QA functions to support 

procurement, including (1) reviewing supplier quality documentation and evaluating the 

supplier’s QA capability, supplier audits, and development and maintenance of the approved 

suppliers list; (2) providing support functions such as source verification or surveillance, receipt 

inspections, installation inspections, and review of procurement documents during receipt 

inspections; and (3) developing and implementing procedures. 

The Engineering Department will assist the QA Manager by performing evaluations of supplier 

technical capabilities and determining the methods of acceptance to be applied to purchased 

items and services. The Engineering Department or QA will be responsible for the approval of 

dispositions and technical evaluations for supplier-generated nonconformances for items and 

services and for providing measures that ensure the proper selection, application, methods of 

acceptance, and use of items. 

12.1.6.3 Implementation 

12.1.6.3.1 Procurement Planning 

The applicant will plan and document procurement activities to ensure a systematic approach to 

the procurement process that will result in the documented identification of procurement 

methods and organizational responsibilities. 

12.1.6.3.2 Supplier Selection 

As described by the applicant, suppliers will be selected in accordance with procedures, and 

selection will be based on evaluation of the supplier’s capability to provide items or services in 

accordance with the requirements of the procurement documents before award of contract. The 

applicant described four methods to assess the supplier’s technical and quality capabilities; 

upon acceptance, the suppliers are placed on the applicant’s Approved Suppliers List. The 

methods described are the following: 
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• Evaluation of the supplier’s history of providing an identical or similar product that 

performs satisfactorily in actual use. The supplier’s history will reflect current 

capabilities. 

• The supplier’s technical and quality capability will be determined by an evaluation of its 

facility and personnel and the implementation of the supplier’s QA program. 

• The supplier implements a QA program accepted by the International Standardization 

Organization (ISO). 

• The supplier maintains a valid ISO certification for the item or service being provided. 

12.1.6.3.3 Source Verification 

Source verification, when used as acceptance, will be performed at intervals consistent with the 

importance to safety and the complexity of the item or service, and it will be implemented to 

monitor, witness, or observe activities. This method provides plans to perform inspections, 

examinations, or tests at predetermined points. Upon purchaser acceptance of source 

verification, documented evidence of acceptance will be furnished to the receiving destination of 

the item, to the purchaser, and to the supplier. 

12.1.6.3.4 Receiving Inspection 

The applicant stated that receiving inspection will be performed for all purchased items to verify 

conformance to procurement documents. Receiving inspection will verify by objective evidence 

features as proper configuration; identification; dimensional, physical, or other characteristics; 

freedom of damage from shipping; cleanliness; and review of supplier documentation when 

procurement documents require the documentation to be furnished. Upon completion of receipt 

inspection, acceptable items will be released for storage or issued for installation or use. If any 

items are determined to be nonconforming after completion of the receipt inspection, those will 

be documented and processed. 

12.1.6.3.5 Post Installation Testing 

If post installation testing is used for acceptance of noncommercial-grade items, the purchaser 

and supplier will establish post installation test requirements and acceptance documentation. 

12.1.6.3.6 Commercial-Grade Items 

The applicant committed in Section A.7..3.3 of the LA (IIFP, 2012a) to establishing a method for 

determining whether an item can be purchased as commercial grade and dedicated for use in 

an IROFS application. The applicant stated that the criteria and methods selected will identify 

the critical characteristics that are essential to ensure that the item will perform its intended 

IROFS function. The applicant stated that Quality Level 1 and Quality Level 2 IROFS may be 

bought commercially, provided that they are subject to a dedication process. Moreover, the 

applicant stated that in cases where commercial-grade items are to be procured and then 

dedicated for use as IROFS or parts thereof, the procurement process procedures will include 

the requirements that IIFP will define to the supplier. Requirements for verifying acceptability of 

critical characteristics are also mentioned and could encompass inspection, tests, or analyses 

after delivery, supplemented as necessary by commercial-grade surveys; product inspections or 
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witness at hold points at the manufacturer’s facility; and analysis of historical records for 

acceptable performance. 

As described by the applicant, as a minimum for acceptance of commercial-grade items, a 

receipt inspection will be performed to (1) determine that damage was not sustained during 

shipment, (2) ensure that the item received is the item ordered, (3) determine that inspection 

and testing were performed by the supplier as required by Engineering, (4) ensure conformance 

with the manufacturer’s published requirements, and (5) ensure that required documentation is 

received and is acceptable. 

Commercial-grade items will be identified as such in the contract or purchase order by the 

manufacturer’s published product description. Alternative commercial-grade items will be 

allowed, provided that the Engineering Department verifies that the alternative 

commercial-grade item will perform its intended IROFS function. 

The NRC staff finds the applicant’s purchase of services and items acceptable because the 

supplier’s history will be evaluated along with its technical and quality capability. In addition, the 

applicant will verify purchased items to ensure they meet all procurement documents. 

12.1.7 Identification and Control of Materials, Parts, and Components 

12.1.7.1 General 

The applicant committed in Section A.8.1 of the LA (IIFP, 2012a) to establishing a system for 

the identification and control of IROFS within the scope of the QAPD (IIFP, 2012c), including 

associated materials, parts, spare parts, components, and subassemblies. 

12.1.7.2 Responsibilities 

The Engineering Department will be responsible for (1) specifying requirements for identification 

methods, traceability, shelf life, and operating life of items when required by codes, standards, 

or specifications; and (2) specifying these requirements during the generation of specifications, 

drawings, procurement documents, or other documents appropriate to the circumstances. 

Managers will be responsible for maintaining and implementing identification, traceability, and 

shelf life and operating life requirements for items under their jurisdiction. In addition, the 

Purchasing Manager will be responsible for receipt, delivery, storage, traceability, identification, 

and control of materials. The QA Manager will be responsible for verifying that items are 

correctly identified through receipt inspection. 

12.1.7.3 Implementation 

Procedures will be established depicting the requirements to be implemented for the 

identification, traceability, and control of materials, parts, and components, including partially 

fabricated assemblies or subassemblies. Some of the requirements that the applicant will 

establish in these procedures will include those described below. 

• Where practical and required by codes, standards, or contractual documents, items or 

documents traceable to them will be identified in a manner that ensures that their 

identification is established and maintained. 
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• When physical identification is either impractical or insufficient to control the item, 

physical separation, procedural controls, or other means will be employed. 

• Measures will be taken to prevent the use of defective, unapproved, incorrect, or 

incomplete materials or equipment. These measures will preclude the use of items 

whose shelf life or operating life has expired. 

• Items will have unique identification and traceability by serial number, part number, 

batch, lot, or specified inspection, test, records, or other appropriate means. 

• Productions of an item at any stage, from initial receipt through fabrication, installation, 

repair, modifications, and use, will be traceable to records such as applicable drawings, 

specifications, purchase orders, manufacturing and inspection documents, deviation 

reports, certified material test reports, or other pertinent applicable design-specifying 

documentation. 

• Permanent physical identification will be placed on the item itself to the maximum extent 

possible, in a manner and location that will not impair or negate the item’s intended use, 

quality, function, or service life. 

• Physical separation, procedural control, or other appropriate means will be used where 

physical identification on the item is impractical or not sufficient. 

• Correct identification of materials, parts, and components will be verified and 

documented on appropriate release documents, work packages, or controlling 

documents, and on materials before subdividing an item or material and before release 

for fabrication, assembly, shipping, and installation. 

• Personnel performing quality activities will receive training, as required, to ensure 

understanding and proper implementation of the applicant’s QA Program and use of 

approved procedures to ensure that improper, uncontrolled, damaged, incorrect, or 

nonconforming material or items are not used or installed. 

• Audits, inspections and surveillances will ensure compliance with established 

procedures. 

• Documents establishing and attesting to proper identification of IIFP-furnished items will 

be maintained as quality records in accordance with design, procurement, and process. 

The NRC staff determined that the applicant’s control of parts and materials is adequate 

because procedures to prevent the use of equipment are implemented. Items will have 

traceability that enables the applicant to determine whether the item is appropriate with its 

intended usage. 

12.1.8 Control of Special Processes 

12.1.8.1 General 

The applicant stated in Section A.9 of the LA Appendix A (IIFP, 2012a) that a system for the 

control of special processes and its requirements will be established. 
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12.1.8.2 Responsibilities 

The Engineering Department will be responsible for determining special processes, and the QA 

Manager will be responsible for the qualification of nondestructive examination personnel, 

including welder/brazing qualifications. Line managers have the responsibility of ensuring that 

identified special processes are performed by qualified personnel, using qualified and approved 

procedures or documents of a type appropriate to the circumstances. 

 

12.1.8.3 Implementation 

The applicant will control special processes affecting the quality of items and services such as 

welding, heat treating, and nondestructive examination through policies, plans, procedures, 

instructions, drawings, checklists, travelers, work orders, or other appropriate means. 

Procedures will ensure that special process parameters are controlled and specified 

environmental conditions are maintained. 

The NRC finds the control of special processes adequate as quality processes, special 

processes will be performed by qualified personnel using qualified equipment when appropriate 

and approved written policies, plans, and/or procedures in accordance with specified 

requirements, codes, or standards. 

Records of qualifications and equipment related to special processes will be maintained. 

12.1.9 Inspection 

12.1.9.1 General 

The applicant will establish a system for inspection of IROFS that will provide measures to 

ensure that maintenance, repair, or modification work is completed satisfactorily. In addition, 

the applicant will identify the requirements for the certification of personnel who perform 

inspection, examination, surveillance, and testing. 

12.1.9.2 Responsibilities 

The QA Manager will be responsible for (1) inspection planning, (2) ensuring that inspections 

are performed, and (3) using qualified and certified inspection personnel, while the engineering 

or quality organization will be responsible for specifying “hold” and “witness” points for inclusion 

in applicable work control documents. Management will establish the appropriate measures to 

ensure that all requirements for inspection are met. 

12.1.9.3 Implementation 

As stated by the applicant, inspections will be planned and executed to verify conformance of an 

item or activity to specified requirements. The characteristics to be inspected and inspection 

methods to be employed will be specified. Inspection personnel will not report directly to the 

immediate supervisors who are responsible for performing the work being inspected. The depth 

and extent of inspections will be established by the Plant Engineering & Maintenance Manager 

and the quality history of the process and will be commensurate with the significance and 

complexity of the IROFS. 
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The applicant will document inspection results and records to identify the item inspected, date of 

inspection, inspector, type of observation, results or acceptability, and references to information 

on action taken in connection with nonconformances. The applicant will plan for inspection 

activities, including the documentation of the identification of characteristics, methods, and 

acceptance criteria. Documentation will also provide for recording objective evidence of 

inspection results. 

The applicant will inspect items in process or under construction to verify quality for certain work 

activities where necessary. If mandatory inspection hold points are required beyond which work 

shall not proceed without the specific consent of the designated representative, appropriate 

documents will indicate the specific hold points; and any waiver will be recorded before the 

continuation of work beyond the designated hold point. 

In-service inspection or surveillance of structures, systems, or components (SSCs) will be 

planned and executed by or for the organization responsible for construction or operation as 

specified in the QAPD (IIFP, 2012c) or the ISA. The inspection methods, including evaluation of 

the performance capability of essential emergency and safety systems and equipment, 

verification of calibration and integrity of instruments and instrument systems, and verification of 

maintenance, as appropriate, will be established and executed to verify that the characteristics 

of an item continue to remain within specified limits 

The applicant will plan final inspections to arrive at a conclusion regarding conformance of the 

item to specified requirements. As described by the applicant, during these final inspections, 

(1) a records review will be performed to check the results and resolution of nonconformances 

identified by prior inspections; (2) completed items will be inspected for completeness, 

markings, calibrations, adjustments, protection from damage, or other characteristics as 

required to verify the quality and conformance of the item to specified requirements; (3) quality 

records will be examined for adequacy and completeness, if not previously examined; (4) the 

acceptance of the item will be documented and approved by authorized personnel; and 

(5) modifications, repairs, or replacement of items performed subsequent to final inspection will 

be required to undergo reinspection or retest, as appropriate, to verify acceptability. 

The NRC staff finds the applicant’s inspections of IROFS adequate because inspections will be 

carried out by approved procedures and the findings will be documented. The applicant will 

ensure actions are taken with respect to nonconformance to prevent any defective item from 

being used. 

12.1.10 Test Control 

12.1.10.1 General 

The applicant will establish a system for design verification testing, acceptance testing, 

preoperational and operational testing, post maintenance testing, and special testing of IROFS 

to ensure that testing is completed satisfactorily. Personnel performing design verification 

testing, acceptance testing, preoperational and operational testing, post maintenance testing, 

and special testing of IROFS will be certified. 
12-18 

12.1.10.2 Responsibilities 

The Engineering Department will be responsible for providing technical criteria for testing, 

evaluating test results, and resolving deficiencies identified from these tests, while line 

managers are responsible for the conduct of testing activities under their cognizance. 

12.1.10.3 Implementation 

The applicant will require tests for verification of conformance to specified requirements and to 

demonstrate satisfactory performance of services to be planned and executed. As described by 

the applicant, tests may include design verification tests, acceptance tests, preoperational and 

operational tests, post maintenance tests, and special tests. 

The NRC staff finds the test control to be acceptable because the applicant will identify specific 

characteristics to be tested and test methods to be employed, and test results will be 

documented. The application specifies that planning for tests may include mandatory hold 

points, as required. In addition, the applicant listed the information that will be contained, as 

appropriate, within the test policies, plans, and procedures, including purpose or objective, 

references, any required provisions, and qualifications. In lieu of test policies, plans, and 

procedures, the applicant will use—when needed—appropriate methods taken from related 

documents, such as external manuals, maintenance instructions, or approved drawings—

among others—provided that those include adequate instructions to ensure the required quality 

of work. The applicant described the contents of the test record information as item tested; test 

date; tester or data recorder; type of observation; test policy, plan, procedure, or reference; 

results and acceptability; actions taken in connection with any deviations noted; and the person 

evaluating the results. 

12.1.11 Control of Measuring and Test Equipment 

12.1.11.1 General 

The applicant will establish a system for the control of measuring and test equipment (M&TE) 

used for measurement, test, and calibration items within the scope of the QA Program. The 

system will include requirements that will establish measures to (1) ensure that tools, gauges, 

instruments, reference and transfer standards, nondestructive test equipment, and other 

measuring and testing devices used in activities affecting quality will be properly controlled, 

calibrated, and adjusted at specified intervals to maintain equipment performance within 

required limits; and (2) ensure that devices and standards used for measurement, tests, and 

calibration activities will be of the proper type, range, accuracy, and tolerance to determine 

conformance to specified requirements. 

12.1.11.2 Responsibilities 

The Plant Engineering & Maintenance Manager will have the overall responsibility for the 

calibration control system for M&TE. In addition, individual area managers will be responsible 

for the implementation of the calibration control system for M&TE under their respective areas. 

12.1.11.3 Implementation 

The applicant will identify a list of devices within the calibration system including, as a minimum, 

the due date of the next calibration and any use limitations. The method and interval for 
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calibration for each item will be defined, based on the type of equipment, stability 

characteristics, required accuracy, intended use, and other conditions affecting measurement 

control. The applicant clarified that calibration controls will not be required for rulers, tape 

measures, levels, and other such devices if the commercial equipment provides adequate 

accuracy. 

M&TE will be properly handled and stored to maintain accuracy and will be calibrated at 

specified intervals or before use against certified equipment having known valid relationships to 

nationally recognized standards. If no nationally recognized standard exists, the bases for 

calibration will be documented. Also, calibrations may be performed when the accuracy of the 

equipment is deemed suspect by personnel performing measurements and tests. Records will 

be maintained, and equipment will be appropriately marked to indicate its calibration status. 

If any M&TE is found to be out of calibration, the applicant will evaluate and document the 

validity of any previous inspection and test results and the acceptability of items previously 

inspected or tested. These devices will be tagged or segregated and will not be used until 

recalibrated. If consistently found to be out of calibration, the M&TE will be repaired or 

replaced. 

The NRC staff finds M&TE acceptable because the NRC staff will have a scheduler with 

information on when the next calibration is due, and the equipment will be properly handled to 

ensure its purpose is not affected. 

12.1.12 Inspection, Test, and Operating Status 

12.1.12.1 General 

The applicant requires the identification of the status of inspection and test activities. In 

addition, status indicators will provide for the operating status of IROFS systems and 

components to prevent inadvertent operation. 

12.1.12.2 Responsibilities 

The QA Manager is responsible for providing a status-indicating system for inspections that will 

be performed in accordance with these requirements, while line managers are responsible for 

the development and implementation of status-indicating systems. 

12.1.12.3 Implementation 

The NRC staff finds that inspections, tests and operating status are adequate because the 

applicant will establish policies, plans, and procedures to ensure that the status of inspection 

and test activities are either marked or labeled on the item or in documents traceable to the 

item. Status indicators are used, as appropriate, to prevent inadvertent operation. Only the 

persons specified by the applicant will authorize the application and removal of tags, markings, 

labels, and stamps. 
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12.1.13 Control of Nonconforming Items 

12.1.13.1 General 

The applicant will establish procedures for the control of items that do not conform to specified 

conditions and related activities and services. These will include requirements for identification, 

segregation, disposition, prevention of inadvertent installation or use, documentation, and 

notification to affected organizations. 

12.1.13.2 Responsibilities 

The applicant affirmed that all personnel participating in quality-affecting activities within the 

scope of the quality system will be responsible for reporting and documenting nonconforming 

items 

The QA Manager will be responsible for implementing the nonconformance control system for 

items that do not meet the established specifications or technical requirements. The 

Engineering Department, ESH, and the quality organization will provide documented technical 

justification for the acceptability of nonconforming items (use-as-is or repair). These 

organizations will also be responsible for applying the design control measures of the QAPD 

(IIFP, 2012c) to nonconformances. The Engineering Department will also ensure that as-built 

records reflect the accepted change. 

12.1.13.3 Implementation 

Nonconforming items will be legibly identified in a manner that does not adversely affect the use 

of the item. If physical identification of the item is not practical, the container, package, or 

segregated storage area will be identified appropriately. Nonconforming items will be 

segregated, when practical, by placing them in a clearly identified and designated area until 

disposition. If segregation is impractical or impossible, other measures will be employed to 

prevent inadvertent use of the item. 

Any nonconforming characteristics will be reviewed and disposition will be recommended based 

on the evaluation. The applicant will define in the appropriate documentation the applicable 

responsibilities and authorities. Specifically, personnel performing evaluations will have an 

adequate understanding of the requirements and have access to pertinent background 

information. Repaired or reworked items will be reexamined in accordance with applicable 

procedures and original acceptance criteria, unless the nonconforming item disposition has 

established alternate acceptance criteria. Further processing, delivery, installation, or use of the 

nonconforming item will be controlled pending an evaluation with an approved disposition and 

an approved notification to affected organizations. 

Nonconforming items or services identified by suppliers will be reviewed to determine 

applicability. The NRC staff finds that identification of nonconforming items is acceptable 

because corrective actions will be initiated which will be evaluated to determine if reporting is 

required. When a nonconformance is identified after the product is shipped to a customer or 

recipient, IIFP management will evaluate whether the nonconformance is potentially reportable 

to the NRC under 10 CFR Part 21. In addition, the customer or recipient will be notified. 
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12.1.14 Corrective Action 

12.1.14.1 General 

The applicant described the corrective action system as a system established for those activities 

and services that are determined to have an adverse effect on the customer or have the 

potential for recurrence. The applicant stated that the corrective action system will (1) establish 

measures to ensure that conditions adverse to quality are identified and corrected as soon as 

practical, (2) ensure that a root cause analysis and corrective action report will be generated to 

document and report the analysis and action to the appropriate levels of management, and 

(3) ensure that follow-up actions are taken to verify implementation of the corrective action. 

12.1.14.2 Responsibilities 

The QA Manager will be responsible for maintenance and implementation of the corrective 

action system and for verifying that management reviews and assesses adverse conditions. 

Moreover, the QA Manager will be responsible for (1) reviewing deficiencies identified by a 

customer audit or assessment; (2) maintaining an orderly file of all audit findings and 

nonconformances related to products and services; (3) ensuring that the requirements of the 

customer and of the procedure are met, as a minimum; (4) evaluating any corrective action that 

results from suggestions by the customer; (5) ensuring that customer suggestions are 

appropriately addressed either by implemented change or resolution; and (6) reviewing and 

evaluating the effectiveness of corrective actions. 

Line managers will be responsible for (1) evaluating and performing assigned corrective actions; 

(2) ensuring the identification and documentation of conditions adverse to quality promptly in 

accordance with applicable procedures; and (3) determining the cause of any deficiency, 

documenting it as a nonconformance, determining the necessary action, and reporting actions 

in a nonconforming report or in the corrective action report. 

12.1.14.3 Implementation 

The applicant committed, in Section A.16.1 of the LA Appendix A (IIFP, 2012a), to promptly 

identifying, documenting, reporting, and correcting conditions adverse to quality. The applicant 

defined “conditions adverse to quality” as failures, malfunctions, deficiencies, deviations, 

defective material and equipment, and nonconformances. 

In addition, the applicant defined “significant conditions adverse to quality” as those for which 

the cause of the condition will be determined and appropriate corrective actions will be taken to 

preclude recurrence. The applicant will evaluate significant conditions adverse to quality to 

determine if stopping work is warranted. In addition, the identification, cause, and corrective 

action for significant conditions adverse to quality will be documented and reported to 

appropriate levels of management. Follow-up actions will verify the implementation of the 

corrective action. 

The applicant defined conditions adverse to quality as follows: 

• a deficiency that would seriously impact an item, activity, or service from meeting or 

performing its intended function or output of ensuring public health and safety 
12-22 

• a deficiency in design that has been approved for fabrication or construction where the 

design deviated from design criteria and bases 

• a deficiency in the fabrication or construction of, or significant damage to, an SSC that 

requires extensive evaluation, redesign, or repair to establish the adequacy of the SSC 

to perform its intended function 

• a deviation from performance specifications that will require extensive evaluation, 

redesign, or repair to establish the adequacy of the SSC to perform its intended function 

• a significant error in a computer program used to support activities affecting quality after 

it has been released for use 

• a deficiency, repetitive in nature, related to an activity or item subject to the IIFP’s QA 

Program 

• a condition that, if left uncorrected, has the potential to have a serious negative impact 

on activities or items subject to the IIFP’s QA Program controls. 

The NRC staff finds the corrective action program acceptable because it recognizes conditions 

adverse to quality and implements actions to prevent them from reoccurring. 

The affected management, including the COO or Plant Manager, when appropriate, will be 

notified of any trends or additional remedial action. 

12.1.15 Quality Assurance Records 

12.1.15.1 General 

The applicant will establish a records management system that will provide measures to control 

QA records. 

12.1.15.2 Responsibilities 

Records Management/Document Control will be responsible for the maintenance and 

implementation of the records control system. Managers will be responsible for (1) identifying 

QA records initiated by their organization, (2) controlling the records within their jurisdiction, and 

(3) transferring records to the Records Manager for retention. All activities related to records 

management and document control will be consistent with the requirements set forth in the QA 

Program. 

12.1.15.3 Implementation 

The applicant affirmed that through written procedures, QA records will be identified, prepared, 

indexed, stored, maintained, preserved, and kept safe in appropriate facilities. Records will be 

stored in authorized facilities that will provide appropriate measures to ensure continuous 

protection of the records. In addition, records will be legible, accurate, and complete and 

retrievable. The applicant will have appropriate measures in place to preclude entry of 

unauthorized personnel into record storage locations and to guard against larceny and 

vandalism. Records will be protected against damage, deterioration, and loss. 
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Documents will be considered valid records only if stamped, initialed, or signed and dated by 

authorized personnel or otherwise authenticated. The applicant will establish methods to 

ensure traceability of the record and the item(s) or activity(ies) to which it applies. The 

originating organization will approve and document any corrections to records. 

The applicant will classify QA records as “lifetime” or “nonpermanent.” Lifetime records must be 

maintained for the life of the particular item. Nonpermanent records will comply with the records 

retention requirements as defined in the IIFP records retention policy. As described by the 

applicant, lifetime records would be of significant value in (1) demonstrating that manufactured 

products meet requirements; (2) maintaining, reworking, repairing, replacing, or modifying 

critical items within the plant or manufactured products; and (3) determining the cause of an 

accident or malfunction of an item within the plant. Quality assurance Records have been 

determined by the NRC staff to be acceptable because the applicant has established 

procedures which keep them safe and ensure their quality is not compromised. 

12.1.16 Audits 

12.1.16.1 General 

The applicant discussed, in Section A.181 of the LA Appendix A (IIFP, 2012a), an audit system 

for activities and services within the scope of the IIFP’s QA Program. The audit system will 

establish planned and periodic audits to verify the compliance and the effectiveness of the QA 

Program in meeting system quality requirements. 

Audits will be executed in accordance with established procedures and will be performed by 

personnel having no direct responsibilities in the areas being audited and with sufficient 

authority and organizational freedom to make the audit process effective. 

As described by the applicant, internal audits will be performed in selected aspects of 

operational activities to ensure that audits of activities are completed within specified time 

periods. External audits to ensure compliance with applicable aspects of the IIFP’s QA Program 

and procurement requirements will be performed for selected suppliers and service contractors 

to verify and evaluate their QA programs, procedures, and activities. 

12.1.16.2 Responsibilities

The QA Manager will be responsible for the development, maintenance, scheduling, and 

performance of the internal audit and external supplier audit system. Audited organizations will 

be responsible for assisting during the planning and performance of audits (e.g., providing 

access to facilities, personnel, documents, and records) and ensuring that requests for 

corrective action are promptly answered and that actions taken to correct any discrepancy are 

adequate and timely. 
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12.1.16.3 Implementation

12.1.16.3.1 Training and Qualification 

The applicant stated that audit personnel will be properly trained to perform the required audits. 

Any technical specialists participating as audit team members will have received the required 

indoctrination and guidance before the audit. 

12.1.16.3.2 Scheduling 

The applicant will schedule internal and external audits in a manner to provide coverage and 

coordination with ongoing QA Program activities at a frequency commensurate with the status 

and importance of the activities. The audit schedules will be reviewed periodically and revised 

as necessary to ensure that coverage is maintained current. To provide adequate coverage, 

additional audits or surveillance of specific subjects will supplement regularly scheduled audits. 

Implementation audits may be scheduled and performed for initial evaluation of suppliers. 

External audits of approved noncommercial-grade suppliers will also be scheduled and 

performed based on the supplier certification program and established quality performance 

parameters. If the suppliers of services perform the work under the IIFP’s QA Program and 

procedures, external audits will not be required. External audits will be required for suppliers of 

services based on the supplier quality certification program, established quality performance 

parameters, and quality level of the service provided. 

12.1.16.3.3 Audit Plan 

Each audit will require an audit plan, developed and documented by the auditing organization. 

Each audit plan will identify the audit scope, requirements, audit personnel, activities to be 

audited, organizations to be notified, applicable documents and written procedures, schedule, 

and approved checklists of questions covering the items to be audited. 

12.1.16.3.4 Personnel and Selection of Audit Team 

The applicant will establish measures for the selection of the audit team and audit team 

familiarization before the start of each audit to ensure consideration of special abilities, 

specialized technical training, prior experience, personal characteristics, and education. The 

applicant described audit teams as containing one or more auditors, with an appointed team 

lead. The team lead will organize and direct the audit, coordinate the preparation and issuance 

of the audit report, and evaluate the responses. The applicant will establish expected audit 

deliverables before the start of the audit. 

12.1.16.3.5 Auditing 

The applicant included requirements for the organizations being audited to provide access and 

assistance to the audit personnel. The audit team will examine objective evidence to determine 

if the QA Program elements are being implemented effectively. Any conditions requiring prompt 

corrective action will be reported immediately to the management of the audited organization. 
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12.1.16.3.6 Reporting 

The audit team leader will sign the audit report, which will generally be issued within 30 days of 

the post audit conference. The audit report will be distributed to responsible management of 

both the auditing and the audited organizations. The applicant listed the following as 

information that will be contained in the report, as appropriate: 

• description of the audit scope 

• identification of the auditors 

• identification of persons contacted during audit activities 

• summary of audit results, including a statement of the implementation effectiveness of 

the QA Program elements that were audited 

• description of each reported adverse audit finding in sufficient detail to enable the 

audited organization to take corrective actions. 

12.1.16.3.7 Response and Follow-up Action 

The management of the audited organization will investigate adverse audit findings. It will also 

identify and schedule corrective actions and measures to prevent recurrence. In addition, 

management will notify the appropriate organization in writing of the actions taken or planned. 

Follow-up action will be taken to verify that corrective action is completed as scheduled. 

12.1.16.3.8 Records 

Audit records will include audit plans, audit reports, written replies, and documented completion 

of corrective actions. 

Audits of services within the QA program have been found acceptable by the NRC staff 

because audit plans will be made before each audit and audit personnel will have access to the 

required documentation to ensure the QA program is being implemented effectively. 

12.2 Evaluation of Findings

The NRC staff reviewed the QA elements that were presented in the form of a QA Program. 

The elements included were organization; QA Program; design control; procurement document 

control; instructions, procedures, and drawings; document control; control of purchased items 

and services; identification and control of materials, parts, and components; control of special 

processes; inspection; test control; control of M&TE; inspection, test, and operating status; 

control of nonconforming items; corrective action; QA records; and audits. Based on its review 

of the LA, the NRC staff has concluded that the applicant has adequately described the 

application of other QA elements (and the applicable QA elements of its principal contractors). 

The NRC staff also concludes the following: 

• The applicant has established and documented a commitment to an organization 

responsible for developing, implementing, and assessing the management measures for 
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providing reasonable assurance of safe facility operations in accordance with the criteria 

in Section 11.4 of NUREG-1520, Revision 0, issued May 2002 (NRC, 2002). 

• The applicant has established and documented a commitment to QA elements, and the 

administrative measures for staffing, performance, assessment of findings, and 

implementation of corrective actions are in place. 

• The applicant has developed a process for preparation and control of written 

administrative plant procedures, including procedures for evaluating changes to 

procedures, IROFS, and tests. The applicant will implement and maintain a process for 

review, approval, and documentation of procedures. 

• The applicant has established and documented surveillances, tests, and inspections to 

provide reasonable assurance of satisfactory in-service performance of IROFS. 

• Periodic independent audits will be conducted to determine the effectiveness of the 

management measures. Management measures will provide for documentation of audit 

findings and implementation of corrective actions. 

• Training requirements have been established and documented to provide employees 

with the skills to perform their jobs safely. Management measures have been provided 

for the evaluation of the effectiveness of training against predetermined objectives and 

criteria. 

• The organizations and persons performing QA element functions have the required 

independence and authority to effectively carry out their QA element functions without 

undue influence from those directly responsible for process operations. 

• QA elements cover the IROFS, as identified in the ISA Summary (IIFP, 2012b), and 

measures are established to prevent hazards from becoming pathways to higher risks 

and accidents. 

Accordingly, the NRC staff concludes that the applicant’s use of additional QA elements meets 

the requirements of 10 CFR Part 70 and provides reasonable assurance of the protection of 

public health and safety and of the environment. 

12.3 References

(IIFP, 2012a) International Isotopes Fluorine Products, Inc., “Fluorine Extraction Process and 

Depleted Uranium Deconversion Plant (FEP/DUP) License Application, Revision B,” May 2012, 

Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML12123A245. 

(IIFP, 2012b) International Isotopes Fluorine Products, Inc., “ISA Summary Rev. B for IIFP,” 

May 2012, Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML12123A245. 

(IIFP, 2012c) International Isotopes Fluorine Products, Inc., “Appendix A QAPD Rev B of IIFP 

License Application,” May 2012, Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML12123A245. 
12-27 

(NRC, 2002) U.S. Nuclear Regulatory Commission, “Standard Review Plan for the Review of a 

License Application for a Fuel Cycle Facility,” NUREG-1520, Rev. 0, March, 2002. 

 
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13-1 

13.0 PHYSICAL PROTECTION AND 

MATERIAL CONTROL AND ACCOUNTING 

13.1 Physical Protection

International Isotopes Fluorine Products, Inc. (IIFP), submitted, along with its license 

application (LA), Revision B of the Physical Security Plan (PSP), dated May 3, 2012 

(IIFP, 2012b), for review and approval by the U.S. Nuclear Regulatory Commission (NRC) in 

accordance with Title 10 of the Code of Federal Regulations (10 CFR) 40.3, “License 

Requirements.” This PSP provides for the protection of radiological material and chemicals of 

interest at IIFP. 

Although 10 CFR Part 40, “Domestic Licensing of Source Material,” does not include physical 

security requirements for licensees, the NRC has issued an order to another 10 CFR Part 40 

licensee with similar materials. IIFP opted to implement physical security in accordance with 

10 CFR 73.67, “Licensee Fixed Site and In-Transit Requirements for the Physical Protection of 

Special Nuclear Material of Moderate and Low Strategic Significance,” for a Category III fuel 

cycle facility. 

Details related to physical protection have been marked by the applicant as “Security-Related 

Information,” pursuant to Title 10 of the Code of Federal Regulations (CFR) 2.390, and are 

therefore not included in the public version of this report. 

13.1.1 Regulatory Guidance 

 The NRC staff’s review and analysis was conducted in accordance with the following regulatory 

guide: 

Regulatory Guide 5.59, Revision 1, “Standard Format and Content for a Licensee 

Physical Security Plan for the Protection of Special Nuclear Material of Moderate or Low 

Strategic Significance,” issued February 1983 (NRC, 1983). 

13.1.2 Staff Review and Analysis 

The IIFP submitted its PSP (IIFP, 2012b) as part of the LA for review and approval by the NRC 

in accordance with 10 CFR 40.3. After reviewing the initial PSP, the NRC submitted an RAI to 

IIFP regarding the security plan, which they responded to on January 10, 2011 (IIFP, 2011). 

The NRC reviewed the PSP and RAI responses to ensure that the applicant committed that it 

will store chemicals of interest only in a controlled access area, which will be monitored by an 

intrusion detection system or other approved procedures. The applicant also committed to 

having a guard force and offsite response force available to respond to unauthorized 

penetrations or activities. Furthermore, the applicant committed to establishing and maintaining 

response procedures for dealing with threats of theft or theft of material and reporting 

safeguards events as required by regulation. 

13.1.3 Evaluation of Findings for Physical Protection 

The NRC staff’s review of the applicant’s PSP for the protection of special nuclear materials of 

low-strategic significance contains information that the applicant has marked as “Official Use 

Only—Security-Related Information,” under 10 CFR 2.390, “Public Inspections, Exemptions, 
13-2 

Requests for Withholding.” The NRC staff reviewed the applicant’s PSP for fixed site physical 

protection of special nuclear materials of low-strategic significance and chemicals of concern. 

The NRC staff finds that the methods and procedures outlined in the PSP satisfy the 

performance objectives, systems capabilities, and reporting requirements specified in 

10 CFR 73.67. The PSP for the facility is acceptable and provides reasonable assurance that 

the requirements for the physical protection of special nuclear materials of low-strategic 

significance and chemicals of concern will be met. 

The NRC staff concludes that Revision B to the PSP (IIFP, 2012b), which incorporates the RAI 

responses (IIFP, 2011), is acceptable and recommends approval. 

13.2 Material Control and Accounting

The U.S. Nuclear Regulatory Commission (NRC) staff’s review of the applicant’s nuclear 

Material Control and Accounting (MC&A) Program is documented in this appendix to the Safety 

Evaluation Report. Appendix H contains information that has been marked as “Proprietary 

Information” and “Export Control Information” by the applicant, pursuant to Title 10 of the Code 

of Federal Regulations (10 CFR) 2.390, “Public Inspections, Exemptions, Requests for 

Withholding,” and 10 CFR 810, respectively. 

The NRC staff concludes that the applicant provided an acceptable Fundamental Nuclear 

Material Control Plan (FNMCP) for the proposed facility that will meet the applicable 

requirements of 10 CFR Part 74, “Material Control and Accounting of Special Nuclear Material.” 

The FNMCP describes acceptable methods for achieving the performance objectives in 

10 CFR 74.33(a) and the system capabilities of 10 CFR 74.33(c). As a result, the NRC staff 

finds that the applicant will meet the requirements in the area of MC&A to operate the proposed 

facility under 10 CFR Part 74. 

The purpose of this review was to verify that the applicant, IIFP, provided sufficient information 

to determine that the MC&A Program meets the applicable regulatory requirements in 

10 CFR Part 40, “Domestic Licensing of Source Material.” 

13.2.1 Regulatory Requirements for Material Control and Accounting 

The requirements for MC&A are specified in 10 CFR 40.64, “Reports.” 

13.2.2 Regulatory Acceptance Criteria 

The application will be considered acceptable if it meets the regulatory requirements specified in 

the following: 

• 10 CFR Part 40 

• NUREG/BR-0006, Rev. 7, “Instructions for Completing Nuclear Material Transaction 

Reports (DOE/NRC Forms 741 and 740M),” effective January 2009 (NRC, 2009a) 

• NUREG/BR-0007, Rev. 6, “Instructions for the Preparation and Distribution of Material 

Status Reports (DOE/NRC Forms 742 and 742C),” effective January 2009 

(NRC, 2009b). 
13-3 

13.2.3 Staff Review and Analysis 

The NRC staff reviewed and evaluated information provided by the applicant for the proposed 

MC&A Program in the LA (IIFP, 2012a). The following aspects of the MC&A Program were 

reviewed: 

• MC&A organization 

• measurements 

• physical inventories 

• recordkeeping and reports. 

13.2.3.1 Material Control and Accounting Organization 

The LA (IIFP, 2012a) describes a management structure and organization positions that have 

responsibilities related to the MC&A Program and that ensure clear responsibility for MC&A 

functions; the use of approved written MC&A procedures; and training for MC&A-related tasks. 

The NRC staff finds that the applicant is capable of implementing a management structure that 

permits effective functioning of the MC&A Program and ensures that the facility management 

structure will not adversely affect the MC&A Program performance. 

13.2.3.2 Measurements 

The LA (IIFP, 2012a) identified and described the measurement systems used for accounting 

purposes. The NRC staff reviewed the overall measurement program and finds the applicant’s 

system of measurements appropriate and acceptable to ensure that all quantities of nuclear 

material in the accounting records are based on reliable measurements. 

13.2.3.3 Physical Inventories 

The applicant’s physical inventory program contains the basic elements for scheduling, 

performing, and reporting annual physical inventories. The NRC staff finds that the physical 

inventory program demonstrates its ability to confirm the presence and quantities of nuclear 

materials. 

13.2.3.4 Recordkeeping and Reports 

The LA (IIFP, 2012a) describes the nuclear material accounting system for the facility and the 

control of this system. In addition, the applicant provided common reporting forms and formats 

to provide nuclear material data in accordance with current regulations. The NRC staff finds 

that the accounting system is secure and adequate. The NRC staff determined that the 

applicant’s establishment, maintenance, and protection of a recordkeeping system are 

acceptable and meet the reporting requirements. 

13.2.4 Evaluation Findings for Material Control and Accounting 

The NRC staff concludes that the applicant provided an acceptable MC&A Program for the 

proposed facility that will meet the applicable 10 CFR Part 40 requirements. The application 
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described acceptable methods for achieving the requirements of 10 CFR 40.64. As a result, the 

NRC staff finds that the applicant meets the requirements in the area of MC&A to operate the 

proposed facility under 10 CFR Part 40. 

13.3 References 

(IIFP, 2012a) International Isotopes Fluorine Products, Inc., “Fluorine Extraction Process and 

Depleted Uranium Deconversion Plant (FEP/DUP) License Application, Revision B,” May 2012, 

Agencywide Documents Access and Management System (ADAMS) Accession 

No. ML12123A245. 

(IIFP, 2012b) International Isotopes Fluorine Products, Inc., “Security Plan Rev B of IIFP 

License Application,” Rev. B, 2012, Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML12123A245. 

(IIFP, 2011) International Isotopes Fluorine Products, Inc., “Submittal of Responses to 

Requests for Additional Information re Application to License a Depleted Uranium Hexafluoride 

De-Conversion and Fluorine Extraction Process Facility,” January 10, 2011, Agencywide 

Documents Access and Management System (ADAMS) Accession No. ML110190058. 

(NRC, 2009a) U.S. Nuclear Regulatory Commission, NUREG/BR-0006, Rev. 7, “Instructions 

for Completing Nuclear Material Transaction Reports (DOE/NRC Forms 741 and 740M),” 

effective January 1, 2009. 

(NRC, 2009b) U.S. Nuclear Regulatory Commission, NUREG/BR-0007, Rev. 6, “Instructions 

for the Preparation and Distribution of Material Status Reports (DOE/NRC Forms 742 

and 742C),” effective January 1, 2009. 

(NRC, 1983) U.S. Nuclear Regulatory Commission, “Standard Format and Content for a 

Licensee Physical Security Plan for the Protection of Special Nuclear Material of Moderate or 

Low Strategic Significance,” Regulatory Guide 5.59, 1983. 
14-1 

14.0 Human Factors 

This appendix describes the human factors engineering (HFE) review of the International 

Isotopes Fluorine Products, Inc.’s (IIFP), license application (LA) (IIFP, 2012a) for a proposed 

depleted uranium deconversion facility. The purpose of this review as stated in NUREG-1520, 

Revision 1, “Standard Review Plan for the Review of a License Application for a Fuel Cycle 

Facility,” Appendix E, page 3-E-1, issued May 2010 (NRC, 2010), is to determine whether the 

applicant has established a HFE Implementation Plan (HFEIP)that will be applied to personnel 

activities identified as safety significant, consistent with the findings of the integrated safety 

analysis (ISA), and the determination of whether an item relied on for safety (IROFS) has 

special or unique safety significance. Specifically, this review addresses the standards, 

guidance, and practices that specify the design and implementation of the human system 

interfaces (HSIs) (e.g., alarms, displays, and controls) for IROFS that require operator actions at 

the proposed facility. 

14.1 Regulatory Requirements

Title 10 of the Code of Federal Regulations (10 CFR) 70.61(e) requires that IROFS remain 

available and reliable to perform their intended function when needed. NUREG-1520, 

Revision 1, Appendix E, page 3-E-1, states, “The application of HFE to personnel activities 

ensures that the potential for human error in the facility operations was addressed during the 

design of the facility by facilitating correct, and inhibiting wrong, decisions by personnel and by 

providing a means for detecting and correcting or compensating for error” (NRC, 2010). In 

addition, the baseline design criteria in 10 CFR 70.64(b)(2) require the facility to be designed to 

reduce challenges to IROFS. This requirement includes a significant human component for 

administrative IROFS. The applicant must demonstrate an adequate HFE program to ensure 

compliance with these requirements. 

14.2 Regulatory Guidance and Acceptance Criteria

Since IIFP submitted their application prior to completion of NUREG-1520, Revision 1 

(NRC, 2010), the bulk of the IIFP review was conducted consistent with NUREG-1520, 

Revision 0 (NRC, 2002). However, this version of the NUREG-does not contain guidance for 

the human factors review. Therefore, the NRC staff and IIFP agreed to conform to guidance in 

NUREG-1520, Revision 1 (NRC, 2010), for this portion of the review. 

In addition, the applicant has committed in the LA to designing its HSIs in accordance with the 

guidelines in NUREG-0700, Revision 2, “Human-System Interface Design Review Guidelines,” 

issued May 2002 (NRC 2002), and to performing its HFE program in accordance with the review 

guidance described in NUREG-0711, Revision 2, “Human Factors Engineering Program Review 

Model,” issued February 2004 (NRC, 2004). 

The key aspects of NUREG-0711 (NRC, 2004) applicable to fuel cycle facilities have been 

distilled into the nine acceptance criteria listed in Appendix E to NUREG-1520, Revision 1 

(NRC, 2010). The NRC staff considers these nine HFE program elements to be the HFE 

activities needed to provide reasonable assurance that the control and maintenance systems 

will be designed to facilitate correct decisions and inhibit wrong decisions by personnel and 

provide a means for detecting, correcting, and compensating for error. HFE is an ongoing 

process that begins with design and continues in conjunction with the corrective action process 

after the startup of a facility. 
14-2 

During the licensing review, portions of the HFE program are often under development and the 

details of design are not available for review; therefore, a programmatic review of the design 

process is performed. At the programmatic level, the applicant is still developing detailed 

procedures. Thus, the applicant has provided the NRC staff with the commitments to be 

performed to achieve the objectives of the HFE program elements. The NRC staff reviewed the 

objectives of each HFE element, the sources on which the method will be developed, and the 

types of results the applicant expects the method to produce. The implementation of this 

program through detailed procedures will be verified by inspection during the readiness review. 

The NRC staff used a graded approach, consistent with Footnote 2 in Appendix E to 

NUREG-1520, Revision 1 (NRC, 2010), to review the IIFP’s HFE program. The proposed 

facility is a depleted uranium deconversion facility. While there are human actions associated 

with the IROFS, the LA, Section 3.1.4.4, indicates a design preference to minimize the need for 

human actions through use of active and passive-engineered controls for IROFS (IIFP, 2012a). 

The facility does not have criticality hazards, large quantities of high-activity beta and gamma 

emitters, or transuranic nuclides. Therefore, the proposed facility has significantly lower risk 

and a less active safety role for the human staff than are typical of a commercial nuclear power 

plant. The NRC staff reviewed the application with the awareness of the detrimental effects that 

automation may have on human performance, situation awareness, and workload, which can 

cause increases in the probability of human error and incorrect decisionmaking when automated 

systems are not appropriately designed for the tasks of the human operator (O’Hara, 2010). 

The implementation of the applicant’s proposed HFEIP will be reviewed in the inspections 

required prior to operations to ensure that the implementation of the HFEIP meets the guidance 

provided in NUREG-1520, Revision 1, Appendix E to Chapter 3 (NRC, 2010). 

For this programmatic level review, the NRC staff used the criteria in NUREG-1520, Revision 1 

(NRC, 2010), to review the applicant’s HFEIP in terms of the overall objective of each HFE 

element, the technical basis (such as industry standards) that the applicant will use to develop 

the detailed HFE program, and the information that should be obtained from each HFE activity. 

14.3 Staff Review and Analysis

The NRC staff reviewed the LA (IIFP, 2012a); the ISA Summary (IIFP, 2012b); and the Quality 

Assurance Program Description (QAPD) (IIFP, 2012c). IIFP provided formal responses to staff 

requests for additional information (RAIs) by letter dated August 22, 2011 (IIFP, 2011b); this 

letter contains the HFE IP, to be added as Section 3.1.4 to the IIFP LA, Chapter 3 (IIFP, 2012a). 

This section presents the results of the review, which are organized by the acceptance criteria in 

Appendix E to Chapter 3 to NUREG-1520, Revision 1 (NRC, 2010), and the applicant’s HFEIP 

(IIFP, 2012a). 

14.3.1 Identification of Personnel Activities 

14.3.1.1 Evaluation 

The HFE program is described as an integral part of the ISA development, plant design, 

operation, and evaluation of operations. IROFS were determined, identified, and documented 

through the ISA process. The ISA process used a “What-if” process for the accident and 

consequence analysis process of the IIFP process hazard analysis (PHA) and relied on the 
14-3 

considerable engineering and operations experience of the ISA team. In a What-if process, the 

subject matter experts consider “worst-case” scenarios in the “What-if” PHA method. The HSI 

was considered as part of the accident analysis review and discussion process. 

14.3.1.2 Conclusion 

The scope of the HFEIP is identified as the administrative IROFS for which human actions are 

relied on to ensure the performance of these administrative controls. Section 3.2.5.8 of the LA 

(IIFP, 2012a) states that safety controls used at the IIFP facility can be characterized as either 

administrative or engineered. Administrative controls may rely on human actions, and IIFP 

stated that they are generally not considered to be as reliable as engineered controls since 

human errors usually occur more frequently than equipment failures. Engineered controls may 

be categorized as being “passive” or “active” and are not reliant on human action. Therefore, 

the NRC staff finds that the focus of the HFEIP on administrative controls is appropriate. 

Personnel actions will be identified and addressed during the operational experience review 

(OER); functional allocation analysis (FAA); task analysis (TA); HSI design, inventory, and 

characterization; procedures; training; and verification and validation (V&V) elements described 

in the HFE IP, Sections 3.1.4.3 through 3.1.4.9 of the LA (IIFP, 2012a). Section 3.1.4.1 of the 

LA indicates that this evaluation will also involve the identification of the structures, systems, 

and components (SSCs) and the boundary definitions of the active-engineered controls and 

passive-engineered controls IROFS. Table 6-3 of the IIFP ISA Summary (IIFP, 2012b) presents 

the IROFS that are subject to the IIFP’s HFEIP and the HFE program. This table includes all 

IROFS that involve administrative components or human actions. In Section 3.1.4.2 of the LA, 

the applicant committed to updating the list through the ISA Summary revision and amendment 

process. 

14.3.2 Human Factors Engineering Design Review Planning 

14.3.2.1 Goals and Scope 

Evaluation

The opening paragraph of the HFEIP in Section 3.1.4.2 of the HFEIP describes the applicant’s 

goals for the plan (IIFP, 2012a). The HFEIP seeks to address the potential for human error in 

the design, procedures, and training for the facility. The goal is to provide the means for 

detecting and correcting or compensating for human errors so that IROFS will remain available 

and reliable. The HFEIP contains the following objectives: 

• Critical personnel tasks are defined and accomplished within applicable time and 

performance criteria. 

• The anthropomorphic standards for the relevant population are defined and applied. 

• HSIs, procedures, staffing and qualifications, training, management, and organizational 

variables support a high degree of operating crew situational awareness. 

• Allocation of functions accommodates human capabilities and limitations. 

• Operator vigilance is maintained and distractions are minimized. 
14-4 

• The operator workload is acceptable. 

• Operator interfaces contribute to an error-free environment. 

• Error detection and recovery capabilities are provided. 

• Control areas minimize stressors and fatigue while ensuring adequate communication. 

This set of goals addresses HSI, procedures, ergonomics, time to respond/perform, vigilance, 

workload, error reduction and fatigue. Figure 3-1 of the LA provides an overview of the 

implementation of the HFEIP (IIFP, 2012a). Section 3.1.4.7 and other sections of the HFEIP 

state that the plan will be applied to procedures and training. 

The LA, Section 3.1.4 (IIFP, 2012a), states that the HFE Program will be applied to personnel 

activities involving administrative IROFS to ensure that the potential for human error is 

addressed during the design of the facility, operating procedure development, and personnel 

training. Tracking and maintenance activities will transition to quality assurance (QA) and 

configuration management functions as operation nears and begins. Therefore, HFE is 

included in the entire life cycle of the facility. 

Conclusions 

The applicant has established appropriate goals and scope to ensure that HFE is incorporated 

into IROFS. These objectives are appropriate for an HFE program because they cover the full 

scope of human performance and human activities within the system. The transition of the HFE 

goals into the QA Program provides the commitment that HFE will be included in the complete 

design cycle. The focus of the HFE program on IROFS that include human actions and 

personnel activities that support safety is of sufficient scope to provide reasonable assurance 

that critical human performance in the plant will be reviewed and that human factors will be 

included in the design to support these activities. Therefore, reasonable assurance is provided 

that the critical aspects of human performance in the operation of the facility will be assessed by 

the HFE review, and the goals and scope of the HFEIP during design, construction and 

operation of the plant is satisfactory. Thus, the goals and scope of the HFE design review 

planning are acceptable and meet the acceptance criterion as described in Appendix E, 

Section (B)(i), to NUREG-1520, Revision 1 (NRC, 2010). 

14.3.2.2 Composition and Experience of the Human Factors Engineering Team 

Evaluation

Section 3.1.4.2, of the LA (IIFP, 2012a) states that an HFE expert is to be added to the team. 

Table 5-1 in the ISA Summary shows the composition of the HFE/ISA team and indicates that 

the team will be multidisciplinary, trained, knowledgeable, and experienced in a wide array of 

ISA methods including hazards identification, PHA, and safety analysis and risk (IIFP, 2012b). 

The HFE expert will be relied on to achieve the stated HFE goals during design and 

construction of the facility. The HFE expert will provide oversight and develop advise on, the 

HFEIP for the IIFP facility before the design of the facility’s IROFS SSCs begins. The HFE 

expert will be responsible for overseeing the incorporation and implementation of the HFEIP 
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elements within the applicable HFE scope during the design, engineering, and startup of the 

facility. IIFP LA, Section 2.2.5.4, presents the minimum qualifications for the HFE professional 

position (IIFP, 2012a). The human factors engineer shall have as a minimum, a bachelor’s 

degree in HFE, or a bachelor’s degree in a scientific discipline or psychology with an emphasis 

on human factors, and at least 3 years of experience in human factors. As discussed in the ISA 

Summary the HFE/ISA team will have responsibility and authority for the following: 

• Identify the HSIs issues and evaluate actions and consequences involved. 

 

• Update the plan before the actual detailed design of IROFS. 

 

• Perform needed interviews to collect information on facilities handling, processing, or 

manufacturing uranium hexafluoride (UF6), uranium tetrafluoride (UF4), and uranium 

oxide and related radiological or chemical hazards 

• Coordinate and communicate with the design team, QA staff, and IIFP’s facility project 

team to develop and implement criteria, requirements, design review (style) guidance, 

and findings of analyses derived from the plan’s implementation processes. 

Conclusions

The experience requirements for the HFE expert, the composition of the HFE/ISA team, and the 

authority and responsibility of the HFE expert and the HFE/ISA team are appropriate to provide 

reasonable assurance that HFE will be considered in the design of HSI for personnel activities. 

Given the breadth of responsibilities of the HFE/ISA team, the organizational authority of the 

team, and the interconnections to the QA and configuration management plans, the HFEIP 

provides reasonable assurance that there will be proper development, execution, oversight, and 

documentation of the HFE function. 

14.3.2.3 Provide a Structured Approach to Human Factors Engineering Using Established 

Processes and Procedures 

Evaluation

Figure 3-1 of the LA (IIFP, 2012a) illustrates an overview of the HFE process to be undertaken 

by IIFP. The figure indicates that the elements identified by NUREG-1520, Revision 1, 

Appendix E, will support and be supported by other design efforts. 

The applicant stated that, for IROFS, an HFE review of the HSIs will be performed using 

NUREG-0700, Revision 2 (NRC, 2002), and NUREG-0711, Revision 2 (NRC, 2004), in 

accordance with the HFEIP (IIFP, 2012a). The results of the HFE review will be documented as 

part of the IROFS. 

Other standards and established processes referenced by the IIFP’s HFEIP to attain the goals 

and scope defined include the following: 

• Institute of Electrical and Electronics Engineers (IEEE) 1023, “Recommended Practice 

for the Application of Human Factors Engineering to Systems, Equipment, and Facilities 

of Nuclear Power Generating Stations and Other Nuclear Facilities” (IEEE, 2004) 
14-6 

• Electric Power Research Institute NP-3659, “Human Factors Guide for Nuclear Power 

Plant Control Room Development,” issued May 1991 (Kincaide, 1991) 

• Military Standard MIL-STD-1478, “Task Performance Analysis,” issued 1991 

(DOD, 1991) 

• MIL-HDBK-46855A, “Human Engineering Requirements for Military Systems, 

Equipment, and Facilities,” issued 2011 (DOD, 2011) 

• NUREG/CR-3331, “A Methodology for Allocation of Nuclear Power Plant Control 

Functions to Human and Automated Control,” issued 1983 (NRC, 1983). 

These standards provide a structured approach for incorporating HFE into the elements of 

design, including FAA and TA; HSI design, inventory, and characterization; staffing; procedure 

development; and training program development. 

Conclusion

The HFEIP provides commitments to implementing the elements defined by NUREG-1520, 

Revision 1, Appendix E (NRC, 2010), in accordance with NRC standards. The LA, 

Section 3.1.4 (IIFP, 2012a), states that the HFE program will be developed during the initial 

design phase and implemented throughout the design and operation of the facility. As the 

detailed design progresses, the elements of the HFE analysis will be reviewed and revised 

when process changes are made through the design control process. This indicates that the 

HFE Program will be an iterative process that is updated throughout the design of the facility, 

which is a critical aspect of a good HFE program. The application provides adequate assurance 

that the HFEIP will follow established the processes and standards listed above. It also 

provides a structured approach to development, design, and evaluation of personnel activities 

and is, therefore, acceptable. 

14.3.2.4 Conclusion to Human Factors Engineering Design Review Planning 

The HFEIP describes the activities to be undertaken including implementing the HSI to prevent 

consequences for IROFS, conducting OERs to identify necessary actions, and designating the 

NRC staff groups responsible for implementing HFE for IROFS. The LA (IIFP, 2012a) provides 

appropriate goals and scope to ensure that HFE practices and guidelines are implemented 

during design as overseen by the HFE expert. Therefore, there is appropriate team 

composition, experience, and organizational authority to ensure that HFE is considered in the 

design of the HSI for IROFS. 

14.3.3 Operating Experience 

The main purpose of conducting an OER is to identify HFE related safety issues that have 

occurred previously in other similar facilities. The OER also seeks to identify relevant HSI and 

equipment that should be considered in the development and design of the facility. The OER 

should provide information on the past performance of similar designs. The issues and lessons 

learned from the operating experience provide a basis for improving the plant design in a timely 

way. The objective of this element is to verify that the applicant has a program to identify and 

analyze HFE-related problems and issues in previous designs that are similar to the design 
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currently under review. In this way, negative features associated with predecessor designs may 

be avoided while positive features are retained. The applicant should do all of the following: 

• Review the HFE-related events or potential events for relevance. 

• Analyze the HSI technology employed for the relevant HFE events or potential events. 

 

• Conduct (or review existing) operator interviews and surveys on the HSI technology for 

the relevant HFE events or potential events. 

14.3.3.1 Evaluation 

Section 3.1.4.3 of the LA describes the applicant’s HFE OER (IIFP, 2012a). Figure 3-1 of the 

IIFP’s LA illustrates that OER is the initial step in the design process once the HFE/ISA team 

has been established. The figure illustrates that OER provides direct input to the identification 

of personnel activities, the FAA, the TA, and staffing and qualifications. The OER will be used 

to determine staffing requirements and categories of personnel and to provide input into HSI 

design and procedure development. 

The applicant committed to reviewing NRC and U.S. Department of Energy (DOE) event reports 

and the information available from the DOE’s Office of Scientific and Technical Information 

Bridge Web site. The review of previous events will include searches for personnel actions, as 

well as HSI technology. In addition, operator interviews will be conducted with personnel who 

have experience in uranium processing. Interviews will be sought with personnel with relevant 

experience (e.g., operating UF6 autoclaves, deconversion of UF6, transporting UF6 and 

radiological wastes, packaging fluorine products, refrigeration systems, and handling and 

transporting fluoride products). Potential sources identified include operating fuel cycle facilities, 

such as the DOE depleted uranium deconversion facilities at Portsmouth, OH; and Paducah, 

KY; the AREVA Eagle Rock Enrichment Facility and URENCO U.S.A. centrifuge enrichment 

facilities, and former plant personnel from the closed Sequoyah Fuels Corporation Depleted UF4

Deconversion Facility. The applicant has hired several engineers from existing plants to import 

their experience into the existing design (see the LA, Sections 2.1.2 and 3.2.7 [IIFP, 2012a]). 

14.3.3.2 Conclusion 

The applicant’s OER will include searches for events related to human error and human action 

in previous events at relevant facilities via accepted databases. It also includes interviews of 

individuals with industry experience and searches for information on HSI and technology. 

Information derived from the OER serves as the basis for the HFE Program and will be 

incorporated into the development of procedures, training, and HSI. Therefore, the operating 

experience meets the acceptance criteria in Appendix E, Section C, of NUREG-1520, 

Revision 1 (NRC, 2010). The NRC staff concludes that the HFE operating experience is 

satisfactory and meets the requirements in 10 CFR 70.61(e) and 10 CFR 70.64(b)(2). 

14.3.4 Functional Allocation Analysis and Task Analysis 

Functional requirements analysis is the identification of functions that must be performed to 

satisfy safety objectives. FAA is the review of the requirements for plant control and the 

assignment of control functions to (1) personnel (e.g., manual control), (2) system elements 

(e.g., automatic control and passive, self-controlling phenomena), and (3) combinations of 

personnel and system elements. The purpose of the functional requirements analysis and 
14-8 

function allocation review is to verify that the plant’s safety functions have been defined and that 

the allocation of those functions to human and system resources takes advantage of human 

strengths and avoids human limitations. The FAA should be based on the OER. Function 

allocations should take advantage of human strengths and avoid allocating functions that would 

be negatively affected by human limitations. 

The functions allocated to personnel define their roles and responsibilities. Human actions 

(HAs) are performed to accomplish these functions. HAs can be further divided into tasks. A 

task is a group of related activities that have a common objective or goal. TA is the 

identification of requirements for accomplishing these tasks (i.e., for specifying the requirements 

for the displays, data processing, controls, and job support aids needed to accomplish tasks). 

TA should be iterative and include the scope, identification of critical tasks, detailed description 

of personnel demands, and the incorporation of job design issues. It should address each 

operating mode and be used to support function allocation. 

14.3.4.1 Functional Analysis 

Evaluation

In Section 3.1.4.4 of the HFEIP in the LA (IIFP, 2012a), the applicant discusses its FAA. The 

safety philosophy of the ISA team was to minimize human involvement in the activation and 

performance of IROFS as much as available technology would allow. The concept of 

preferential use of active- and passive-engineered controls for IROFS is consistent with NRC 

guidance. Where engineered-controlled IROFS and administrative action IROFS are used in 

the same accident sequence, the engineered-control IROFS take precedence over human 

actions. 

IIFP proposes to follow the guidance and practices for FAA contained in NUREG/CR-3331 

(NRC, 1983). NUREG/CR-3331 is one of the technical bases for the guidance criteria on 

function allocation and functional analysis provided in NUREG-0711, Revision 2 (NRC, 2004), 

and NUREG-1520, Revision 1, Appendix E (NRC, 2010). It provides best practices for team 

composition, methods for performing the function allocation and functional analysis, assessment 

of the quality of the analysis, and guidance on appropriate allocation of functions between 

human and system so that the strengths of the human are leveraged and human weaknesses 

are avoided. 

Conclusion

The IIFP’s HFEIP indicates that the OER, ISA, and actual operating experience form the basis 

for determination of allocation of functions. The use of the processes, procedures, and 

guidance defined in NUREG/CR-3331 (NRC, 1983) provide reasonable assurance that 

personnel activities will be allocated to take advantage of human strengths and avoid demands 

that are not compatible with human capabilities. Therefore, the NRC staff concludes the 

functional analysis meets the acceptance criteria in Appendix E, Section d(i), of NUREG-1520, 

Revision 1 (NRC, 2010). 
14-9 

14.3.4.2 Task Analysis 

In the LA, Section 3.1.4.4, of the HFEIP (IIFP, 2012a), the applicant discussed TAs and stated 

that the TA involves determining the requirements of personnel to successfully and safely 

perform complex real-time control actions as part of the job assignments resulting from function 

allocation. 

The scope of the TA is defined as those IROFS that involve human actions. The IROFS that 

involve administrative components are defined in ISA Summary Table 6.3 (IIFP, 2012b). 

Section 3.2.5.8 of the LA (IIFP, 2012a) states that safety controls used at the IIFP facility can be 

characterized as either administrative or engineered. Administrative controls may rely on 

human actions, and IIFP stated that they are generally not considered to be as reliable as 

engineered controls, since human errors usually occur more frequently than equipment failures. 

Engineered controls may be categorized as being “passive” or “active” and are not reliant on 

human action. Therefore, the scope of the TA is appropriately defined as IROFS that involve 

administrative components. TA is defined as an ongoing, and thus iterative, activity that 

extends through the training process. The TA scope will include startup, shutdown, and 

emergency operations. TA is an input to function allocation and HSI design. Section 3.1.4.4 of 

the HFEIP indicates the applicant’s commitment to considering job design issues (IIFP, 2012a). 

IIFP proposes to follow the methods described in military standards MIL-STD-1478 (DOD, 1991) 

and MIL-HDBK-46855A (DOD, 2011). MIL-STD-1478 provides a process for incorporating FAA 

and TA into the design, which allows for flexibility in the performance of TA. Users are directed 

to select the established method of TA that is most cost effective if other guidance does not 

specify a method. The format and content of a task performance analysis product are 

described, inputs from TA to other design aspects such as HSI design and function allocation 

are indicated, and inputs from TA to procedure development are described. MIL-STD-1478 

indicates that for each task in the TA, the minimum data collected and analyzed should be 

equipment acted on, consequence of the action, and feedback information resulting from the 

action, which would support HSI design and functional allocation. MIL-HDBK-46855A contains 

recommended HFE program tasks and procedures and specifies the need for TA, the inputs 

from other processes to TA, and the outputs from TA to other elements. Section 8.3 of 

MIL-HDBK-48655A summarizes a number of widely accepted TA methods and the stage of 

design to which they apply. The TA meets the acceptance criteria in Appendix E, Section D(ii), 

to NUREG-1520, Revision 1 (NRC, 2010). 

14.3.4.3 Functional Allocation Analysis and Task Analysis Conclusion 

Based on the NRC staff’s review of the applicant’s FAA and TA, using the criteria described in 

Appendix E, Section D, to NUREG-1520, Revision 1, the NRC staff concludes that the FAA and 

TA are adequate and meet the requirements in 10 CFR 70.61(e) and 10 CFR 70.64(b)(2). 

14.3.5 Human System Interface Design, Inventory, and Characterization 

HSI design is the process by which HSI design requirements are developed and HSI designs 

are identified and refined. HSI should incorporate FAA and TA into the safety-significant HSI 

(alarms, displays, controls, and operator aids) via systematic application of HFE. The work 

environment, workspace layout, control panel and console design, control and display device 

layout, and information and control interface design details should be included. HFE should be 

applied to HSIs required to perform personnel activities. The HSI design process should 
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exclude the development of extraneous controls and displays. HSI documentation should 

include a complete HSI inventory and the basis for the HSI characterization. 

14.3.5.1 Evaluation 

The LA, Section 3.1.4 (IIFP, 2012a), contains the applicant’s proposal to follow the guidance 

provided in NUREG-0700, Revision 2 (NRC, 2002), which provides the guidelines for the 

development of HSI. The guidelines address the physical and functional characteristics of HSIs. 

Part I contains guidelines for the basic HSI elements: displays, user interface interaction and 

management, and controls. Part II contains the guidelines for the alarm system, group view 

display system, soft control system, computer-based procedure system, computerized operator 

support system, and communication system. Part III provides guidelines for workstations and 

workplaces. Part IV provides guidelines for HSI support (i.e., maintainability of digital systems). 

In the LA, Section 3.1.4.5, the applicant proposed to follow the design process methods in 

IEEE 1023, (IEEE, 2004). IEEE 1023 provides industry-accepted guidance and methods for the 

development of HSI, including considerations of the work environment and testing and 

validation of the designs. 

In the LA, Section 3.1.4.5, of the HFEIP (IIFP, 2012a), the applicant discussed HSI design. In 

this section, the applicant stated that the minimum inventory of HSIs, displays, alarms, and 

control instruments will be developed from the IROFS control function, the IROFS functional 

information, and input derived from the OER, FAA, and TA results, which would provide a basis 

for the characterization of the HSI. This minimum inventory will be updated during the design 

stage and during facility maintenance. It will be documented in accordance with the QA 

Program. The applicant committed to considering the work environment. The applicant also 

made a commitment that the implementation process for each HFE element will be applied 

accordingly to ensure consideration of the HFE aspects for the IROFS SSCs once those SSCs 

are defined and before actual design and design review. 

The IIFP’s QAPD will document the methodology for creation of the inventory of HSIs that are 

implemented to support IROFS and its results (IIFP, 2012c). This formal methodology is part of 

the facility’s Configuration Management Program and will ensure that the inventory list of HSIs 

is kept up to date. Extraneous displays and controls will be identified during task support 

verification. 

IIFP’s LA, Section 11.1.5.3 (IIFP, 2012a), states that the facility modification procedure will 

include HFE as a review and evaluation activity for any modifications that may impact HSIs. 

The applicant stated that if the assessment reveals that the modification affects HSI, the HFE 

process will be applied. The IIFP’s HFEIP states that if any HSIs are identified that have no 

associated tasks because the FAA or TA was incomplete, then the FAA or TA will be reviewed 

and shortcomings will be resolved. 

14.3.5.2 Conclusion 

The applicant committed to using the OER, FAA, and TA to establish the types of HSI to 

incorporate into the IROFS design. The HSI will be applied, as part of the HFE Program, to 

each control process involving the administrative components identified in the IROFS and their 

work environments. The inventory of HSIs will be tracked through the QA Program. The HSIs 

will be inventoried as part of the QA process. Based on the review of the applicant’s HSI 

design, inventory, and characterization using the acceptance criteria in Appendix E, Section E, 

to NUREG-1520, Revision 1 (NRC, 2010);, and to follow the structured approach provided in 
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IEEE 1023 (IEEE, 2004) and the development guidance provided in NUREG-0700, Revision 2 

(NRC, 2002), the NRC staff concludes that the HSI design, inventory, and characterization are 

adequate and meet the requirements in 10 CFR 70.61(e) and 10 CFR 70.64(b)(2). 

14.3.6 Staffing 

Staffing and qualification analyses identify the requirements for the number and qualifications of 

personnel. These analyses should reflect a thorough understanding of task and regulatory 

requirements. The applicant should conduct this review in a systematic manner that 

incorporates FAA and TA results. 

14.3.6.1 Evaluation 

Multiple sections of the LA and HFEIP address staffing. Sections 4.1.1 and 4.3 of the LA 

(IIFP, 2012a) state that staffing will be consistent with guidance in Regulatory Guide 8.2, “Guide 

for Administrative Practice in Radiation Monitoring,” issued February 1973 (NRC, 1973); and 

Regulatory Guide 8.10, Revision 1-R, “Operating Philosophy for Maintaining Occupation 

Radiation Exposures as Low as Is Reasonably Achievable,” issued May 1977 (NRC, 1977). 

Chapter 2 of the IIFP’s LA describes the qualifications for staff positions (IIFP, 2012a). The LA, 

Section 2.3.3 (IIFP, 2012a), provides information on the training and qualifications of staff. 

In the LA, Section 3.1.4.6 (IIFP, 2012a), the HFEIP discusses staffing and qualification. The 

HFEIP states that the applicant will ensure adequate staffing to operate the facility safely; 

furthermore, the OER, FAA, and TA will provide the initial basis for determining staffing 

requirements. These requirements will be considered throughout the design process. The 

initial estimates of staffing requirements and acceptability of staffing goals will be reassessed as 

the detailed design progresses. Therefore, assessment of staffing and use of the TA will be 

iterative. Categories of personnel activities will be derived from the TA, FAA, and OER. HSI 

design will be considered in the determination of staffing requirements and goals. During the 

design and build phase of operations, contractors will have the primary responsibility for 

ensuring adequate staffing, with oversight from an HFE expert or staff. As part of the HFE 

team, the HFE expert will review and provide oversight and advice on staffing needs. Staffing 

assignments, personnel tasks, and ergonomics will be validated during the V&V portion of the 

HFE IP. 

14.3.6.2 Conclusion 

The staffing and qualification levels are based on issues derived from the OER, the FAA, and 

the TA. Procedure development, HSI design, and V&V will be used to inform staffing levels, 

along with appropriate regulatory requirements. The number and qualifications of staff will be 

reviewed and updated throughout the design and build phase of operations, with oversight and 

review from the HFE team. Categories of tasks will be derived from the FAA and TA. 

Therefore, the NRC staff concludes the staffing plan meets the acceptance criteria in 

Appendix E, Section F, to NUREG-1520, Revision 1 (NRC, 2010). Thus, based on the review of 

the applicant’s staffing using the criteria described in Appendix E, Section F, of NUREG-1520, 

Revision 1, the staffing commitments are adequate and meet the requirements in 

10 CFR 70.61(e) and 10 CFR 70.64(b)(2). 
14-12 

14.3.7 Procedure Development 

Procedure development addresses the verification that human engineering principles and 

guidance are applied, along with all other design requirements, to develop procedures that are 

technically accurate, comprehensive, explicit, easy to use, and validated. Procedures should be 

derived from the same design process and analyses as the HSI and be subject to the same 

evaluation processes. 

14.3.7.1 Evaluation 

Chapter 11 of the LA covers nine elements for procedure development: identification, 

development, verification, review and comment resolution, approval, validation, issuance, 

change control, and periodic review (IIFP, 2012a). Section 11.1.5 of the LA provides a formal 

change control process for modifying procedures. The LA, Section 3.1.4.7, of the HFEIP 

(IIFP, 2012a), states that procedures will govern the production work aside from routine 

custodial work and office duties. ISA results, including IROFS, will be used to identify needed 

procedures. The adequacy and fidelity of procedures, and the allocation of functions in 

procedures, will be validated during the HFE integrated system validation as described in the 

LA, Section 3.1.4.9 (IIFP, 2012a). One objective of the HSI task support verification is to 

confirm that all procedures required for personnel tasks are provided. 

In a letter dated August 22, 2011 (IIFP, 2011b), IIFP stated that procedure development and 

refinement are ongoing activities. As construction nears completion, technical subject matter 

experts will perform walkthroughs to validate and refine procedures before use. Flowcharts of 

procedure steps will be created as part of V&V, as described in the LA, Section 3.1.4.9 

(IIFP, 2012a). Procedures will include required actions for normal and off-normal conditions of 

production. The QA elements and Configuration Management Program at the facility provide 

reasonable assurance that current procedures are available and used at all work locations. 

The applicant committed to considering the HFE program elements identified in NUREG-1520, 

Revision 1, Appendix E (NRC, 2010), in the development of procedures. IROFS and other 

safety-related items will be highlighted in work procedures, typically as “cautions” and 

“warnings.” The responsible organizations will develop and approve procedures. Employees 

will be trained on all procedures they follow as part of their work assignments. Work procedures 

and supplemental safety-related procedures are expected to be located in the general work 

areas. Facility and process changes will require procedure updates in the form of revisions, and 

such revisions shall be in place before the operation can recommence. The relevant 

procedures will incorporate changes to safety systems and safety basis documentation. 

Refinements and changes to procedures that are within the scope of the HFE Program will 

receive review for consideration of human factors. Employees will be retrained on the revised 

procedures before the restart of work. 

Procedures throughout the facility will identify IROFS and associated management measures. 

Procedures go through a validation process as part of the QA and Configuration Management 

Programs. The functional managers must also approve procedures. The LA, Section 3.1.4.2 

(IIFP, 2012a), states that the IIFP staff will implement procedure development and the HFE 

team will provide review, advice, and oversight. 
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14.3.7.2 Procedure Development Conclusion 

Each technical review area described in the application contains commitments to developing 

detailed procedures to ensure safe operations; the HFEIP contains commitments to developing 

procedures with consideration of the elements defined in NUREG-1520, Revision 1, Appendix E 

(NRC, 2010). The procedures will be developed and maintained consistent with the change 

control process and configuration management defined in Chapter 11 of the LA (IIFP, 2012a). 

In addition the HFE team will oversee and review the procedures to ensure that they incorporate 

HFE principles. The procedures are also reviewed by the V&V Program and evaluated during 

development via walkthrough. Therefore, based on the review of the applicant’s procedure 

development using the criteria described in Appendix E, Section G, to NUREG-1520, 

Revision 1, the NRC staff concludes that the procedure development commitments are 

adequate and meet the requirements in 10 CFR 70.61(e) and 10 CFR 70.64(b)(2). 

14.3.8 Training Program Development 

The applicant’s training program is designed to help ensure that personnel have the knowledge 

and skills necessary to perform their tasks. The development of the training program indicates 

how knowledge and skill will be evaluated, how training development will be coordinated with 

the HFE design process, and how the training program will be implemented in a manner 

consistent with human factors principles and practices. The training program development 

addresses the areas of review in Chapter 11 of NUREG-1520, Revision 1 (NRC, 2010), and 

results in a training program that provides personnel with qualifications commensurate with their 

activities. 

14.3.8.1 Evaluation 

In the LA, Section 3.1.4.8, of the HFEIP (IIFP, 2012a), the applicant discussed training program 

development. The objective of the training program is to ensure the job proficiency of facility 

personnel through effective training and qualification. The training development will address all 

personnel activities in the performance of job tasks involving IROFS. The training program will 

be coordinated with other activities of the HFE design process. 

The training combines both classroom and on-the-job training and involves formal evaluation to 

ensure that individuals demonstrate a minimum level of proficiency. The training uses a graded 

approach to emphasize safety-significant procedures, such as those that involve IROFS. 

Chapter 11 of the LA and each technical area throughout the application (IIFP, 2012a) describe 

additional commitments to the training program. These descriptions demonstrate that the 

training program is comprehensive and ensures that staff members have the proper skills to 

meet their responsibilities. 

The HFE is incorporated into the training program by regular job TAs to ensure that the training 

imparts the knowledge required to do the tasks and to iteratively improve the courses. In an 

RAI response (IIFP, 2011a), IIFP indicated that the methods used to perform job TA are 

contained in the same response as the description of the TA. LA Figure 3-1 (IIFP, 2012a) 

indicates that operation TA includes HFE TA as a subpart and that TA, functional allocation, 

staffing and qualifications, and OER form the basis for HSI development and procedure 

development and thus for HFE V&V and training program development. 

The HFEIP makes a commitment that the HFE team will review and provide oversight and 

advice to the training program as the facility nears operating status. Thus, the HFE team will 
14-14 

review the training program to ensure that training on IROFS incorporates human factors 

principles. 

The training also ensures that the specific tasks and details associated with each IROFS are 

addressed. The training program covers all appropriate staff for the facility and must be 

approved by the cognizant management. The training is reviewed regularly and incorporates 

input from the job task analyses. The development of training and regular review receives input 

from the HFE expert. 

14.3.8.2 Conclusion 

Each technical review area described in the application contains commitments to appropriate 

training. In addition, Chapter 11 of the LA states commitments to a comprehensive training 

program, which emphasizes safety-significant procedures. The training incorporates HFE 

practices and principles and is verified by the HFE expert or staff. Specific training is provided 

for actions related to IROFS. Therefore, based on the review of the applicant’s training program 

development and using the criteria described in Appendix E, Section H, to NUREG-1520, 

Revision 1 (NRC, 2010), the NRC staff concludes that the training program development 

commitments are adequate and meet the requirements in 10 CFR 70.61(e) and 

10 CFR 70.64(b)(2). 

14.3.9 Human Factors Verification and Validation 

V&V confirms that the design incorporates HFE into HSI in a manner that enables the 

successful completion of personnel activities. The V&V should be applied to personnel activities 

and HSI design as described in the V&V section of Chapter 3, Appendix E to NUREG-1520, 

Revision 1 (NRC, 2010). 

14.3.9.1 Evaluation 

Section 3.1.4.9 of the HFEIP (IIFP, 2012a) indicates that the purpose of the V&V Program will 

be to verify that the HSIs have the appropriate controls, procedures, and alarms so that the 

IROFS functions and controls remain available and reliable. The V&V Program will be used 

during the design phase of operations and to review individual actions needed to support 

IROFS. 

HSI Task Support Verification. The HSI task support verification (TSV) portion of the LA, 

Section 3.1.4.9 (IIFP, 2012a), provides a high-level overview of the process to be followed to 

perform TSV. This portion states that the criteria for TSV come from TAs of HSI requirements. 

HSIs and their characteristics will be compared to the personnel task requirements derived from 

the TA. Human engineering discrepancies (HEDs) are identified when needed HSIs are not 

available, characteristics do not match requirements, or HSIs are unnecessary or have no 

associated task. On this basis, the TSV demonstrates regulatory compliance with the 

acceptance criteria in Appendix E, Section (I)(i), to NUREG-1520, Revision 1 (NRC, 2010). 

HFE Design Verification. HFE design verification will be performed to determine that each HSI 

identified for personnel activity in the IROFS has HFE incorporated into its design and that the 

HSI has been verified to conform with HFE criteria and style guidance. Section 3.1.4.5 states 

that design review criteria and style guidance, with consideration of human factors for affected 

layouts, locations, and configurations of alarms, displays, and control instrumentation will be 

developed for use by the design engineers before they begin the actual design. The HFE issue 
14-15 

resolution verification portion of the HFEIP states that the HFEIP will be subject to the 

requirements of the QAPD. All HFE requirements and design documents will be controlled 

under the design control provisions of the Configuration Management Program, subject to the 

same change control as analysis, specifications, and drawings. The LA, Section A.3.3.4, of the 

QAPD (IIFP, 2012c), states that testing used for design verification shall demonstrate adequacy 

of performance under conditions that simulate the most adverse design conditions. The tests 

used for design verification must meet all the design requirements. 

Integrated System Validation (ISV). The applicant stated in LA, Section 3.1.4.9, that the 

purpose of the ISV is as follows: 

• validate the role of plant personnel; 

• validate staffing assignments; 

• validate each human ergonomic function; 

• validate specific personnel tasks; 

• validate that the integrated system performance is tolerant of failures; and 

• validate procedure adequacy, allocation, and fidelity. 

IIFP committed to having ISV done by knowledgeable personnel performing task walkthroughs 

and talk throughs in a part-task simulator, which provides a representation of the process 

control rooms and local control panels. The part-task simulator will have sufficient fidelity to be 

used as a training tool and to upgrade software or design new software. Walkthroughs and talk 

throughs are widely accepted HFE techniques for validating processes. When combined with 

simulation, they provide an effective method of assessing the integration of all aspects of design 

in a realistic task setting, which provides reasonable assurance that HFE issues will be 

identified. Reviews of TA and functional analysis findings, flowcharts of procedure steps, and 

operational task charts will also be used in combination with the part-task simulation 

walkthroughs and talk throughs to provide reasonable assurance of a link to the HSI and task 

requirements developed earlier. 

HFE Issue Resolution Verification. The QAPD governs HFE issue resolution verification. As 

stated in the QAPD and in the HFE IP, qualified reviewers who were not involved in the design 

process will perform HFE issue resolution. All HFE requirements and design documents will be 

controlled under the design control provisions of the Configuration Management Program and 

will be subject to the same change control as analysis, specifications, and drawings. 

Results of V&V activities will be summarized in a report, and discrepancies will be identified as 

HED. An HED report will include the results of resolution. The QAPD contains the facility’s 

existing Corrective Action and QA Programs described in the LA, Chapter 11, Section 11.6 

(IIFP, 2012a), and in the QAPD (IIFP, 2012c), respectively. The V&V involves a check and 

review of the HFE program and HSIs. The report will identify the changes needed and describe 

how they are resolved. Identification of HFE issues will be incorporated into the Root-Cause 

Analysis Program, and issues will be resolved via the corrective action process. The 

commitment to include a root-cause analysis of HFE into the Corrective Action Program 
14-16 

provides reasonable assurance of regulatory compliance with the acceptance criteria in 

Appendix E, Section (i)(iv), to NUREG-1520, Revision 1 (NRC, 2010). 

14.3.9.2 Conclusion 

Based on the review of the applicant’s human factors V&V and using the criteria described in 

Appendix E, Section I, to NUREG-1520, Revision 1 (NRC, 2010), the NRC staff concludes that 

the human factors V&V commitments are adequate and meet the requirements in 

10 CFR 70.61(e) and 10 CFR 70.64(b)(2).

14.4 Evaluation Findings

The NRC staff reviewed the applicant’s LA (IIFP, 2012a); ISA Summary (IIFP, 2012b); the 

QAPD (IIFP, 2012c); and the Configuration Management Program description in the LA for the 

application of HFE to personnel activities for the proposed facility. The NRC staff conducted the 

review at a programmatic level, as described in Appendix E, Section E.2, above. 

The documents named above and the HFEIP (IIFP, 2012a), with reliance on the HFE standards 

identified in the HFEIP elements, provide acceptable objectives, expected results, and plans for 

documentation. The NRC staff concludes that the applicant has committed to incorporating 

accepted HFE guidance and practices to design and implement those HSIs that support IROFS 

requiring operator actions. The guidance and practices, when implemented as described in the 

LA (IIFP, 2012a), demonstrate compliance with the regulatory criteria in Appendix E to Chapter 

3 of NUREG-1520, Revision 1 (NRC, 2010), and should result in IROFS that will perform their 

intended functions and meet the requirements of 10 CFR Part 70, “Domestic Licensing of 

Special Nuclear Material.” 

The detailed procedures for implementation of the HFE program will be reviewed and verified 

during the inspection program. 

14.5 References

(DOD, 2011) U.S. Department of Defense, MIL-HDBK-46855A, “Department of Defense 

Standard Practice: Human Engineering Requirements for Military Systems, Equipment, and 

Facilities,” Washington, DC, May 2011. 

(DOD, 1991) U.S. Department of Defense, Military Standard MIL-STD-1478, “Task 

Performance Analysis,” Washington, DC, May 1991. 

(IEEE, 2004) Institute of Electrical and Electronics Engineers, IEEE 1023, “Recommended 

Practice for the Application of Human Factors Engineering to Systems, Equipment, and 

Facilities of Nuclear Power Generating Stations and Other Nuclear Facilities,” Piscataway, NJ, 

December 8, 2004. 

(IIFP, 2012a) International Isotopes Fluorine Products, Inc., “Fluorine Extraction Process and 

Depleted Uranium Deconversion Plant (FEP/DUP) License Application, Revision B,” May 2012, 

Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML12123A245. 
14-17 

(IIFP, 2012b) International Isotopes Fluorine Products, Inc., “ISA Summary Rev. B for IIFP,” 

May 2012, Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML12123A245. 

(IIFP, 2012c) International Isotopes Fluorine Products, Inc., “Appendix A QAPD Rev B of IIFP 

License Application,” May 2012, Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML12123A245. 

(IIFP, 2011a) International Isotopes Fluorine Products, Inc., “Official Responses to Human 

Factors RAIs,” Rev. A, May 2011. Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML11130A128. 

(IIFP, 2011b) International Isotopes Fluorine Products, Inc., “Official Responses to Human 

Factors RAIs,” Rev. B, August 2011. Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML11250A055. 

(Kincaide, 1991) Kincaide, R.G., and J. Anderson, Electric Power Research Institute, “Human 

Factors Guide for Nuclear Power Plant Control Room Development,” Electric Power Research 

Institute EPRI/NP-3659, MIL-STD-1478, “Military Standard: Task Performance Analysis,” 

May 1991. 

(NRC, 2010) U.S. Nuclear Regulatory Commission, NUREG-1520, “Standard Review Plan for 

the Review of a License Application for a Fuel Cycle Facility,” Rev. 1, Washington, DC, 

May 2010. 

(NRC, 2004) U.S. Nuclear Regulatory Commission, NUREG-0711, Rev. 2, “Human Factors 

Engineering Program Review Model,” Washington, DC, February 2004. 

(NRC, 2002) U.S. Nuclear Regulatory Commission, NUREG-0700, Rev. 2, “Human-System 

Interface Design Review Guidelines,” Washington, DC, May 2002. 

(NRC, 1983) U.S. Nuclear Regulatory Commission, NUREG/CR-3331, “A Methodology for 

Allocation of Nuclear Power Plant Control Functions to Human and Automated Control,” 

Washington, DC, 1983. 

(NRC, 1977) U.S. Nuclear Regulatory Commission, Regulatory Guide 8.10, Rev. 1-R, 

“Operating Philosophy for Maintaining Occupational Radiation Exposures as Low as Is 

Reasonably Achievable,” Washington, DC, May 1977. 

 (NRC, 1973) U.S. Nuclear Regulatory Commission, Regulatory Guide 8.2, “Guide for 

Administrative Practice in Radiation Monitoring,” Washington, DC, February 1973. 

(O’Hara, 2010) O’Hara, J., and J. Higgins, “Human-System Interfaces to Automatic Systems: 

Review Guidance and Technical Basis,” Brookhaven National Laboratory Technical Report 

BNL-91017-2010, Upton, NY, 2010. 

 
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15-1 

15.0 SAFETY EVALUATION REPORT PREPARERS 

The individuals and organizations listed below are the principal contributors to the preparation of 

this Safety Evaluation Report. The U.S. Nuclear Regulatory Commission’s (NRC’s) staff 

directed the effort and contributed to the technical evaluations. The NRC staff also used 

contractor support from the Center for Nuclear Waste Regulatory Analyses in the preparation of 

this document. 

Table 15-1 U.S. Nuclear Regulatory Commission Contributors 

Matthew Bartlett, Office of Nuclear Material Safety 

and Safeguards (NMSS) 

NRC Project Manager 

General Information and Organization and 

Administration 

Maria Guardiola, NMSS NRC Project Manager 

Patti Silva, NMSS Branch Chief, Conversion, Deconversion 

and Enrichment Branch (CDEB) 

Mary Adams, NMSS Decommissioning 

Damaris Arroyo, NMSS Management Measures and Quality 

Assurance 

Julie Marble, Office of Nuclear Regulatory 

Research (Research) Human Factors Engineering 

Gregory Chapman, NMSS Health Physics and Accident Analysis 

Jonathan De Jesus-Segarra, NMSS Chemical Safety 

James Downs, NMSS Fire Safety 

Stan Echols, NMSS Environmental Protection 

Madhumita Sircar, Research Seismic and Structural 

Roman Przygodzki, Office of Federal and State 

Materials and Environmental Management 

Programs (FSME) 

Financial Qualification and 

Decommissioning Financial Assurance 

Mike Norris, Office of Nuclear Security Incidence 

Response (NSIR) Emergency Management 

James Anderson, NSIR Emergency Management 

Tom Pham, NMSS Material Control and Accountability 

Steve Ward, NMSS Material Control and Accountability 

Yawar Faraz, NMSS Integrated Safety Analysis 

Mary Adams, NMSS Integrated Safety Analysis 

Barry Wray, NSIR Physical Protection 

 
15-2 

Table 15-1 Contributors to the Safety Evaluation Report (continued) 

Center For Nuclear Waste Regulatory Analyses Contributors 

Asadul Chowdhury Structural Design 

Sui-Min (Simon) Hsiung Natural Phenomena 

John Stamatakos Seismic Engineering 

George Adams Electrical Engineering and Instrumentation 

and Controls 

Stanley Consultants Contributors 

Chris Storms Stanley Consultants, Muscatine, Iowa 

 
A-1 

Appendix A ACCIDENT ANALYSIS 

The U.S. Nuclear Regulatory Commission's (NRC) staff independently evaluated the possible 

consequences of a subset of potential accident sequences identified in the applicant's 

Integrated Safety Analysis (ISA) Summary. The primary purpose of this analysis is to ensure 

that the modeling used by the applicant for assessing accidents is acceptable. This evaluation 

is documented in Appendix A to this Safety Evaluation Report. Appendix A contains information 

that has been marked by the applicant as "Security Related Information” and is designated by 

the NRC as “Official Use Only—Security Related Information," pursuant to Title of the Code of 

Federal Regulations (10 CFR) 2.390. 

The five accident sequences analyzed by the NRC staff are presented in Appendix A. The 

selected accident consequences vary in severity from high- to low-consequence events, and 

include accidents initiated by natural phenomena, operator error, and equipment failure. 

The NRC staff concluded that the proposed International Isotopes Fluorine Products, Inc.’s 

(IIFP), design would reduce the risk (likelihood) of these accidents by using items relied on for 

safety (IROFS) and defense-in-depth measures. In addition, the facility’s Emergency Plan (EP) 

addresses these types of events. The NRC staff independently verified the applicant's accident 

analysis by performing confirmatory calculations and modeling. The NRC staff concluded that 

through the combination of plant design, passive- and active-engineered IROFS, administrative 

IROFS, and defense-in-depth features, the consequences of potential accidents at the proposed 

IIFP will pose an acceptably low safety risk to workers, the public, and the environment. As a 

result, the NRC staff determined that the applicant meets the requirements to operate the 

proposed facility under 10 CFR Part 40, “Domestic Licensing of Source Material.” 

Details related to the accident analysis have been marked by the applicant as “Security-Related 

Information,” pursuant to Title 10 of the Code of Federal Regulations (CFR) 2.390, and are 

therefore not included in the public version of this report. 

 
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B-1 

Appendix B STRUCTURAL AND GEOTECHNICAL DESIGN 

The U.S. Nuclear Regulatory Commission (NRC) staff independently evaluated the protection of 

building structures against natural phenomena described in the applicant’s License Application 

(LA) (IIFP, 2012a); Integrated Safety Analysis (ISA) Summary (IIFP, 2012b); and IIFP 

responses to NRC requests for additional information (IIFP, 2011a, 2011b, and 2011c). This 

evaluation is documented in this appendix of the Safety Evaluation Report (SER). Appendix B 

contains information that has been marked as “Security Related Information” by the applicant 

and is designated by the NRC as “Official Use Only—Security Related Information.” 

The NRC staff reviewed the structural and geotechnical design of the proposed facility. This 

included reviewing the building designs, general design criteria, loading conditions, and 

applicable American Society of Civil Engineers (ASCE) standards. The proposed International 

Isotopes Fluorine Products, Inc.’s (IIFP), design reduces the risk (likelihood) of the accident by 

identifying items relied on for safety (IROFS), and defense-in-depth features. 

The NRC staff concluded that, through the combination of plant design, passive- and 

active-engineered IROFS, administrative IROFS, and defense-in-depth features, building 

structures will have adequate protection against natural phenomena so that the safety risk to 

workers, public, and the environment will be acceptably low. As a result, the NRC staff finds 

that the applicant meets the requirements to operate the proposed facility under Title 10 of the 

Code of Federal Regulations (10 CFR) Part 70, “Domestic Licensing of Special Nuclear 

Material.” 

Details related to the structural and geiotechnical design have been marked by the applicant as 

“Security-Related Information,” pursuant to Title 10 of the Code of Federal Regulations (CFR) 

2.390, and are therefore not included in the public version of this report. 

 
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C-1 

Appendix C INSTRUMENTATION AND CONTROL 

The U.S. Nuclear Regulatory Commission (NRC) staff's review of the instrumentation and 

control (I&C) is documented in Appendix C to this Safety Evaluation Report (SER). Appendix C 

contains information that has been marked by the applicant as "Security-Related Information” 

and is designated by the NRC as “Official Use Only—Security Related Information," pursuant to 

Title 10 of the Code of Federal Regulations (10 CFR) 2.390. 

The NRC staff reviewed accident sequences related to I&C. This included reviewing the 

baseline design criteria (BDC), general design criteria, applicable Institute of Electrical and 

Electronics Engineers (IEEE) standards, and applicable American National Standards Institute 

standards. The proposed International Isotopes Fluorine Products, Inc.’s (IIFP), design reduces 

the risk (likelihood) of the accident by identifying items relied on for safety (IROFS), and 

defense-in-depth features. The NRC staff concludes that through the combination of plant 

design, passive and active engineered IROFS, administrative IROFS, and defense-in-depth 

features, accidents related to the electrical system and I&C at the proposed IIFP will pose an 

acceptably low safety risk to workers, public, and the environment. As a result, the NRC staff 

determined that the applicant meets the requirements to operate the proposed facility under 

Title 10 of the Code of Federal Regulations (10 CFR) Part 70, “Domestic Licensing of Special 

Nuclear Material.” 

Details related to instrumentation and control have been marked by the applicant as “SecurityRelated Information,” pursuant to Title 10 of the Code of Federal Regulations (CFR) 2.390, and 

are therefore not included in the public version of this report. 

 
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D-1 

Appendix D ELECTRICAL SYSTEMS 

The U.S. Nuclear Regulatory Commission (NRC) staff's review of the electrical system is 

documented in Appendix D to this Safety Evaluation Report. Appendix D contains information 

that has been marked by the applicant as "Security-Related Information," pursuant to Title 10 of 

the Code of Federal Regulations (10 CFR) 2.390 and is designated by the NRC as “Official Use 

Only-Security Related Information.”. 

The NRC staff reviewed the electrical power system for the facility. The review verified that the 

design of the electrical system will meet the regulatory requirements specified in 

10 CFR Part 70, “Domestic Licensing of Special Nuclear Material,” Subpart H. The NRC staff 

evaluated the adequacy of the proposed conceptual design and intended operations of these 

systems as reflected in the applicant’s commitments and goals with respect to that design. The 

NRC staff evaluated the applicant’s commitments for completing the design of the electrical 

system in a manner that addresses specific regulatory acceptance criteria, identified in 

Section 3.2 of this report. These commitments and goals are described within the Integrated 

Safety Analysis (ISA) Summary (IIFP, 2012b) and the applicant’s responses to the NRC staff’s 

request for additional information (RAI) (IIFP, 2011a). 

Details related to electrical systems have been marked by the applicant as “Security-Related 

Information,” pursuant to Title 10 of the Code of Federal Regulations (CFR) 2.390, and are 

therefore not included in the public version of this report. 

 
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