NUREG-1827

Safety Evaluation Report

for the National Enrichment Facility 

in Lea County, New Mexico

\

Docket No. 70-3103

Louisiana Energy Services

 U.S. Nuclear Regulatory Commission

Office of Nuclear Material Safety and Safeguards

Washington, DC 20555-0001
 

 NUREG-1827 

 

Safety Evaluation Report

for the National Enrichment Facility

in Lea County, New Mexico

Docket No. 70-3103

Louisiana Energy Services

Manuscript Completed: June 2005

Date Published: June 2005

Division of Fuel Cycle Safety and Safeguards

Office of Nuclear Material Safety and Safeguards

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001
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iii

ABSTRACT

The report documents the U.S. Nuclear Regulatory Commission (NRC) staff review and safety

and safeguards evaluation of the Louisiana Energy Services’ (LES) (the applicant) application

for a license to construct a gas centrifuge uranium enrichment facility and possess and use

special nuclear material (SNM), source material, and byproduct material in a gas centrifuge

uranium enrichment facility. LES proposes that the gas centrifuge uranium enrichment facility

be located in Lea County, New Mexico, near the city of Eunice, New Mexico. The facility will

possess natural, depleted, and enriched uranium, and will enrich uranium up to a maximum of 5

weight percent uranium-235.

The objective of this review is to evaluate the potential adverse impacts of operation of the

facility on worker and public health and safety under both normal operating and accident

conditions. The review also considers physical protection of SNM and classified matter,

material control and accounting of SNM, and 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 and safeguards of facility

operations and that operation of the facility does not pose an undue risk to worker and public

health and safety.
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TABLE OF CONTENTS

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

TABLE OF CONTENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v

EXECUTIVE SUMMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xv

LIST OF ACRONYMS AND ABBREVIATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xix

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 ACCEPTANCE CRITERIA . . . . . . . . . . . . . . . . . . . . 1-1

1.1.3 STAFF REVIEW AND ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1

1.1.4 EVALUATION FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3

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

1.2.1 REGULATORY REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3

1.2.2 REGULATORY ACCEPTANCE CRITERIA . . . . . . . . . . . . . . . . . . . . 1-3

1.2.3 STAFF REVIEW AND ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3

1.2.3.1 Corporate Identity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3

1.2.3.2 Foreign Ownership, Control, or Influence . . . . . . . . . . . . . . 1-5

1.2.3.3 Financial Qualifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5

1.2.3.4 Type, Quantity, and Form of Licensed Material . . . . . . . . . 1-8

1.2.3.5 Authorized Uses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-8

1.2.3.6 Special Exemptions or Special Authorizations . . . . . . . . . . 1-9

1.2.3.7 Security of Classified Matter . . . . . . . . . . . . . . . . . . . . . . . 1-10

1.2.4 EVALUATION FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-11

1.3 SITE DESCRIPTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-11

1.3.1 REGULATORY REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . 1-12

1.3.2 REGULATORY ACCEPTANCE CRITERIA . . . . . . . . . . . . . . . . . . . 1-12

1.3.3 STAFF REVIEW AND ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-12

1.3.3.1 Site Geography . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-12

1.3.3.2 Demographics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-14

1.3.3.3 Meteorology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-16

1.3.3.4 Geology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-19

1.3.3.5 Hydrology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-23

1.3.4 EVALUATION FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-24

1.4 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-24

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

2.1 REGULATORY REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1

2.2 REGULATORY ACCEPTANCE CRITERIA . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1

2.3 STAFF REVIEW AND ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1

2.3.1 Organizational Responsibilities and Qualifications . . . . . . . . . . . . . . . 2-2

2.3.2 Management Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-8

2.3.2.1 Configuration Management . . . . . . . . . . . . . . . . . . . . . . . . . 2-8

2.3.2.2 Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-8

2.3.2.3 Training and Qualifications . . . . . . . . . . . . . . . . . . . . . . . . . 2-8

2.3.2.4 Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-9

2.3.2.5 Audits and Assessments . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-9

2.3.2.6 Safety Review Committee . . . . . . . . . . . . . . . . . . . . . . . . . . 2-9

2.3.2.7 Incident Investigations . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-10

2.3.2.8 Employee Concerns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-10

2.3.2.9 Records Management . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-10

2.3.2.10 Written Agreements with Offsite Emergency Agencies . . . 2-10

2.3.3 Transition from Design and Construction to Operations . . . . . . . . . . 2-10

2.4 EVALUATION FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-11

2.5 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-11

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

3.1 REGULATORY REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1

3.2 REGULATORY ACCEPTANCE CRITERIA . . . . . . . . . . . . . . . . . . . . . . . . . . 3-3

3.3 STAFF REVIEW AND ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-3

3.3.1 General Information . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-4

3.3.1.1 Site Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-5

3.3.1.2 Facility Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-25

3.3.1.3 Processes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-32

3.3.2 Baseline Design Criteria and Defense-In-Depth . . . . . . . . . . . . . . . . 3-38

3.3.3 Safety Program and Integrated Safety Analysis . . . . . . . . . . . . . . . . 3-40

3.3.3.1 Safety Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-40

3.3.3.2 Integrated Safety Analysis and ISA Summary . . . . . . . . . 3-42

3.3.3.3 Summary of ISA Results . . . . . . . . . . . . . . . . . . . . . . . . . . 3-52

3.4 EVALUATION FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-53

3.5 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-56

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

4.1 REGULATORY REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1

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

4.1.2 As Low As is Reasonably Achievable (ALARA) Program . . . . . . . . . . 4-1

4.1.3 Organization and Personnel Qualifications . . . . . . . . . . . . . . . . . . . . . 4-1

4.1.4 Written Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1

4.1.5 Training . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1

4.1.6 Ventilation and Respiratory Protection Programs . . . . . . . . . . . . . . . . 4-1

4.1.7 Radiation Survey and Monitoring Programs . . . . . . . . . . . . . . . . . . . . 4-2

4.1.8 Additional Program Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-2

4.2 REGULATORY ACCEPTANCE CRITERIA . . . . . . . . . . . . . . . . . . . . . . . . . . 4-2

4.3 STAFF REVIEW AND ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-2

4.3.1 Radiation Protection Program Implementation . . . . . . . . . . . . . . . . . . 4-2

4.3.2 ALARA Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-4

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

4.3.4 Written Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-5

4.3.5 Training . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-6

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

4.3.7 Radiation Survey and Monitoring Programs . . . . . . . . . . . . . . . . . . . . 4-8

4.3.8 Additional Program Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . 4-10

4.4 EVALUATION FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-11

4.5 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-11

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

5.1 REGULATORY REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1

5.2 REGULATORY ACCEPTANCE CRITERIA . . . . . . . . . . . . . . . . . . . . . . . . . . 5-2

5.3 STAFF REVIEW AND ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-2

5.3.1 Possession Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-2

5.3.2 NRC NCS Regulatory Guide 3.71 and ANSI/ANS Series 8 NCS

Standards . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-2

5.3.3 Management of the NCS Program . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-4

5.3.4 NCS Organization and Administration . . . . . . . . . . . . . . . . . . . . . . . . 5-5

5.3.4.1 NCS Organization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-5

5.3.4.2 NCS Administration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-5

5.3.4.3 NCS Organization and Administration Conclusions . . . . . 5-10

5.3.5 NCS Management Measures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-10

5.3.6 NCS Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-13

5.3.6.1 NCS Methodologies and Technical Practices . . . . . . . . . . 5-13
viii

5.3.6.2 NCS Criticality Accident Alarm System . . . . . . . . . . . . . . . 5-26

5.3.6.3 NCS Subcriticality of Operations and Margin of 

Subcriticality for Safety . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-28

5.3.6.4 NCS Baseline Design Criteria . . . . . . . . . . . . . . . . . . . . . . 5-30

5.3.6.5 NCS in the Integrated Safety Analysis (ISA) Program 

and the ISA Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-30

5.3.6.6 Additional NCS Program Commitments . . . . . . . . . . . . . . 5-35

5.4 EVALUATION FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-36

5.5 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-38

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

6.1 REGULATORY REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1

6.2 REGULATORY ACCEPTANCE CRITERIA . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1

6.3 STAFF REVIEW AND ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1

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

6.3.1.1 Gas Centrifuge Process . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-2

6.3.1.2 Chemical Screening and Classification . . . . . . . . . . . . . . . . 6-3

6.3.1.3 Hazardous Chemicals and Chemical Interactions . . . . . . . 6-4

6.3.1.4 UF6 and Interaction Chemicals . . . . . . . . . . . . . . . . . . . . . . 6-4

6.3.2 Chemical Accident Sequences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-7

6.3.3 Chemical Accident Consequences . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-7

6.3.4 Items Relied on for Safety (IROFS) and Management Measures . . . . 6-7

6.3.4.1 Chemical-Process IROFS . . . . . . . . . . . . . . . . . . . . . . . . . . 6-7

6.3.4.2 Management Measures . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-8

6.3.5 Baseline Design Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-12

6.4 EVALUATION FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-13

6.5 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-13

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 Process Fire Hazards and Special Hazards . . . . . . . . . . . . . . . . . . . . 7-1

7.3.2 Accident Scenarios . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-4

7.3.2.1 Fire in the Centrifuge Test Facility . . . . . . . . . . . . . . . . . . . 7-4
ix

7.3.2.2 Fire in the CRDB . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-4

7.3.2.3 Fire Involving Cylinder Transporters/Movers . . . . . . . . . . . 7-5

7.3.2.4 Fire in the Cascade Halls . . . . . . . . . . . . . . . . . . . . . . . . . . 7-6

7.3.2.5 Fire in the Process Services Area . . . . . . . . . . . . . . . . . . . 7-6

7.3.2.6 Fire in the UF6 Handling Area/Blending and Liquid 

Sampling Area . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-7

7.3.2.7 Fire in the TSB . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-7

7.3.2.8 UBC Storage Pad . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-8

7.3.2.9 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-9

7.3.3 IROFS Related to Fire Safety . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-10

7.3.4 Facility Fire Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-10

7.3.4.1 Fire Safety Management Measures . . . . . . . . . . . . . . . . . 7-10

7.3.4.2 Facility Passive-Engineered Fire-Protection Systems . . . 7-12

7.3.4.3 Facility Active-Engineered Fire-Protection Systems . . . . . 7-12

7.3.4.4 Fire Safety and Emergency Response . . . . . . . . . . . . . . . 7-14

7.4 EVALUATION FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-15

7.5 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-15

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 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-3

8.3.5 Detection of Accidents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-3

8.3.6 Mitigation of Consequences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-4

8.3.7 Assessment of Releases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-4

8.3.8 Responsibilities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-4

8.3.9 Notification and Coordination . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-5

8.3.10 Information To Be Communicated . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-6

8.3.11 Training . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-6

8.3.12 Safe Shutdown (Recovery and Facility Restoration) . . . . . . . . . . . . . . 8-7

8.3.13 Exercises and Drills . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-7

8.3.14 Responsibilities for Developing and Maintaining the Emergency 

Program and Its Procedures Current . . . . . . . . . . . . . . . . . . . . . . . . . 8-9

8.4 EVALUATION FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-9

8.5 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-9

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

9.1 REGULATORY REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-1

9.2 REGULATORY ACCEPTANCE CRITERIA . . . . . . . . . . . . . . . . . . . . . . . . . . 9-2

9.3 STAFF REVIEW AND ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-2

9.3.1 Radiation Safety . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-2

9.3.1.1 As Low As is Reasonably Achievable (ALARA) Goals for 

Air and Liquid Effluent Control . . . . . . . . . . . . . . . . . . . . . . 9-2

9.3.1.2 Air Effluent Controls to Maintain Public Doses ALARA . . . . 9-4

9.3.1.3 Liquid Effluent Controls to Maintain Public Doses ALARA . 9-6

9.3.1.4 ALARA Reviews and Reports to Management . . . . . . . . . . 9-6

9.3.1.5 Waste Minimization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-6

9.3.1.6 Safe Handling of Radioactive Wastes . . . . . . . . . . . . . . . . . 9-7

9.3.2 Effluent and Environmental Monitoring . . . . . . . . . . . . . . . . . . . . . . . . 9-8

9.3.2.1 Air Effluent Monitoring . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-8

9.3.2.2 Liquid Effluent Monitoring . . . . . . . . . . . . . . . . . . . . . . . . . 9-11

9.3.2.3 Laboratory Quality Control . . . . . . . . . . . . . . . . . . . . . . . . 9-13

9.3.2.4 Environmental Monitoring . . . . . . . . . . . . . . . . . . . . . . . . . 9-13

9.3.3 ISA Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-14

9.4 EVALUATION FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-15

9.5 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-16

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

10.1 REGULATORY REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-1

10.2 REGULATORY ACCEPTANCE CRITERIA . . . . . . . . . . . . . . . . . . . . . . . . . 10-1

10.3 STAFF REVIEW AND ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-2

10.3.1 Conceptual Decontamination and Decommissioning Plan . . . . . . . . 10-2

10.3.1.1 Decommissioning Program . . . . . . . . . . . . . . . . . . . . . . . . 10-2

10.3.1.2 Decommissioning Steps . . . . . . . . . . . . . . . . . . . . . . . . . . 10-3

10.3.1.3 Management/Organization . . . . . . . . . . . . . . . . . . . . . . . . 10-6

10.3.1.4 Health and Safety . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-7

10.3.1.5 Waste Management . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-7

10.3.1.6 Security/Nuclear Material Control . . . . . . . . . . . . . . . . . . . 10-7

10.3.1.7 Recordkeeping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-7

10.3.1.8 Decontamination . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-8

10.3.1.9 Decommissioning Costs . . . . . . . . . . . . . . . . . . . . . . . . . 10-10

10.3.1.10 Financial Assurance for Decommissioning . . . . . . . . . . . 10-13

10.4 EVALUATION FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-15
xi

10.5 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-16

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

11.1 REGULATORY REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-1

11.2 REGULATORY ACCEPTANCE CRITERIA . . . . . . . . . . . . . . . . . . . . . . . . . 11-2

11.3 STAFF REVIEW AND ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-2

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

11.3.1.1 CM Policy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-2

11.3.1.2 Design Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-3

11.3.1.3 Document Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-4

11.3.1.4 Change Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-5

11.3.1.5 Assessments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-6

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

11.3.2.1 Corrective Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . 11-8

11.3.2.2 Preventative Maintenance (PM) . . . . . . . . . . . . . . . . . . . . 11-8

11.3.2.3 Surveillance and Monitoring . . . . . . . . . . . . . . . . . . . . . . . 11-8

11.3.2.4 Functional Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-9

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

11.3.3.1 Organization and Management of the Training Function

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-12

11.3.3.2 Analysis and Identification of Functional Areas Requiring

Training . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-12

11.3.3.3 Position Training Requirements . . . . . . . . . . . . . . . . . . . 11-12

11.3.3.4 Basis for Training and Training Objectives . . . . . . . . . . . 11-18

11.3.3.5 Lesson Plans and Other Training Guides . . . . . . . . . . . . 11-18

11.3.3.6 Evaluation of Trainee Accomplishment . . . . . . . . . . . . . . 11-18

11.3.3.7 Conduct of On-the-Job Training . . . . . . . . . . . . . . . . . . . 11-18

11.3.3.8 Evaluation of Training Effectiveness . . . . . . . . . . . . . . . . 11-19

11.3.3.9 Personnel Qualification . . . . . . . . . . . . . . . . . . . . . . . . . . 11-19

11.3.3.10 Periodic Personnel Evaluations . . . . . . . . . . . . . . . . . . . 11-20

11.3.4 Procedure Development and Implementation . . . . . . . . . . . . . . . . . 11-20

11.3.5 Audits and Assessments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-21

11.3.6 Incident Investigations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-23

11.3.7 Records Management . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-25

11.3.8 Other QA Elements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-26

11.4 EVALUATION FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-28

11.4.1 CM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-28

11.4.2 Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-29

11.4.3 Training and Qualification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-30

11.4.4 Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-30

11.4.5 Audits and Assessments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-30

11.4.6 Incident Investigations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-31

11.4.7 Records Management . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-31
xii

11.4.8 Other QA Elements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-31

11.5 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-33

12.0 MATERIAL CONTROL AND ACCOUNTING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12-1

13.0 PHYSICAL PROTECTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13-1

14.0 PHYSICAL SECURITY OF THE 

TRANSPORTATION OF SPECIAL NUCLEAR MATERIAL 

OF LOW STRATEGIC SIGNIFICANCE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14-1

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

APPENDIX A - ACCIDENT ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1

A.1 ACCIDENT ANALYSIS METHODOLOGY . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1

A.1.1 Selection of Representative Accident Scenarios . . . . . . . . . . . . . . A-1

A.1.2 Source Term Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1

A.1.3 NRC Performance Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . A-5

A.1.4 Consequence-Assessment Approach for Acute Health Effects . . . A-5

A.1.5 Consequence-Assessment Methodology for Chronic Health 

Effects . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-8

A.2 ACCIDENT ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-11

A.2.1 Generic Inadvertent Nuclear Criticality . . . . . . . . . . . . . . . . . . . . . A-11

A.2.2 Hydraulic Rupture of a UF6 Cylinder in the Blending and LiquidSampling Area . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-13

A.2.3 Natural Phenomena Hazard - Earthquake . . . . . . . . . . . . . . . . . . A-14

A.2.4 Fire in a UF6 Handling Area . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-14

A.2.5 Process Line Rupture in a Product Low-Temperature Takeoff 

Station (LTTS) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-16

A.3 ACCIDENT ANALYSIS SUMMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-16

A.4 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-18
xiii

TABLES

Table 1.2-1 Proposed Possession Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-9

Table 5.3-1 Safety Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-14

Table 5.3-2 Application of Safety Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-15

Table 5.3-3 Identification of IROFS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-32

Table 7.3-1 Identification of IROFS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-11

Table 9.3-1 Building Ventilation Air Discharge Locations Safety Criteria . . . . . 9-9

Table 9.3-2 Summary of Air and Liquid Effluent Sampling Points in the 

Radiological Effluent Monitoring Program . . . . . . . . . . . . . . . . . . 9-12

Table 9.3-3 Types of Samples Collected under the REMP . . . . . . . . . . . . . . . 9-14

Table A.1-1 Accident Types . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-2

Table A.1-2 Accident Scenarios Representative of Potential Facility Events . . . A-2

Table A.1-3 Definition of High- and Intermediate-Consequence Events . . . . . . A-5

Table A.1-4 Values of Atmospheric Dispersion Factors for the Facility

Boundaries . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-7

Table A.1-5 Ingestion Parameters Used in GENII to Calculate Collective

Radiological Dose to the Public from Accidents . . . . . . . . . . . . . . . A-9

Table A.1-6 Public Population in Sectors Surrounding the Facility . . . . . . . . . A-10

Table A.2-1 Health Effects Resulting from Inadvertent Nuclear Criticality . . . . A-12

Table A.2-2 Health Effects Resulting from Hydraulic Rupture of a UF6

Cylinder . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-14

Table A.2-3 Health Effects Resulting from an Earthquake . . . . . . . . . . . . . . . . A-15

Table A.2-4 Health Effects Resulting from Fire in a UF6 Handling Area . . . . . A-15

Table A.2-5 Acute Health Effects Resulting from Process Line Rupture in a

Product LTTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-16

Table A.3-1 Summary of Health Effects Resulting from Accidents . . . . . . . . . A-17
xiv

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xv

EXECUTIVE SUMMARY

On December 12, 2003, Louisiana Energy Services (LES) (the applicant) submitted, to the U.S.

Nuclear Regulatory Commission (NRC), an application requesting a license, under 10 CFR

Parts 30, 40, and 70, to possess and use byproduct, source, and special nuclear material

(SNM) in a gas centrifuge uranium enrichment facility. LES proposes that the facility be located

in Lea County, New Mexico, and have a nominal capacity of 3 million separative work units

(SWUs). (A SWU is a unit of enrichment that measures the effort required to separate isotopes

of uranium.) The facility will possess natural, depleted, and enriched uranium, and will enrich

uranium up to a maximum of 5 percent uranium-235. The applicant also requested a facility

clearance for classified information, under 10 CFR Part 95.

The NRC staff conducted its safety review in accordance with NUREG-1520, “Standard Review

Plan for the Review of a License Application for a Fuel Cycle Facility.” The staff’s safeguards

review involved reviews of the applicant’s Fundamental Nuclear Materials Control Plan

(FNMCP); the Physical Security Plan, which includes transportation security; and a “Standard

Practice Procedures Plan for the Protection of Classified Matter.” The staff also reviewed the

applicant’s Quality Assurance Program Description and Emergency Plan. Where the applicant’s

design or procedures should be supplemented, the NRC staff has identified license conditions

to provide assurance of safe operation.

The applicant also submitted an Environmental Report, which was used to prepare, in a

separate document, an Environmental Impact Statement for the facility.

A summary of NRC’s review and findings in each of the review areas is provided below:

General Information

The applicant provided an adequate description of the facility and processes so that the staff

has an overall understanding of the relationships of the facility features as well as the function of

each feature. Financial qualifications were properly explained and outlined in the application. 

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, administrative policies, and sufficient competent resources have been

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

“Integrated Safety Analysis” (ISA) and ISA 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. 
xvi

For nuclear criticality safety safe-by-design components, the applicant identified the hazards

and demonstrated that the failure of those components would be highly unlikely. No mitigated

event consequence exceeds the performance requirements of 10 CFR 70.61. The applicant

also provided adequate information about items relied on for safety (IROFS). License

conditions have been added to the license to ensure that IROFS boundaries will be defined

using the applicant’s IROFS boundary definition procedure and that the applicant will submit

license amendment requests if digital instrumentation and controls are used in IROFS.

Radiation Protection

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

application adequately describes: (a) the qualification requirements; (b) written radiation

protection procedures; (c) the radiation work permit (RWP) program; (d) the program for

ensuring that worker and public doses are as low as is reasonably achievable (ALARA); and (e)

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.

Nuclear Criticality Safety

The applicant provided adequate information to evaluate the Nuclear Criticality Safety (NCS)

program. The applicant committed to having an adequate group of qualified staff to develop,

implement, and maintain the NCS program in accordance with the facility organization and

administration and management measures. The program meets the regulatory requirements.

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 those 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 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 license application, meet safety requirements and provide reasonable

assurance that the facility is protected against fire hazards.

Emergency Management

The applicant provided an adequate Emergency Plan, for the facility, that meets the regulatory

requirements. The applicant commits to maintaining and executing an Emergency Plan 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. 

The requirements of the Emergency Plan are implemented through approved written

procedures.
xvii

Environmental Protection

The applicant committed to adequate environmental protection measures, including,”1)

environmental and effluent monitoring; and (2) effluent controls to maintain public doses ALARA

as part of the radiation protection program. The applicant’s proposed controls are adequate to

protect the environment and the health and safety of the public and comply with the regulatory

requirements.

Decommissioning

The applicant provided a conceptual decommissioning plan, for the facility, that addresses: (a)

contamination control; (b) control of worker exposures and waste volumes; (c) waste disposal;

(d) the final radiation survey; (e) control of SNM; (f) control of classified matter; and (g) recordkeeping for decommissioning.

The applicant provided a decommissioning funding plan, for the facility, that 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 applicant also submitted an

exemption request to allow for incremental funding for depleted uranium disposition based on

depleted uranium tails generation rates. The decommissioning funding plan and the

incremental approach for funding depleted uranium disposition costs will provide adequate

assurance for decommissioning funding because sufficient funding will be available to

decommission the facility and disposition the inventory of depleted uranium on-site at any point

in time. The applicant also provided proposed language for a surety bond, with a standby trust

agreement. The surety bond and standby trust agreement will be executed before the applicant

takes possession of licensed material. The applicant will update the site-specific cost estimate

at least every 3 years, to reflect inflation and changes in site inventories and conditions, that

could affect the cost of decommissioning. A license condition has been added to the license to

ensure that the applicant takes possession of no licensed material until the surety bond and

standby trust agreement are executed and are acceptable to NRC. The decommissioning

funding plan is acceptable because it provides sufficient funding to ensure 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 information about management measures that will be applied to the

project. The information describes: (a) the overall configuration management program and

policy; (b) the maintenance program; (c) training; and (d) the process for the development,

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

assessments program as well as incident investigations and 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).
xviii

Materials Control and Accountability

The applicant provided information describing the Fundamental Nuclear Material Control Plan

(FNMCP) for the project. The FNMCP describes the programs to be used to control and

account for SNM in the facility. The program meets the applicable regulatory requirements in

Part 74.

Physical Protection

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

to protect SNM of low strategic significance used and possessed at the facility. This information

is acceptable and meets the requirements in Part 73.

The applicant also provided information on the protection of classified matter, including security

controls and procedures, to ensure that classified matter is used, processed, stored,

reproduced, transmitted, transported, and destroyed. This program is acceptable and in

accordance with the regulatory requirements in Part 95 for a facility clearance.

Transportation Security

The applicant provided information in the Physical Security Plan on the policies, methods, and

procedures to be implemented to protect SNM of low strategic significance in transit to and from

the facility. This information is acceptable and meets the requirements in Part 73.
xix

LIST OF ACRONYMS AND ABBREVIATIONS

ACI American Concrete Institute

AEGL Acute Exposure Guideline Level

AHU Air handling unit

AISC American Institute of Steel Construction

ALARA As low as is reasonably achievable

Al2O3 Aluminum oxide

ANSI American National Standards Institute

ARF Airborne release fraction

ASCE American Society of Civil Engineers

ASM Additional Safety Measures

ASME American Society of Mechanical Engineers

BDC Baseline design criteria

BNFL-EL BNFL Enrichment Limited

CAA Controlled access area

CAP Corrective Action Program

CAS Central alarm station

CCTV Closed circuit television

CEC Claiborne Enrichment Center

CEDE Committed Effective Dose Equivalent

CFR Code of Federal Regulations

cm centimeter

CM Configuration management

CPD Core Plant Design

CRDB Cylinder Receipt and Dispatch Building

CS Chemical safety

CUB Central Utility Building

DOE U.S. Department of Energy

DOEQAP U.S. Department of Energy Quality Assurance Program

DR Damage ratio

EECP Entry/exit control point

EO Emergency Organization

EOC Emergency Operations Center

EP Emergency Plan

EPA U.S. Environmental Protection Agency

EPIP Emergency Plan Implementing Procedures

ER Environmental Report

ERO Emergency response organization

FAA Federal Aviation Administration
xx

FNMCP Fundamental Nuclear Material Control Plan

FOCI Foreign ownership, control, or influence

ft feet

ft/s feet per second

g acceleration of gravity

GEVS Gaseous Effluent Ventilation System

gpm gallons per minute

ha hectare

HAZOP hazard and operability

HEPA High efficiency particulate air

HF Hydrogen fluoride

HPS Health Physics Society

HS&E Health, safety, and environment

HVAC Heating, ventilating, and air conditioning

ICBO International Conference of Building Officials

IEC International Electrotechnical Commission

in inch

IRB Industrial revenue bond

IROFS Items relied on for safety

ISA Integrated safety analysis

ISO International Organization for Standardization

kg kilogram

km kilometer

kPa kiloPascals

kPa/s kiloPascals per second

L liter

lb pound

LEL Lower explosive limit

LES Louisiana Energy Services

LLW Low-level radioactive waste

LPF Leak path factor

Lpm liters per minute

LTTS Low temperature takeoff station

m meter

m3 cubic meter

MAPEP Mixed Analyte Performance Evaluation Program

MAR Material at risk

MC&A Material control and accounting

mg milligram

mi mile

mm millimeter

MOU Memorandum of Understanding
xxi

m/s meter per second

MT metric ton

NAC National Advisory Committee

Na2CO3 Sodium carbonate

NaF Sodium fluoride

NCS Nuclear criticality safety

NELAC National Environmental Laboratory Accreditation Conference

NFPA National Fire Protection Association

NIOSH National Institute for Occupational Safety and Health

NMAC New Mexico Administrative Code

NOAA National Oceanic and Atmospheric Administration

NRC Nuclear Regulatory Commission

OSHA U.S. Occupational Safety and Health Administration

PFPE Perfluorinated polyether

PM Preventive maintenance

psf pounds per square foot

psf/s pounds per square foot per second

psi pounds per square inch

psia pounds per square inch absolute

QA Quality assurance

QAPD Quality Assurance Program Description

RAI Request for additional information

RASCAL Radiological Assessment System for Consequence Analysis

rem Roentgen equivalent man

REMP Radiological Environmental Monitoring Program

RF Respirable fraction

RP Radiation protection

RWP Radiation Work Permit

SAR Safety Analysis Report

SER Safety Evaluation Report

SM Source Material

SNM Special Nuclear Material

SNM-LSS Special Nuclear Material - Low Strategic Significance

SOP Standard Operating Procedure

SPPP Standard Practice Procedures Plan

SRC Safety Review Committee

SSC Structures, systems, and components

Sv Sievert

SWU Separative Work Unit

T Total Likelihood Index

TEEB Treated Effluent Evaporative Basin

TID Tamper indicating device
xxii

TLD Thermoluminescent dosimeter

TSB Technical Services Building

TWA Time-weighted average 

UBC Uranium byproduct cylinder

UCN Ultra-Centrifuge Nederland NV

μg microgram

UF4 Uranium tetrafluoride

UF6 Uranium hexafluoride

UO2F2 Uranyl fluoride

UPS Uninterruptible power supply

USEC U.S. Enrichment Corporation

USGS U.S. Geological Survey

wt weight
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 proposed

Louisiana Energy Services (LES) facility and process description is to determine whether the

application includes an overview of the facility layout and a summary description of the

proposed processes. A more detailed description of the facility and processes is contained in

the “Integrated Safety Analysis (ISA) Summary” (LES, 2005b).

1.1.1 REGULATORY REQUIREMENTS

The regulations in 10 CFR 30.33, 10 CFR 40.32, and 10 CFR 70.22 require each application for

a license to include information on the proposed activity and the equipment and facilities that

will be used by the applicant to protect health and minimize danger to life and property. In

addition, the regulations in 10 CFR 70.65 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.

1.1.2 REGULATORY ACCEPTANCE CRITERIA

The acceptance criteria applicable to NRC’s review of the facility and process description

section of the application are contained in Section 1.1.4.3 of the “Standard Review Plan for the

Review of a License Application for a Fuel Cycle Facility,” NUREG-1520 (NRC, 2002).

1.1.3 STAFF REVIEW AND ANALYSIS

In Section 1.1 of the Safety Analysis Report (SAR) (LES, 2005a), the applicant provides a

summary description of the proposed gas centrifuge uranium enrichment plant and processes. 

This description includes discussion of the major chemical and mechanical processes to be

used in the facility. The facility is proposing to use a gas centrifuge enrichment process to

enrich uranium from its natural isotopic concentration of about 0.7 percent uranium-235 (U-235)

to 5 percent U-235. The proposed plant will have a nominal enrichment capacity of 3 million

Separative Work Units (SWUs). (A SWU is a measure of the effort required to perform isotopic

separation.) The process uses uranium in the chemical form of uranium hexafluoride (UF6). 

Gaseous UF6 enters a high-speed rotor at subatmospheric conditions where centrifugal forces

press the heavier isotope of uranium, uranium-238 (U-238), to the outer wall of the rotor. The

lighter isotope, U-235, remains closer to the center, away from the rotor wall. Internal scoops

are used to collect the heavier and lighter fractions and circulate them to other centrifuges piped

in a cascade arrangement.

The proposed plant will be constructed to have three Separations Building Modules, each

having two Cascade Halls, with each Cascade Hall having eight cascades. Each Separations

Building consists of a UF6 Feed System, Cascade Systems, a Product Take-off System, and a

Tails Take-off System. The plant also has a Product Liquid Sampling System and a Product

Blending System.
1-2

Natural uranium feed is shipped to the plant primarily by truck in cylindrical steel containers

having a capacity of up to 12.7 metric tonnes (MT) (14 tons) of UF6. Under ambient conditions,

the UF6 is a solid. Feed containers are vented to remove air and hydrogen fluoride (HF) gases

and then heated to sublime the solid UF6 to a gas. The feed system is designed to preclude the

UF6 from becoming liquid. The light gases and gaseous UF6 pass through the Feed Purification

Subsystem to remove the light gases that are directed through the Gaseous Effluent Ventilation

System (GEVS) to ensure that HF and UF6 are removed and not released to the atmosphere. 

After the venting is complete, the UF6 feed from the Solid Feed Stations is directed to a cascade

for enrichment.

After enrichment in gas centrifuge machines, both depleted and enriched products are

withdrawn from the cascade and desublimed at subatmospheric pressure in the Tails Take-off

System and the Product Take-off System, respectively. Tails and Product Take-off Systems are

designed to preclude UF6 from becoming a liquid.

Sampling to verify the assay level is performed in the Product Liquid Sampling Autoclave. In the

autoclave, UF6 is heated to a liquid; the cylinder is tilted so that UF6 can flow into sample

manifold and sample bottles; and the cylinder is returned to its original horizontal position. This

is the only system in the plant where UF6 is in a liquid form.

To produce enriched uranium meeting customer-assay specifications, the Product Blending

System is used to mix enriched uranium at two different enrichment levels to one meeting the

customer specifications. This system can also be used to transfer product between cylinders.

Facility information contained in the ISA Summary (LES, 2005b) is provided in the application in

layout drawings of the plant buildings and the location of plant systems within the buildings. 

Geographical features and transportation routes are also provided on these drawings.

The proposed facility is expected to possess natural, enriched, and depleted uranium. It is

expected to handle, on an annual basis, approximately 690 nominal 12.5-MT (14-ton) or 9.5-MT

(10.5-ton) natural uranium feed cylinders; 350 nominal 2-MT (2.5-ton) enriched-uranium product

cylinders; and 625 nominal 12.5-MT (14-ton) depleted uranium tails cylinders.

Gaseous airborne effluents will be released from the proposed facility. The applicant estimates

that less than 10 grams (0.35 ounces (oz)) of uranium and less than 1 kilogram (kg) (2.2 pounds

(lbs)) of HF will be released annually in 2.47 × 109

 cubic meters of air discharge. These

effluents are significantly below 10 CFR Part 20 and U.S. Environmental Protection Agency

National Emission Standards for Hazardous Air Pollutants (NESHAPs) airborne release limits.

Liquid discharges include contaminated process effluents, cooling tower blowdown, and

stormwater discharges. Liquid effluents will be significantly below Part 20 liquid effluent

requirements.

Wastes expected to be generated include non-hazardous industrial, Class A radioactive,

hazardous, and mixed wastes. Construction wastes will also be generated in construction of

the plant. Radioactive wastes will be disposed of at properly licensed low-level radioactive

waste disposal facilities. Hazardous chemical wastes will be properly treated and disposed of at

permitted treatment and disposal facilities. Mixed low-level radioactive and chemically

hazardous wastes will be treated and disposed of at facilities having the proper licenses and
1-3

permits for these wastes. Depleted uranium tails will be stored on-site on the Uranium

Byproduct Cylinder (UBC) pad until they are transferred to another licensee for commercial use

or they are designated for disposal as waste. If designated as waste, the applicant is

proposing to use either a commercial disposition path or the U.S. Department of Energy (DOE)

disposition path set out in the USEC Privatization Act of 1996. The applicant has committed to

not store depleted uranium tails for longer than the 30-year life of the plant.

1.1.4 EVALUATION FINDINGS

The staff has reviewed the proposed general facility and process descriptions according to

Section 1.1 of the Standard Review Plan. The applicant has adequately described: (1) the

facility and processes so that the staff has an overall understanding of the relationships of the

facility features; and (2) the function of each feature. The staff concludes that the applicant has

met the requirements and acceptance criteria applicable to this section.

1.2 INSTITUTIONAL INFORMATION

The purpose of NRC’s review of institutional information is to establish whether the license

application includes adequate information identifying the applicant, the applicant’s

characteristics, and the proposed activity.

1.2.1 REGULATORY REQUIREMENTS

The regulations in 10 CFR 30.32 and 10 CFR 40.31 require each application for a license to

include: (a) information on the identity of the applicant; (b) name, chemical and physical form,

and maximum amount that will be possessed; and (c) purpose for which the licensed material

will be used. The regulations in 10 CFR 70.22 require each application for a license to include: 

(a) information on the corporation applying for a license; (b) the location of the principal office;

(c) the names and citizenship of the principal officers; (d) information concerning ownership and

control; (e) the proposed site activities; (f) financial qualifications; and (g) the name, amount,

and specifications of the licensed material to be used. The regulations in 10 CFR Part 95

contain provisions for obtaining a facility security clearance. The regulations in 10 CFR 140.13b

require applicants for uranium enrichment facilities to provide and maintain liability insurance.

1.2.2 REGULATORY ACCEPTANCE CRITERIA

The acceptance criteria applicable to NRC’s review of the facility and process description

section of the application are contained in Section 1.2.4.3 of the “Standard Review Plan for the

Review of a License Application for a Fuel Cycle Facility,” NUREG-1520 (NRC, 2002).

1.2.3 STAFF REVIEW AND ANALYSIS

1.2.3.1 Corporate Identity

In Section 1.2.1 of the SAR (LES, 2005a), the applicant provides information on the corporate

organization. LES is a Limited Partnership chartered in Delaware. If was formed solely to 

provide uranium enrichment services to the commercial nuclear power sector. LES has a 100-

percent-owned subsidiary, NEF Series 2004, LLC, a limited liability company, organized under
1-4

the laws of the State of Delaware, formed to purchase Industrial Revenue Bonds issued by Lea

County. The General Partners are as follows:

1. Urenco Investments, Inc., a Delaware corporation and wholly owned subsidiary of

Urenco Limited, a corporation formed under laws of the United Kingdom and owned

equally by BNFL Enrichment Limited (BNFL-EL); Ultra-Centrifuge Nederland NV (UCN);

and Uranit GmbH (Uranit) companies, formed under English, Dutch, and German law,

respectively. BNFL-EL is wholly owned by British Nuclear Fuels plc, which is wholly

owned by the Government of the United Kingdom. UCN is 99 percent owned by the

Government of the Netherlands and 1 percent owned by the Royal Dutch Shell Group,

DSM, Koninklijke Philips Electronics N.V., and Stork N.V. Uranit is owned equally by

Eon Kernkraft GmbH and RWE Power AG, which are corporations formed under laws of

the Federal Republic of Germany.

2. Westinghouse Enrichment Company, LLC, a Delaware limited liability company and

wholly owned subsidiary of Westinghouse Electric Company LLC, also a Delaware

limited liability company, whose ultimate parent, through two intermediary Delaware

corporations and one corporation formed under the laws of the United Kingdom, is

British Nuclear Fuels plc, which is wholly owned by the Government of the United

Kingdom.

The Limited Partners are as follows:

1. Urenco Deeinemingen B.V., a Netherlands corporation and wholly owned subsidiary of

Urenco Nederlands B.V.

2. Westinghouse Enrichment Corporation, LLC, a Delaware limited liability company and

wholly owned by Westinghouse Electric Company, LLC.

3. Entergy Louisiana, Inc., a Louisiana corporation and wholly owned subsidiary of Entergy

Corporation, a publicly owned Delaware corporation and a public utility holding

company.

4. Claiborne Energy Services, Inc., a Louisiana corporation and wholly owned subsidiary of

Duke Energy Corporation, a publicly owned North Carolina corporation.

5. Cenesco Company, LLC, a Delaware limited liability company and wholly owned

subsidiary of Exelon Generation Company, LLC, a Pennsylvania limited liability

company.

6. Penesco Company, LLC, a Delaware limited liability company and wholly owned

subsidiary of Exelon Generation Company, LLC, a Pennsylvania limited liability

company.

Urenco owns 70.5 percent of the partnership while Westinghouse Electric Company owns 19.5

percent. Entergy, Duke Energy Corporation, and Exelon Generation Company own the

remaining 10 percent in equal shares.

No other companies will be present or operating on the uranium enrichment plant property other
1-5

than where the applicant has contracted such services. The principal location for business is

Albuquerque, New Mexico.

The applicant provided the name of the President of LES, who is a citizen of the United States.

1.2.3.2 Foreign Ownership, Control, or Influence

With respect to a foreign ownership, control, or influence (FOCI) determination for LES’ National

Enrichment Facility, the NRC staff received a letter from the Department of Energy (DOE) dated

March 31, 2005 (DOE, 2005), which states in part that “...any additional FOCI mitigation

measures placed on LES would provide no additional benefit to the National Security of the

U.S.” The letter further recommends that the NRC waive the requirement for FOCI mitigation

associated with the granting of a nuclear facility licence to LES. The NRC accepts this finding

by DOE based on an Interagency Agreement between NRC and DOE dated May 6, 2002. 

(DOE, 2002).

1.2.3.3 Financial Qualifications

1.2.3.3.1 Project Costs

The applicant estimates the total construction cost of the facility to be approximately $1.2 billion,

in 2002 dollars, which excludes escalation, interest during construction, tails disposition,

decommissioning, and any replacement equipment required during the operating life of the

facility. The facility SAR (LES, 2005a) and supporting supplements addressing NRC’s April 19,

2004, Request for Additional information (RAI), provide detailed bases that supported the $1.2

billion estimate. The supporting supplements included detailed proprietary construction cost

estimates for the facility.

As part of the financial review, before starting the detailed review of the cost estimate, the staff

conferred with the technical reviewers assigned to evaluate the support systems/structures

necessary to support the safe operation of the facility to confirm that the necessary systems had

been identified in the SAR (LES, 2005a). The staff also conducted a detailed review of the

SAR, Section 1.1.1 (LES, 2005a), which provided a detailed description of each supporting

structure/system, and then compared the support systems for each building with the systems

identified in the cost estimate, to confirm that the cost estimate and the facility description were

consistent. The cost estimate is based on a reasonable estimate of the cost of the supporting

systems and structures, as well as confirming that all the major equipment necessary to support

safe operation were included.

The applicant identified the principal buildings necessary to support the operation of the facility. 

The buildings are: (a) Separation Building/3-modules; (b) Technical Services Building (TSB); (c)

Centrifuge Assembly Building; (d) site infrastructure; (e) central utilities; (f) Cylinder Receipt and

Dispatch Building (CRDB); and (g) blending and liquid sampling area. The applicant also

identified each of the principal components/systems necessary to support its operation and

further divided the cost estimate by components/systems and buildings necessary to support

the operation of the facility. For each of the major structures, the applicant typically divided the

systems into the following: 1) engineering and project management support; 2) centrifuge

mounting and equipment; 3) UF6 systems; 4) control and instrumentation; 5) auxiliary systems;

6) electrical systems; 7) site, building, and landscape; 8) miscellaneous costs including start-up
1-6

cost; and 9) a contingency of approximately of 10 percent. The staff reviewed each of the

detailed costs for the major facilities and its supporting components, and based on its review,

the staff has concluded the costs for each of the major structures is reasonable.

The validity of the estimated cost and its supporting assumptions were also key factors in

determining if the applicant is financially qualified to construct the facility. The applicant stated

that it plans to meet contingencies for cost overruns and construction revenue shortfalls in

several ways. Unforeseen construction contingencies will be minimized by the use of a turnkey

contractor for the engineering, procurement, and construction of the facility. For cost overruns

not covered under the turnkey provisions of the contract, the applicant will seek additional

partner equity contributions. However, if cost overruns are much higher than could be

anticipated, the applicant would cancel the project and leave an allowance for site stabilization. 

The facility will begin operation in a phased approach, with each separation module capable of

production of one-third of the plant’s capacity. This will allow the facility to generate income

while coming up to full production.

The staff considers that the construction cost estimate, as presented in the application, and

supporting supplements in response to the staff’s RAIs, is reasonable and, therefore, the staff

concluded that the $1.2 billion estimate is a reasonable estimate to construct the facility.

1.2.3.3.2 Financial Qualifications

The applicant made commitments that construction of the facility will not begin before funding is

fully committed. Of this funding (equity and debt), the applicant will have in place, before

construction, a minimum of equity contributions of 30 percent of the project’s estimated costs of

$1.2 billion from the parents and affiliates of the partners, and firm commitments ensuring funds

for the remaining project costs. The applicant plans to fund the construction phase of the

project with a mix of approximately 50 percent debt and 50 percent equity contributions by the

two major partners. The applicant’s reliance on approximately 50 percent equity is viewed as a

positive endorsement because, by contrast, some analogous construction projects rely on 100

percent financing, which often proves to be difficult to secure from financial institutions.

The applicant has no reported income statements. However, the partners have assets to

support their respective equity ownership portions of LES. Urenco Investments, Inc., is a wholly

owned subsidiary of Urenco Limited, which in turn is owned in equal shares by BNFL EL; UCN;

and Uranit - companies that are formed under English, Dutch and German law, respectively. 

BNFL EL is wholly owned by British Nuclear Fuels plc, which is wholly owned by the

Government of the United Kingdom. UCN is 99 percent owned by the Government of the

Netherlands, with the remaining 1 percent owned collectively by private consortiums. Uranit is

equally owned by Eon Kernkraft GmbH and RWE Power AG, which are both German

companies.

For the year ending December 31, 2003, Urenco Group had total assets of €1.49 billion, with

cash assets of €14.4 million. Urenco Group’s net income in 2003 was €107.9 million. Urenco

Limited is the holding company for the Urenco Group.

Cenesco Company and Penesco Company are both wholly owned subsidiaries of Exelon

Generation Company. For the year ending December 31, 2003, Exelon Generation Company

had total assets of $14.76 billion, with cash or near-cash assets of $233 million. The company
1-7

sustained a net loss of $133 million in 2003. Net losses in 2003 can be attributed primarily to

operating expenses, in particular the costs of purchased fuel, purchased power, impairment of

long-lived assets, and other operating and maintenance expenses. Furthermore, for the 9

months ending September 30, 2004, the company had a positive net income of $599 million.

Duke Energy Corporation, a publicly held North Carolina corporation, is the owner of Claiborne

Energy Services, Inc., which is also a 3.33 percent owner of LES. For the year ending

December 31, 2003, Duke Energy Corporation had total assets of $56.2 billion, with cash or

near-cash assets of $1.16 billion. Duke Energy Corporation sustained a net loss of $1.3 billion

in 2003. Net losses in 2003 can be attributed primarily to operating expenses, in particular the

costs of purchased natural gas and petroleum products. Furthermore, for the 9 months ending

September 30, 2004, the corporation had a positive net income of $1.13 billion.

Entergy Corporation, a public utility holding company, is the owner of Entergy Louisiana Inc.,

which is a 3.33 percent owner of LES. For the year ending December 31, 2003, Entergy

Corporation had total assets of $28.55 billion, with cash or near-cash assets of $692 million. 

Entergy Corporation’s net income in 2003 was $950 million.

Westinghouse Enrichment Company, LLC, is a wholly owned subsidiary of Westinghouse

Electric Company, LLC, whose ultimate parent, through two intermediary Delaware corporations

and one corporation under United Kingdom laws, is British Nuclear Fuels plc. British Nuclear

Fuels plc is wholly owned by the Government of the United Kingdom.

For the year ending March 31, 2004, Westinghouse Electric Company had total assets of £1.02

billion. The company’s pre-tax net income was £17 million for that financial year. 

Westinghouse Electric Company is a subsidiary of British Nuclear Fuels, which had total assets

of £23.94 billion for the year ending March 31, 2004. British Nuclear Fuels sustained a net loss

of £194 million after taxes for that financial year (pre-tax losses were £299 million).

The remaining 50 percent of the estimated $1.2 billion construction costs will be financed

through financial institutions and bond holders.

Lea County will serve as the lessor-owner of the facility during the 30-year term of Industrial

Revenue Bonds (IRB) issuance by the State of New Mexico. In this capacity, Lea County will

hold the legal title to the uranium enrichment facility, including all related buildings, storage,

infrastructure, and equipment, and will hold legal title or a possessory interest in the site on

which the facility is located during the term of the IRB. This financial structure will allow the

applicant to take advantage of certain tax abatements, tax avoidance, and make other

payments in lieu of taxes available under New Mexico law. The IRB is not a vehicle for

financing the plant.

Lea County will have no authority to operate the facility as a business or otherwise use or

acquire the facility for any purpose, except in its limited role as lessor. During the term of the

lease, the applicant is solely responsible, on behalf of, and as agent for, the County, for

acquiring, constructing, and installing the equipment into the facility. At the conclusion of the

30-year lease, which corresponds to the 30-year term of the IRB, the applicant will purchase the

land and facility from Lea County for the sum of $1.00.

On December 3, 2003, the applicant announced that the first round of contracts with several
1-8

U.S. nuclear power plants, including Exelon, were signed. These contracts represent at least

70 percent of the facility’s first 10 years of production. As the project construction progresses,

LES will make a decision to continue, based on a comparison of future incremental construction

and operations and maintenance costs to the expected revenues generated from enrichment

services sales.

The NRC staff finds that LES and its partner-owners appear to be financially qualified to build

and operate the proposed facility, in accordance with 10 CFR 70.23(a)(5). The applicant

identified sources of debt and equity for construction, and has reasonable assurance of

securing additional financial resources, if needed.

1.2.3.3.3 Liability Insurance

Under 10 CFR 140.13b, a uranium enrichment facility is required to carry liability insurance to

cover public claims arising from any occurrence, within the U.S. that causes, within or outside

the U.S., bodily injury, sickness, disease, death, loss of, or damage to, property, or loss of use

of property arising from the radioactive, toxic, explosive, or other hazardous properties of

chemicals containing licensed material. The applicant is proposing to have and maintain up to

$300 million to satisfy the 10 CFR 140.13b requirement. The applicant has already obtained a

nuclear energy liability policy with a limit of $1 million as a standby policy until the facility is

ready to begin operations. At that time, the applicant will increase the amount to approximately

$300 million.

Because full liability insurance coverage will not be provided until prior to receipt of licensed

material, NRC staff is imposing the following license condition:

“The licensee shall provide proof of full liability insurance as required under 10 CFR

140.13b, at least 30 days prior to the planned date for obtaining licensed material. If the

licensee is proposing to provide less than $300 million of liability insurance coverage,

the licensee shall provide, to the NRC for review and approval, an evaluation supporting

liability insurance coverage in amounts less than $300 million at least 120 days prior to

the planned date for obtaining licensed material.”

1.2.3.4 Type, Quantity, and Form of Licensed Material

Table 1.2-1 of the SAR (LES, 2005a) lists the type, quantity, and form of the licensed material

proposed for possession. The applicant proposes to use and possess the amounts of special

nuclear material, source material, and byproduct material given in Table 1.2-1. The quantities of

Tc-99 and transuranics from residual contamination as a consequence of the historical feed of

recycled uranium at other facilities are expected to have no significant radiological impact. 

1.2.3.5 Authorized Uses

The application is for the issuance of licenses under 10 CFR Parts 30, 40, and 70. The

applicant is proposing to use special nuclear material and source material in the enrichment of

uranium. The uranium enrichment services would be sold to clients for the production of lowenriched uranium that would be ultimately used in the manufacture of fuel for commercial

nuclear power plants. Byproduct material would be used in instrument-calibration sources and

may be present as contamination as a consequence of the historical feed of recycled uranium at
1-9

other enrichment facilities. Feed cylinders that have been previously used to transport or store

recycled uranium must be decontaminated before being allowed on the facility site. In addition,

natural UF6 supplied to the facility will meet American Society for Testing and Materials (ASTM)

ASTM C787, “Standard Specification for Uranium Hexafluoride for Enrichment” (ASTM, 2003),

and periodic audits of suppliers will be performed to ensure that these conditions are met. The

applicant intends to identify specific byproduct calibration sources in future license amendment

requests. The applicant proposed a 30-year license term. The applicant also requested

approval of a classified-matter facility clearance, under 10 CFR Part 95.

Table 1.2-1

Proposed Possession Limits

Source or Special Nuclear

Material

Physical and Chemical Form Maximum Amount to be

Possessed at Any One Time

Uranium (natural and

depleted) and daughter

products

Physical: Solid, Liquid, and

Gas

Chemical: UF6, UF4, UO2F2,

oxides and other compounds

136,120,000 kg (300,093,231

lbs)

Uranium enriched in isotope

U-235 up to 5 percent by

weight and uranium daughter

products

Physical: Solid, Liquid, and

Gas

Chemical: UF6, UF4, UO2F2,

oxides and other compounds

545,000 kg (1,201,519 lbs)

Tc-99, transuranic isotopes

and other contamination

Any Amount that exists as

contamination as a

consequence of the historical

feed of recycled uranium at

other facilities

Note: Tc-99 - Technetium-99

UF4 - Uranium Fluoride

UO2F2 - Uranyl Fluoride

1.2.3.6 Special Exemptions or Special Authorizations

In Section 1.2.5 of the SAR (LES, 2005a), the applicant addressed an exemption request to 10

CFR 40.36 and 10 CFR 70.25 to provide incremental funding for decommissioning to reflect its

phased approach for enrichment capacity at the facility and its expected depleted uranium tails

generation rate. As discussed in Section 10.2.2 of the SAR (LES, 2005a), the applicant stated

that it would initially provide funding for the projected cost of facility decontamination and

decommissioning, assuming operation at full capacity, and disposition of the tails generated

during the first three years of operation. Thereafter, the applicant will provide NRC with revised

funding instruments for depleted uranium disposition on an annual forward-looking incremental

basis. In the event that the applicant does not employ all projected modules as expected,

updates required under 10 CFR 40.36 and 10 CFR 70.25 could reflect a corresponding

reduction in the anticipated facility decommissioning costs based on the actual number of
1-10

modules used. NRC staff will review revisions to the cost estimate and the financial instrument,

which are presented in Section 10.2.2 of the SAR (LES, 2005a), before the applicant takes

possession of licensed material. NRC staff will also review all subsequent revisions to the cost

estimate and financial instruments.

Under 10 CFR 40.14 and 10 CFR 70.17, the Commission may grant exemptions from the

requirements of the regulations as it determines are authorized by law and will not endanger life

or property or the common defense and security and are otherwise in the public interest. NRC

staff evaluated the exemption request and determined that such exemption is not prohibited by

law. Staff also determined that, because the incremental funding approach proposed by the

applicant will provide funding for the all applicant’s decommissioning obligations at any point

time, the approach will not endanger life or property or the common defense and security. 

Because the incremental funding approach will reduce the applicant’s expenses from having to

fund a 30-year decommissioning obligation when, in actuality, the decommissioning obligations

prior to the end of the 30-year operating period are less, the staff has determined that the

proposed approach will be in the public interest by reducing unnecessary regulatory costs. 

Therefore, the staff grants the requested exemption as provided in Section 1.2.5 of the SAR. A

license condition will be included in the license that will address the applicant’s commitments for

updating the decommissioning funding plan over time. This license condition is discussed

further in Section 10.3.1.10 of this SER.

1.2.3.7 Security of Classified Matter

The purpose of this review is to verify that the applicant provided sufficient information to

conclude that there is an adequate Standard Practice Procedures Plan (SPPP) for the

protection of classified matter at the proposed facility to be located in Lea County, New Mexico,

and a facility clearance can be issued.

1.2.3.7.1 Regulatory Requirements

10 CFR 70.22(m) provides the regulatory requirements for the SPPP that describes the facility’s

proposed security procedures and controls, as set forth in 10 CFR 95.15(b).

The applicable portion of 10 CFR 70.22(m) identifies that the requirements to protect against

unauthorized viewing of classified enrichment equipment and unauthorized disclosure of

classified matter are contained in 10 CFR Parts 25 and 95.

1.2.3.7.2 Regulatory Acceptance Criteria

The LES SPPP was reviewed for compliance with the requirements of 10 CFR Parts 25 and 95,

by using “Standard Practice Procedures Plan Standard Format and Content for the Protection of

Classified Matter for NRC Licensee, Certificate Holder and Others Regulated by the

Commission” (NRC, 1999).

1.2.3.7.3 Staff Review and Analysis

The staff reviewed and evaluated information provided by LES in the facility’s proposed security

procedures and controls to ensure that classified matter is used, processed, stored, reproduced,

transmitted, transported, and destroyed in accordance with the requirements of 10 CFR Parts
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25 and 95.

NRC staff reviewed the LES SPPP and found it satisfied the requirements of 10 CFR Parts 25

and 95. The applicant has made commitments that meet the requirements of 10 CFR Parts 25

and 95 by providing an acceptable SPPP that establishes controls to ensure that classified

matter is used, processed, stored, reproduced, transmitted, transported, and destroyed only

under conditions that will provide adequate protection and prevent access by unauthorized

persons. By meeting these requirements, the applicant complies with the requirements of 10

CFR 70.22(m). On the basis of these findings, the staff concludes that the SPPP is acceptable

for implementation.

1.2.4 EVALUATION FINDINGS

The staff reviewed the institutional information for the proposed LES uranium enrichment facility,

according to Section 1.2 of the Standard Review Plan. The applicant has adequately described

and documented the corporate identity, structure, and financial information, and is in

compliance with those parts of 10 CFR 30.32, 10 CFR 40.31, 10 CFR 70.22, and 10 CFR 70.65

related to institutional information.

The staff reviewed the information provided by the applicant on liability insurance. This

information meets the requirements of 10 CFR 140.13b. Because full liability insurance

coverage will not be provided until prior to receipt of licensed material, NRC staff is imposing the

following license condition:

“The licensee shall provide proof of full liability insurance as required under 10 CFR

140.13b, at least 30 days prior to the planned date for obtaining licensed material. If the

licensee is proposing to provide less than $300 million of liability insurance coverage,

the licensee shall provide, to the NRC for review and approval, an evaluation supporting

liability insurance coverage in amounts less than $300 million at least 120 days prior to

the planned date for obtaining licensed material.”

In addition, in accordance with 10 CFR 30.32, 10 CFR 40.31, and 10 CFR 70.22(a)(2) and (4),

the applicant has adequately described the types, forms, and quantities and proposed purpose

and authorized uses of licensed materials to be permitted at the facility. The applicant provided

information on an exemption request related to decommissioning funding that meets the

requirements of 10 CFR 40.14 and 10 CFR 70.17. The applicant has also adequately described

information related to FOCI and its plans to secure classified matter for a facility clearance

under 10 CFR Parts 25 and 95. The staff concludes that the applicant has met the

requirements and acceptance criteria applicable to this section.

1.3 SITE DESCRIPTION

The purpose of a site description review is to determine whether the information provided by an

applicant adequately describes the geographic, demographic, meteorological, geologic,

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

description is a summary of the information that the applicant used in preparing the

environmental report, emergency plan, and integrated safety analysis summary.
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1.3.1 REGULATORY REQUIREMENTS

The regulations in 10 CFR 30.33, 10 CFR 40.32, 10 CFR 70.22, 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).

1.3.2 REGULATORY ACCEPTANCE CRITERIA

The acceptance criteria applicable to the NRC review of the site description section of the

application are contained in Section 1.3.4.3 of NUREG–1520 (NRC, 2002).

1.3.3 STAFF REVIEW AND ANALYSIS

1.3.3.1 Site Geography

1.3.3.1.1 Location

The proposed site is in Southeastern New Mexico in Lea County, approximately 1.6 km (1 mi)

west of the New Mexico-Texas border on the north side of New Mexico Highway 234. Andrews

County, Texas, lies across the border from the site. The site is about 8 km (5 mi) east of

Eunice, New Mexico, and 32 km (20 mi) south of Hobbs, New Mexico. The site is 220 ha (543

acres) in size and is located within County Section 32, Township 21 South, Range 38 East. The

site is owned by Lea County.

The proposed site is relatively flat with elevations between 1033 and 1045 m (3,390 to 3,430 ft)

above sea level. The site slopes to the southwest, is undeveloped, and is used for domestic

livestock grazing.

1.3.3.1.2 Nearby Highways

Information concerning public roads is provided in Section 1.3.2.4 of the SAR (LES, 2005a) and

Section 3.2.1.2 of the ISA Summary (LES, 2005b). The New Mexico State Highway 234 passes

along the southern boundary of the proposed facility.

Based on review of the information provided on nearby highways, staff concludes that the data

used in the analysis are accurate and are from acceptable sources.

1.3.3.1.3 Nearby Gas Pipelines

Information concerning gas pipelines passing through or located near the proposed facility site

is provided in Sections 1.3.2.4 of the SAR (LES, 2005a) and 3.2.2.4 of the ISA Summary (LES,

2005b).

Natural Gas Pipeline

The applicant identified an underground natural gas pipeline located along the south property
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line running parallel to New Mexico State Highway 234. A parallel gas pipeline is also identified,

but is not in use. 

Carbon Dioxide Pipeline

An underground carbon dioxide pipeline currently runs across the property. The ISA Summary

(LES, 2005b) indicated this pipeline will be relocated along the western and southern boundary

of Section 32 so the pipeline will be positioned at least 396.2 m (1300 ft) from the facility

restricted area and is approximately 945m (3100 ft) [estimated from Figure 3.2-3 of the ISA

Summary (LES, 2005b)] from a Separations Building Module, a safety-significant structure that

houses two cascade halls. The applicant concluded that, at this distance from the proposed

facility, the pipeline was not a safety concern. Staff agrees with the applicant’s assessment that

the carbon dioxide pipeline is not a safety concern to the Separations Building Module.

Onsite Natural Gas Pipeline

The proposed facility will include an on-site natural gas pipeline (Harper, 2003a). This pipeline

will be used to provide natural gas for heating the boiler in the Central Utility Building (CUB).

Summary

Staff reviewed information provided in the SAR (LES, 2005a) and ISA Summary (LES, 2005b)

on nearby highways, natural gas pipelines, and carbon dioxide pipelines and finds the data

used to be accurate and from reliable sources.

1.3.3.1.4 Nearby Air Transportation

Information relating to local air transportation is provided in Sections 1.3.2.4 of the SAR

(LES, 2005a), and Section 3.2.1.2.4 of the ISA Summary (LES, 2005b). The information

included the number of operations and holding patterns of six local airports. These

airports include:

• Lea County Regional Airport—40 km (25 statute miles) northwest of the proposed facility

site;

• Eunice Airport—24 km (15 statute miles) west of the proposed facility site;

• Lea County/Jal Airport—40 km (25 statute miles) south-southwest of the proposed

facility site;

• Andrews County Airport—48 km (30 statute miles) east of the proposed facility;

• Gaines County Airport—48 km (30 statute miles) northeast of the proposed facility; and

• Seminole Spraying Services (private)—48 km (30 statute miles) northeast of the

proposed facility.

The information about the number of operations and holding patterns of each airport was

obtained from the Federal Aviation Administration (FAA) (Yeung, 2003). Military flights are
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operated from the Lea County Regional Airport. The number of military operations is included in

the number of operations for the Lea County Regional Airport.

Based on information from the FAA (Yeung, 2003), the applicant concluded the holding patterns

for four of the six airports were, in general, more than the 3.2-km (2-statute miles) proximity

criterion (third criterion) provided in Section 3.5.1.6 of NUREG–0800 (NRC, 1981) for airway

distance from the site of interest. The applicant pointed out no specific holding patterns existed

for Eunice or Lea County/Jal Airport (LES, 2004a; LES, 2005a). For the Eunice Airport, the

annual operations are small, approximately 480 flights per year. This number is substantially

smaller than the threshold limit (225,000 annual operations) provided in NUREG–0800 (NRC,

1981). For the Lea County/Jal Airport, the applicant indicated the airport is more than 32 km (20

mi) away from the proposed facility, and the landing procedure usually will not be initiated until

an aircraft is within 32 km (20 mi) of the airport (LES, 2004a; LES,

2005a). Therefore, even if an aircraft is placed in a holding pattern, it will not bring the aircraft

near the proposed facility.

The applicant identified a low-level Federal airway passing within 9 km (-6 statute miles)

northeast of the proposed facility. This airway was analyzed in the ‘ISA Summary” and shown

to pose no hazard to the proposed facility.

Three military training routes were identified in the region. The closest route to the proposed

facility was approximately 26 km (16 statute miles) southwest of the site.

Based on the review of aircraft-crash risk assessment, staff concluded that the aircraft

transportation information used in the analysis is accurate and was obtained from a reliable

source.

1.3.3.2 Demographics

Information about demographics is provided in Section 1.3.2 of the SAR (LES, 2005a).

1.3.3.2.1 Local Population and Land Use

The proposed site is in Lea County, New Mexico, and is about 1.6 km (1 mi) from the New

Mexico-Texas border. Andrews County, Texas abuts the border on the Texas side. Together,

the counties have a combined population of 68,515, based on the 2000 census. In 1990, the

combined population was 70,130. This decrease is counter to the trends within New Mexico

and Texas, which had state-wide population increases of 20.1 percent and 22.8 percent,

respectively, over that 10-year period. The population decreases in Lea and Andrews Counties

are caused by decrease in petroleum industry jobs since the mid-1980s. It is expected that

population growth in these two counties in the next 30 years will be at a lower rate than the

overall rates in New Mexico and Texas.

Lea County covers 11,378 km2

 (4393 mi2), which is about three times the size of the State of

Rhode Island. Andrews County covers 3,895 km2

 (1504 mi2).

Major population centers near the proposed site include the following:
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Eunice, New Mexico, about 8 km (5 mi) west of the site;

Hobbs, New Mexico, about 32 km (20 mi) north of the site;

Jal, New Mexico, about 37 km (23 mi) south of the site;

Lovington, New Mexico, about 64 km (39 mi) north-northwest of the site;

Andrews, Texas, about 51 km (32 mi) east of the site;

Seminole, Texas (in Gaines County) about 51 km (32 mi) east-northeast of the site; and

Denver City, Texas (in Gaines County) 65 km (40 mi) north-northeast of the site.

Outside of these population centers, population density is very low. The nearest residences are

located about 4.3 km (2.6 mi) west of the proposed facility site.

Within 8 km (5 mi) of the site, land is primarily open land used for cattle grazing. Oil and gas

potentials are absent within this range, although operations are widespread beyond this area. 

Nearby industrial activities include a quarry and a “produced-water” reclamation company. Lea

County operates a county landfill on the south side of New Mexico Highway 234, and about 1.6

km (1 mi) east of the proposed site, Waste Control Specialists operates a hazardous chemical

waste disposal facility and has licenses for the treatment and storage of low-level radioactive

and mixed wastes. A natural gas processing plant is located about 6 km (4 mi) from the site.

1.3.3.2.2 Local Public Services

Fire fighting services are provided locally by Eunice Fire and Rescue, which is located 8 km (5

mi) from the proposed site. It is staffed by a full-time fire chief and 34 volunteer firefighters.

Equipment includes three pumpers, one tanker, and three grass trucks. Eunice Fire and

Rescue also has agreements for mutual assistance with all Lea County fire departments.

Police and law enforcement services are provided by the Eunice Police Department, which has

five full-time officers. The Lea County Sheriff’s Department also has a substation in Eunice. 

Agreements between Lea and Andrews Counties provide mutual support when needed. The

New Mexico State Police can also provide support.

Educational institutions in Eunice include an elementary school, a middle school, a high school,

and a private K-12 school. The nearest other schools are in Hobbs, New Mexico, 32 km (20 mi)

north of Eunice. The nearest schools in Andrews County, Texas, are in Andrews, Texas, about

51 km (32 mi) from the proposed site.

There are two hospitals in Lea County – one is located in Hobbs, New Mexico, 32 km (20 mi)

north of the proposed site, and the other in Lovington, New Mexico, 64 km (39 mi) northnorthwest of the site. The hospital in Hobbs is a 250-bed facility capable of handling acute and

stable chronic care patients. The hospital in Lovington is a full-service, 27-bed facility. The

Eunice clinic is the nearest medical center to the proposed facility site. The nearest nursing

home facilities are in Hobbs, New Mexico.

There are no recreation facilities near the site. The Eunice Golf Course is located

approximately 15 km (9.2 mi) west of the site. A historical marker and picnic area are located

about 3.2 km (2 mi) west of the proposed site at the intersection of New Mexico Highways 234

and 18.

1.3.3.2.3 Water Use
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Southeast New Mexico has a semi-arid climate with an average annual precipitation of 33 to 38

cm (13 to 15 in.). The proposed site has no surface water and/or drainage features. Essentially

all precipitation either infiltrates the soil or is evapotranspirated. There are no significant bodies

of water or navigable waterways in the vicinity of the proposed site. There is also no agricultural

activity in the site vicinity although there are various crops grown in Lea and Andrews Counties. 

Cattle grazing does occur at the proposed site and in the nearby vicinity. Dairy farming is

important in Lea County, although none takes place near the site and/or in Andrews County.

Known sources of water near the site include: a man-made pond at the quarry, adjacent to the

proposed site, stocked with fish for private use; Baker Spring, which is an intermittent surfacewater feature, located about 1.6 km (1 mi) northeast of the proposed site; and several cattle

watering holes, where groundwater is pumped by windmills and stored in above-ground tanks.

1.3.3.2.4 Summary

The staff reviewed the site demographic information presented by the applicant and finds that

the applicant has adequately described and summarized general site demographical

information related to local population, identification of population centers, schools, commercial

facilities, land use, and water use. Population information is provided based on the latest

census information.

1.3.3.3 Meteorology

1.3.3.3.1 Tornado Hazard and Tornado-Generated Missiles

Information about the tornadoes and design-basis tornado at the proposed facility is provided in

Sections 1.3.3.3 of the SAR (LES, 2005a), and Section 3.2.3.4.1 of the ISA Summary (LES,

2005b).

There is an average of nine tornadoes a year in New Mexico, and the occurrence of tornadoes

in the vicinity of the proposed facility is rare. Tornadoes are classified using the Fujita Tornado

Damage Scale (F-scale) with classifications ranging from F0-F5 (NOAA, 2005). Eighty-seven

tornadoes of low magnitude (F0 to F2) were reported in Lea County, New Mexico, between

January 1, 1950, and December 31, 2004. Only one additional tornado was reported as F3 on

May 17, 1954. Two tornadoes, one in 1998 and the second in 1999, had a magnitude of F0 and

were located near Eunice. All the reported tornadoes were associated with very light damage

(NCDC, 2005).

The tornado-generated missiles the applicant considered for the proposed facility included three

classes of missiles. These missiles were: (i) a 6.8-kg (15-lb), 10.2- × 30.5-cm (2- × 4-in.) timber

plank; (ii) 34-kg (75-lb), 7.6-cm (3-in.)-diameter steel pipe; and (iii) 1361-kg (3000-lb)

automobile. The associated vertical and horizontal impact velocities for each missiles also were

provided in the ISA Summary (LES, 2005b). According to the applicant (Harper, 2003b), the

tornado-generated missiles were determined based on DOE–STD–1020–2002 (DOE, 2002).

Based on the review of the information concerning tornados and tornado-generated missiles,

NRC concludes: (i) the information is accurate and is from reliable sources; and (ii) the designbases tornado-generated missiles are acceptable because they were determined based on an
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appropriate DOE standard. The use of a DOE standard is an acceptable approach to NRC

staff.

1.3.3.3.2 High Winds and Hurricanes

Information about high winds at the proposed facility is provided in Sections 1.3.3.1 of the SAR

(LES, 2005a); Section 3.2.3.4.2 of the ISA Summary (LES, 2005b), and Section 3.6.1.4 of the

Environmental Report (ER) (LES, 2005c). 

According to the SAR (LES, 2005a), no meteorological data were available for the proposed

facility site. Although the measured wind data at Midland–Odessa, Texas, and Roswell, New

Mexico, were discussed in the SAR (LES, 2005a) and ER (LES, 2005c), the Midland–Odessa

annual extreme wind data were used exclusively to estimate the high-wind hazard at the

proposed facility site (LES, 2005b; Harper, 2003b). The annual extreme data used range from

1973 through 1999. The wind speeds were 3-second gust speeds measured at 10 m (32.8 ft)

above ground. The Midland–Odessa weather station is located at the regional airport

approximately 103 km (64 mi) east-southeast of the proposed site, whereas the Roswell station

is approximately 161 km (100 mi) northwest of the proposed site. The climate data for both

locations were collected by the National Oceanic and Atmospheric Administration (NOAA) (LES,

2005c).

The largest wind speed for the annual extreme straight-line winds from 1973 through 1999 at

Midland–Odessa was 140 km/h (87 mph) and the smallest annual extreme straight-line wind

speed was 84 km/h (52 mph) (Harper, 2003b). The mean and standard deviation wind speeds

were 111.5 and 16.6 km/h (69.3 and 10.3 mph).

The high-wind hazard assessment was performed by fitting the annual extreme wind data using

the Fisher–Tippett Type I distribution model. The applicant chose the speed of a wind with an

annual probability of 1.0 × 10!5

 for the design-basis straight-line wind speed for the proposed

facility. This design-basis straight-line wind speed was 252 km/h (157 mph) (LES, 2005b).

Because the proposed facility is not located near the coastal area [805 km (500 mi) from the

coast], hurricanes affecting the coastal area will have no effect on the performance of the

proposed facility. Consequently, consideration of hurricane hazards on the design of the

proposed facility is not needed.

Based on the review of the information concerning high winds, the staff concludes that highwind hazards and the associated design-basis straight-line winds have been addressed

acceptably because the data used for assessment were from a recognized source and the

method used for assessing high-wind hazards is a commonly used and accepted method.

1.3.3.3.3 Temperature Extremes

Information about the temperature at the proposed facility site is provided in Section 3.6.1.2 of

the ER (LES, 2005c).

The regional temperatures in Hobbs, New Mexico [32 km (20 mi) north of the proposed facility

site]; Midland–Odessa, Texas; and Roswell, New Mexico, are discussed in the ER. The

discussions are based on 30-year records (from 1971 through 2000). As indicated previously,
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NOAA collected the climate data for Midland–Odessa and Roswell. However, the Western

Regional Climate Center collected the climate data for Hobbs (LES, 2005c).

The highest recorded monthly mean maximum temperature was 38.9 °C (102.1 °F), and the

lowest recorded monthly mean minimum temperature was !5.1 °C (22.8 °F) for Hobbs,

New Mexico. No such data were presented in the ER for Midland–Odessa or Roswell. The

highest daily maximum and lowest daily minimum temperatures were 46.7 °C (116.0 °F) and

!23.9 °C (!11.0 °F) for Midland–Odessa, and 45.6 °C (114.0 °F) and !22.8 °C (!9.0 °F) for

Roswell. No such data were presented for Hobbs. As indicated, the highest daily maximum

and the lowest daily minimum temperatures for Midland–Odessa and Roswell were similar. 

The staff reviewed the temperature information and find the information acceptable because

recognized data sources were used and the temperature extremes are properly determined.

1.3.3.3.4 Extreme Precipitation

Section 1.3.3.2 of the SAR (LES, 2005a); and Sections 3.2.3.2 and 3.2.3.4.4 of the ISA

Summary (LES, 2005b) discuss the rainfall precipitation at the proposed facility site. The

precipitation data for Hobbs, Midland–Odessa, and Roswell were listed in Tables 3.2-14 through

3.2-16 of the ISA Summary (LES, 2005b). These data were collected from the Western

Regional Climate Center and NOAA and are based on data from 1971 through 2000

(LES, 2005c). The maximum monthly totals were 35.13 cm (13.83 in.) for Hobbs; 24.6 cm (9.7

in.) for Midland–Odessa; and 17.5 cm (6.88 in.) for Roswell. The minimum monthly totals were

zero for all locations. The highest 24-hour precipitation was 15.2 cm (5.99 in.) for

Midland–Odessa and 12.5 cm (4.91 in.) for Roswell. 

According to the SAR (LES, 2005a), the local intense probable maximum precipitation was

estimated from NOAA data (NOAA, 1982). The local intense probable maximum precipitation

was approximately 43.9 cm (17.3 in.) in 1 hour, over a 2.6-km2

 (1-mi2) area. 

The staff reviewed the information concerning regional precipitation and local intense probable

maximum precipitation presented in the SAR (LES, 2005a) and the ISA Summary (LES, 2005b)

and find the information acceptable because recognized data sources, such as NOAA, were

used. 

1.3.3.3.5 Snow

Section 1.3.3.2 of the SAR (LES, 2005a), and Section 3.2.3.3 of the ISA Summary (LES, 2005b)

discuss the regional snowfall. NOAA collected the snowfall data. The maximum monthly

snowfall/ice pellets were 24.9 cm (9.8 in.) for Midland–Odessa and 53.3 cm (21.0 in.) for

Roswell. The maximum snowfall/ice pellets during a 24-hour period were 12.47 cm (4.91 in.) for

Midland–Odessa and 41.91 cm (16.5 in.) for Roswell. No snowfall information was available for

Hobbs, New Mexico.

The staff reviewed the information concerning snow precipitation presented in the SAR (LES,

2005a) and the ISA Summary (LES, 2005b) and find the information acceptable because

recognized data sources, such as NOAA, were used.
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1.3.3.3.6 Lightning and Thunderstorms

Section 1.3.3.3 of the SAR (LES, 2005a), and Section 3.2.3.4.5 of the ISA Summary

(LES, 2005b) describe the potential of thunderstorms and lightning strikes at the proposed

facility site. The applicant indicated thunderstorms occur every month and are most common in

spring and summer at the proposed facility site.

The applicant estimated the lightning strike frequency at the proposed facility site to be 1.36

flashes per year. The applicant also stated in the ISA Summary (LES, 2005b) that the proposed

facility will be designed for lightning protection.

The staff reviewed the information about lightning and find the lightning strike frequency

determined for the site is acceptable and appropriate. Staff further concludes the design

approach proposed by the applicant to protect the proposed facility from lightning effects

is acceptable.

1.3.3.3.7 Sandstorms

Section 1.3.3.3 of the SAR (LES, 2005a) describes the potential of sandstorms at the proposed

facility site.

According to the SAR (LES, 2005a), blowing sand and dust may occur occasionally. Large dust

storms with the potential of covering a large region are rare. Staff reviewed the information

about sandstorms presented in the SAR (LES, 2005a) and find the information sufficient and

acceptable.

1.3.3.4 Geology

1.3.3.4.1 Seismic Hazard

Seismic hazards are discussed in Section 1.3.5 of the SAR (LES, 2005a); and Sections 3.2.5

and 3.2.6 of the ISA Summary (LES, 2005b).

The following areas concerning the seismic hazard applicable to the safety analysis and design

of the proposed facility were reviewed:

• Seismic source characterization; 

• Ground motion attenuation;

• Seismic hazard calculation;

• Development of site-specific spectra; and

• Surface faulting.

1.3.3.4.1.1 Seismic Source Characterization

Geological and Tectonic Settings
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Section 1.3.5 of the SAR (LES, 2005a) and Section 3.2.5 of the ISA Summary (LES, 2005b)

provide a description of the regional and local geological and tectonic settings. The proposed

facility site is located within the Central Basin Platform area. The Central Basin Platform Area is

situated between the Midland and Delaware Basins, all of which are part of the Permian Basin,

a 250-million-year-old structure. The Permian Basin is a downward flexure of a large thickness

of originally flat-lying bedded, sedimentary rock. The base of the Permian Basin sediments

extends to approximately 1525 m (5000 ft) beneath the proposed facility site. The top of the

Permian section is approximately 434 m (1425 ft) below ground surface. These sediments are

overlain by sedimentary strata of the Triassic Age Dockum Group. The upper formation of the

Dockum Group is the Chinle Formation, locally overlain by either the Tertiary Ogallala, Gatuña,

or Antlers Formations, or Quaternary alluvium. At the proposed facility site, geotechnical

borings identified up to 0.6 m (2 ft) of loose eolian sand underlain by dense to very dense, fineto medium-grained sand and silty sand of the Gatuña Formation. The sands of the Gatuña

Formation are locally cemented with caliche. Beneath the Gatuña Formation, the Chinle

claystone, a hard and highly plastic clay, was encountered in geotechnical borings at depths

from 10.7 to 12.2 m (35 to 40 ft).

As noted in the ISA Summary (LES, 2005b), the Southeast New Mexico–West Texas area is

presently structurally stable. The Laramide Orogeny (late Cretaceous to Early Tertiary time)

uplifted the region to its present elevation, and there has been no substantial tectonic activity

since this early Tertiary deformation. The Permian Basin has subsided slightly since the

Laramide Orogeny. However, this subsidence is believed to be a result of dissolution of the

Permian evaporite layers by groundwater or possibly compaction from oil and gas extraction. 

As stated in the ISA Summary (LES, 2005b), no active faults have been identified at the site. 

Faulting consists of geologically older subsurface faults in the Permian Basin subregion

related to the development of the Permian Basin and the Laramide Orogeny. The nearest

evidence of Quaternary faulting is 161 km (100 mi) west of the site, in the Basin and Range

tectonic province.

Historical Seismicity

Section 1.3.5.2 of the SAR (LES, 2005a) and Section 3.2.6.1 of the ISA Summary (LES, 2005b)

summarizes the historical seismicity at the proposed facility site. As stated in the ISA Summary

(LES, 2005b), the assessment of historical seismicity included earthquakes in the region of

interest known from felt or damage records and from more recent instrumental records (since

the early 1960s). The largest earthquake known to occur within 322 km (200 mi) of the site was

the August 16, 1931, earthquake near Valentine, Texas. This earthquake had an estimated ML

(Local Magnitude) of 6.0 to 6.4 and produced a maximum epicentral intensity of VIII on the

Modified Mercalli intensity scale. This earthquake occurred approximately 237 km (147 mi) from

the proposed site location. Within 80 km (50 mi) of the site, the largest historical earthquake

was a ML 5.0 event in 1992, approximately 16 km (10 mi) southwest of the site. Other

significant events between 322 km (200 mi) and 80 km (50 mi) of the proposed facility site

ranged in ML from 4.0 to 5.7. 

Earthquakes in the region of the proposed facility site include isolated and small clusters of lowto-moderate-magnitude events toward the Rio Grande Valley of New Mexico and in Texas,

southeast of the proposed site. According to the ISA Summary (LES, 2005b), no earthquakes

in the site region are known to be correlated to specific faults. An earthquake catalog based on
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the historic seismicity in the region [322-km (200-mi) radius] was presented in the ISA Summary

(LES, 2005b). This catalog was composed of data from: the Advanced National Seismic

System (NCEDC, 2004); University of Texas Institute for Geophysics (UTIG, 2002); New Mexico

Tech Historical Catalog (NMIMT, 2003); and New Mexico Tech Regional catalogs. The catalog

identified a substantial cluster of seismic activity that has occurred on and near the Central

Basin Platform since the mid-1960s. It was suggested by DOE (DOE, 2003) and noted in the

ISA Summary (LES, 2005b) that Central Basin Platform earthquakes are not tectonic in origin

but instead are related to water injection and withdrawal resulting from secondary recovery

operations in oil fields in the Central Basin Platform area. The ISA Summary (LES, 2005b)

noted, however, the January 2, 1992, event was attributed to a tectonic origin because of its

determined focal depth of approximately 12 km (7 mi) and is not correlated with oil or gas

drilling. At the proposed facility site, postulated earthquakes that could impact safe operation of

the proposed facility are associated with zones of crustal weakness in the Central Basin

Platform and the Basin and Range tectonic province.

The staff concludes that information concerning seismic source characterization presented in

Section 1.3.5 of the SAR (LES, 2005a) and the ISA Summary (LES, 2005b) is acceptable. The

information provides a complete summary of seismicity and potential fault and tectonic sources

and thereby demonstrates compliance with regulatory requirements in 10 CFR 30.33, 10 CFR

40.32, 10 CFR 70.22, and 10 CFR 70.65(b)(1). 

1.3.3.4.1.2 Ground Motion Attenuation

Details of ground motion attenuation functions used to compute the hazard are described in

Section 3.2.6.4.1 of the ISA Summary (LES, 2005b). Several attenuation models were used in

the ISA Summary. The Nuttli attenuation model developed by the U.S. Department of the Army,

Waterways Experiment Station (USDA, 1973) was primarily selected because it was used in the

DOE (DOE, 2003) seismic hazard assessment. The Toro, et al. (Toro, 1997) attenuation model

also was used in the hazard calculations for comparison.

The attenuation models used in the ISA Summary were applicable to locations within the

Central U.S.. The proposed facility site is located at 103° west longitude, slightly east of the

105° west longitude cutoff for Central and Eastern U.S. sites, as specified in Regulatory Guide

1.165 (NRC, 1997). In addition, Frankel, et al. (Frankel, 1996) specified attenuation zones for

the U.S. in its hazard mapping project. The U.S. Geological Survey (USGS) boundary

separating the Western U.S. and the Central and Eastern U.S. attenuation zones also is located

at approximately 105° west longitude and slightly to the west of the proposed facility site. The

proposed facility site is thus situated within the area in which both the Central and Eastern U.S.

attenuation models are applicable.

1.3.3.4.1.3 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 (LES, 2005b), the nearest 

recent faulting is located more than 161 km (100 mi) west of the site. Therefore, surface

faulting was not considered a credible disruptive event for the proposed facility.

Recently, a fault was discovered at the nearby Waste Control Specialists (WCS) site. However,
1-22

subsequent fault investigations revealed that the faulting is inactive because no faults exist in

formations younger than Triassic age (205 to 240 million years old) (LES, 2004a).

1.3.3.4.2 Slope Stability

Section 1.3.1.2 of the SAR (LES, 2005a), and Section 3.2.1.1 of the ISA Summary (LES, 2005b)

describe the topography at the proposed facility site. The SAR (LES, 2005a) and ISA Summary

(LES, 2005b) indicated the site topography is relatively flat, with a gradual elevation increase

from southwest to northeast. The staff site visit on May 27–28, 2004 (NRC, 2004), confirmed

the area at the proposed facility is relatively flat. Consequently, slope stability is not a safety

concern for this proposed facility.

1.3.3.4.3 Liquefaction

Liquefaction potential of soils beneath the proposed facility is discussed in Section 3.2.7.1 of the

ISA Summary (LES, 2005b). According to the ISA Summary (LES, 2005b), except for a top

layer of loose sand [up to 0.6m (2ft)], the soils at the proposed facility site are dense to very

dense and the groundwater level is at least 30 m (98 ft) below ground surface. In Section

3.3.2.1 of the ER, the applicant (LES, 2005c) indicates the groundwater table at the site is

65 to 68 m (214 to 222 ft) below ground surface. Consequently, the applicant concluded the

potential for liquefaction was remote (LES, 2005b, c).

Geotechnical investigation indicated the soil beneath the proposed facility site is a layer of loose

eolian sand underlain by the Gatuña Formation (dense to very dense sand and silty sand). 

Below the Gatuña Formation is the Chinle claystone, a very hard, highly plastic clay. The

Chinle claystone was encountered at depths approximately 10.7 to 12.2 m (35 to 40 ft). For the

top 7.6 m (25 ft) of soils, the blow-count values ranged from 20 to 76. Beneath the 7.6-m (25-ft)

horizon, typical blow-count values were more than 60, with even larger blow-count values for

the Chinle claystone.

The staff reviewed the geotechnical investigation information presented in the ISA Summary

(LES, 2005b, c) and concurs with the applicant that the potential for liquefaction of soils at the

site may not be a safety concern for the proposed facility. The applicant committed in Section

3.3.9 of the SAR (LES, 2005a) to perform additional geotechnical investigations at the site to

confirm that liquefaction is not a safety concern for the proposed facility. Additional site testing

will be evaluated in accordance with NRC Regulatory Guide 1.198, “Procedures and Criteria for

Assessing Seismic Soil Liquefaction at Nuclear Power Plant Sites” (NRC, 2003).

1.3.3.4.4 Settlement

Settlement of foundations for the proposed facility is discussed in Section 3.2.7 of the ISA

Summary (LES, 2005b). In its ISA Summary, the applicant stated that only five borings were

drilled at the proposed facility site to determine the suitability of the site. The applicant

recognized the geotechnical results obtained from the five borings were not sufficient for final

design purposes. The applicant committed in Section 3.3.9 of the SAR (LES, 2005a) that the

settlement and differential settlement for the design of the proposed facility will be determined

based on the information that will be obtained from the additional geotechnical investigations. 

Allowable soil bearing pressures will be evaluated in accordance with Naval Facilities

Engineering Command Design Manual NAVFAC DM-7.02, “Foundations and Earth Structures”
1-23

(NAVFAC, 1996). Building settlement analyses will be performed in accordance with Naval

Facilities Engineering Command Design Manual NAVFAC DM-7.01, “Soil Mechanics”

(NAVFAC, 1986) and Winterkorn and Fang, “Foundation Engineering Handbook” (Winterkorn,

1975).

The staff reviewed the information presented concerning differential settlements and find the

applicant’s commitment to perform additional geotechnical investigations using acceptable

geotechnical standards for final facility design to be acceptable.

1.3.3.5 Hydrology

Site surface water and groundwater hydrology is discussed in Sections 1.3.4 of the SAR (LES,

2005a) and 3.2.4 of the ISA Summary (LES, 2005b). The applicant obtained hydrological data

principally from previous investigations conducted by WCS, which is located 1.6 km (1 mi) east

of the proposed site. WCS operates a hazardous chemical treatment and disposal facility. The

applicant performed a limited number of geotechnical studies that demonstrate that the WCS

data are applicable to the proposed site.

The proposed site contains no surface water and/or surface water drainage features, with

essentially all precipitation subject to either infiltration or evapotranspiration.

The applicant performed subsurface studies of the alluvial material that overlies the Chinle red

bed clays. These alluvial deposits are 9 to 18 m (30 to 60 ft) thick. The Chinle formation

consists of a low-permeability clay unit having a thickness of 323 to 333 m (1060 to 1092 ft) and

is the upper formation within the Triassic Age Dockum Group. No perched water systems in

the alluvial deposits were found, although one well produced water samples, because of a

limited groundwater occurrence.

The low permeability Chinle formation essentially isolates the deep and shallow groundwater

systems. Within the Chinle formation are two distinct groundwater systems, with no

interconnections. The first is a siltstone or silty sandstone unit with some saturation at 65 to 68

m (214 to 222 ft) below the surface. This unit is a low-permeability formation that does not yield

groundwater easily. The second unit is a saturated siltstone layer approximately 30.5 m (100 ft)

thick, at an elevation of 183 m (600 ft) below the surface. The Santa Rosa formation lies below

the Chinle Unit, but within the Dockum Group, at 340 m (1115 ft) below the surface. The Santa

Rosa unit is the first occurrence of a well-defined aquifer system. However, this system is

considered non-potable because of high concentrations of dissolved solids.

At the quarry site, north of the proposed site, there are shallow groundwater occurrences. 

These shallow perched systems, however, are intermittent and limited and caused by a layer

caliche or caprock at the surface that in places is fractured and can lead to rapid infiltration of

precipitation forming the perched water system. Caprock, however, is not present at the

proposed site, and, therefore, it is not expected that significant perched water systems would be

produced.

Baker Spring is located about 1.6 km (1 mi) northeast of the proposed site. However, this

spring is intermittent and flows only after precipitation events.

Several localized shallow perched groundwater systems exist to the east of the proposed site
1-24

and are used to supply water pumped by windmills to tanks for grazing livestock. These

perched systems are located above the Chinle clays, but the volume of water produced is

limited.

Because of the lack of sufficient surface and groundwater supplies, the applicant will not make

withdrawals of groundwater at the site. Instead, the applicant is proposing to obtain water for

plant use using Eunice and Hobbs municipal supplies. These water supplies are obtained from

well fields near Hobbs, New Mexico. The applicant is also not proposing to inject water into

groundwater systems at the site.

Since there are no surface water bodies in the immediate vicinity of the site, flooding is not a

design objective. The only potential flooding at the plant would occur from intense local

precipitation events. Flood protection is provided by establishing building floor levels above the

calculated depth of ponded water caused by intense precipitation events (see Sections 1.3.3.3.4

and 1.3.3.3.5 of this SER).

The staff reviewed the applicant’s hydrological data and finds that it provides sufficient

information to assess site flooding hazards and ground- and surface water impacts, and is

consistent with information in the ISA Summary.

1.3.4 EVALUATION FINDINGS

The staff has reviewed the site description for the proposed LES uranium enrichment facility

according to Section 1.3 of the Standard Review Plan. The applicant has adequately described

and summarized general information pertaining to: (1) the site geography, including its location

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

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

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

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

for the site; and (4) applicable 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; and that it demonstrates compliance with regulatory

requirements in 10 CFR 30.33, 10 CFR 40.32, 10 CFR 70.22, and 10 CFR 70.65(b)(1).

1.4 REFERENCES

(ASTM, 2003) American Society for Testing and Materials (ASTM). ASTM C787, “Standard

Specification for Uranium Hexafluoride for Enrichment,” 2003.

(DOE, 2002) U.S. Department of Energy (DOE). Interagency Agreement Between the U.S.

Department of Energy Oak Ridge Operations and the U.S. Nuclear Regulatory Commission

Concerning Foreign Ownership, Control, and Influence Determinations Associated with Certain

NRC Licensees. May 6, 2002.

(DOE, 2003) U.S. Department of Energy (DOE). DOE/WIPPS95S2065, Rev. 7, “Waste

Isolation Pilot Plant Contact-Handled (CH) Waste Safety Analysis Report,” 2003.

(DOE, 2004) U.S. Department of Energy (DOE). DOE–STD–1024, “Natural Phenomena
1-25

Hazards Design and Evaluation Criteria for Department of Energy Facilities,” 2004.

(DOE, 2005) U.S. Department of Energy (DOE). Letter to U.S. Nuclear Regulatory

Commission on Louisiana Energy Services Foreign Ownership, Control, or Influence issues, 

March 31, 2005.

(Frankel, 1996) Frankel, A., et al. “National Seismic Hazards Maps, June 1995

Documentation.” U.S. Geological Survey. 1996.

(Harper, 2003a) Harper, G.A. “Assessment of Other External Event Hazards at NEF for ISA

and Design Basis.” Document Identifier 51-2400552-00. Framatome ANP, Inc. 2003.

(Harper, 2003b) Harper, G.A. “Assessment of Tornado, Tornado Missiles and High Wind

Loads at NEF for ISA and Design Basis—Attachment 2.” Document Identifier 51-2400548-00. 

Framatome ANP, Inc. 2003.

(LES, 2004a) Louisiana Energy Services (LES), letter to U.S. Nuclear Regulatory Commission,

“Clarifying Information Related to External Hazards Analysis,” August 31, 2004.

(LES, 2004b) Louisiana Energy Services (LES), letter to U.S. Nuclear Regulatory Commission,

“Clarifying Information Related to Probabilistic Seismic Hazard Analysis,” November 22, 2004.

(LES, 2005a) Louisiana Energy Services (LES). “National Enrichment Facility Safety Analysis

Report,” Revision 6, 2005.

(LES, 2005b) Louisiana Energy Services (LES). “Integrated Safety Analysis Summary,” 

Revision 4, 2005.

(LES, 2005c) Louisiana Energy Services (LES). “National Enrichment Facility Environmental

Report,” Revision 5, 2005.

(NCDC, 2005) National Climatic Data Center (NCDC). “Storm Events.”

<http://www4.ncdc.noaa.gov/cgi-win/wwcgi.dll?wwevent~storms> (Accessed 3/17/05).

(NAVFAC, 1986) U.S. Naval Facilities Engineering Command (NAVFAC). Design Manual

NAVFAC DM-7.01, “Soil Mechanics,” 1986.

(NAVFAC, 1996) U.S. Naval Facilities Engineering Command (NAVFAC). Design Manual

NAVFAC DM-7.02, “Foundations and Earth Structures,” 1996.

(NCEDC, 2004) Northern California Earthquake Data Center (NCEDC). “Composite

Earthquake Catalog,” 2004, <http://quake.geo.berkeley.edu/anss/>

(NMIMT, 2003) New Mexico Institute of Mining and Technology (NMIMT). “Earthquake

Catalogs for New Mexico and Bordering Areas: 1869–1998,” 2003.

(NOAA, 1982) National Oceanic and Atmospheric Administration (NOAA). “Application of

Probable Maximum Precipitation Estimates—United States East of the 105th Meridian,”

Hydrometeorological Report No. 52, 1982.
1-26

(NOAA, 2005) National Oceanic and Atomospheric Administration. “Tornadoes.”

<http://www.noaa.gov/tornadoes.html> (Accessed 3/17/05).

(NRC, 1981) U.S. Nuclear Regulatory Commission (NRC). NUREG–0800, “Standard Review

Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,” 1981.

(NRC, 1997) U.S. Nuclear Regulatory Commission (NRC). Regulatory Guide 1.165,

“Identification and Characterization of Seismic Sources and Determination of Safe Shutdown

Earthquake Ground Motion,” 1997.

(NRC, 1999) U.S. Nuclear Regulatory Commission (NRC). “Standard Practice Procedures

Plan Standard Format and Content for the Protection of Classified Matter for NRC Licensee,

Certificate Holder and Others Regulated by the Commission,” 1999.

(NRC, 2002) U.S. Nuclear Regulatory Commission (NRC). NUREG–1520, “Standard Review

Plan for the Review of a License Application for a Fuel Cycle Facility,” 2002.

(NRC, 2003) U.S. Nuclear Regulatory Commission (NRC). Regulatory Guide 1.198,

“Procedures and Criteria for Assessing Seismic Soil Liquefaction at Nuclear Power Plant Sites,”

2003.

(NRC, 2004) Graves, H.L., U.S. Nuclear Regulatory Commission (NRC). Summary of Meeting

with Louisiana Energy Services on Structural, Seismic, High Winds, and Tornado Safety

Analysis. June 29, 2004.

(Toro, 1997) Toro, G.R., N.A. Abrahamson, and J.H. Schneider, “Model of Strong Ground

Motions from Earthquakes in Central and Eastern North America: Best Estimates and

Uncertainties,” Seismological Research Letters, Vol. 68, No. 1, pp. 41S5, 1997.

(USDA, 1973) U.S. Department of Army, Waterways Experiment Station (USDA). “Design

Earthquake for the Central United States,” Miscellaneous Paper S–731–1, 1973.

(UTIG, 2002) The University of Texas Institute for Geophysics. “Compendium of Texas

Earthquakes,” 2002.

(Winterkorn, 1975) Winterkorn, H.F., and Fang, H.Y., “Foundation Engineering Handbook,”

1975.

(Yeung, 2003) Yeung, W.S., “Aircraft Hazard Risk Determination,” Document Identifier 32-

2400569-00, Framatome ANP, Inc., 2003.
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2.0 ORGANIZATION AND ADMINISTRATION

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

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

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

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

provided adequate qualification descriptions for key management positions.

2.1 REGULATORY REQUIREMENTS

10 CFR 30.33, 10 CFR 40.32, 10 CFR 70.22, 10 CFR 70.23, and 10 CFR 70.62(d) require a

management system and administrative procedures for the effective implementation of health,

safety, and environment (HS&E) protection functions concerning the applicant’s corporate

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

procedures to provide adequate safety for workers, the public, and the environment.

2.2 REGULATORY ACCEPTANCE CRITERIA

The acceptance criteria applicable to the U.S. Nuclear Regulatory Commission’s (NRC’s) review

of the organization and administration section of the application are contained in Section 2.4.3

of the “Standard Review Plan for Fuel Cycle Facilities,” NUREG-1520 (NRC, 2002). 

2.3 STAFF REVIEW AND ANALYSIS

In Section 2.1 of the applicant’s Safety Analysis Report (SAR) (LES, 2005), the applicant

provides a functional description of specific organization groups responsible for managing the

design, construction, and operation of the facility. Included in this section are the plans for the

transition from the start-up phase to operations.

In Section 2.2 of the applicant’s SAR (LES, 2005), the applicant describes the qualifications,

responsibilities, and authorities for key supervisory and management personnel, along with a

listing of the shift crew composition.

In Section 2.3 of the applicant’s SAR (LES, 2005), the applicant includes administration

procedures for effective implementation of HS&E functions, using written procedures and

reporting of unsafe conditions or activities, along with written agreements with offsite emergency

resources and a commitment to establish formal management measures to ensure availability of

Items Relied on for Safety (IROFS).

Figure 2.1-1 of the applicant’s SAR (LES, 2005) shows the LES organization during design and

construction phase of the facility. Before beginning operations, this organization transitions to

the one shown in Figure 2.1-2.
2-2

2.3.1 Organizational Responsibilities and Qualifications

The Chief Operating Officer is appointed by the President and is responsible for ensuring that

the facility complies with all applicable regulatory requirements. The Chief Operating Officer

directs these responsibilities through the Plant Manager. 

The Plant Manager will be appointed by, and reports to, the LES Chief Operating Officer. The

plant manager has direct responsibility for operation of the facility in a safe, reliable, and

efficient manner. He/she is responsible for proper selection of staff for all key positions,

including positions on the Safety Review Committee (SRC). The Plant Manager is responsible

for the protection of the facility staff and the general public from radiation and chemical

exposure or any other consequences of an accident at the facility and also bears the

responsibility for compliance with the facility license. He/she or designee(s) has the authority to

approve and issue procedures. The Plant Manager will have, as a minimum, a bachelor's

degree (or equivalent) in an engineering or scientific field and 10 years of responsible nuclear

experience.

The Quality Assurance (QA) Director is appointed by, and reports to, the President, and has

overall responsibility for development, management, and implementation of the LES QA

Program. He/she will have, as a minimum, a bachelors degree (or equivalent) in an engineering

or scientific field, and at least 6 years of responsible nuclear experience in the implementation

of a QA program. The QA Director will have at least 4 years experience in a QA organization at

a nuclear facility.

The QA Manager reports to the Plant Manager and is responsible for establishing and

maintaining the QA Program for the facility. The facility line managers and their staff who are

responsible for performing quality-affecting work are responsible for ensuring implementation of

and compliance with the QA Program. The QA Manager position is independent from other

management positions at the facility, to ensure that the QA Manager has access to the Plant

Manager for matters affecting quality. In addition, the QA Manager has the authority and

responsibility to contact the LES President through the QA Director, with any QA concerns. The

QA Manager will have, as a minimum, a bachelor's degree (or equivalent), in an engineering or

scientific field, and at least 5 years of responsible nuclear experience in the implementation of a

QA program. The QA Manager shall have at least 2 years experience in a QA organization at a

nuclear facility.

The HS&E Manager reports to the Plant Manager and has the responsibility for assuring safety

at the facility through activities including maintaining compliance with safeguards, appropriate

rules, regulations, codes – and has the responsibility for implementation and control of the

Fundamental Nuclear Material Control Plan (FNMCP). This includes HS&E activities associated

with nuclear criticality safety, radiation protection, chemical safety, environmental protection,

emergency preparedness, and industrial safety. The HS&E Manager works with the other

facility managers to ensure consistent interpretations of HS&E requirements, performs

independent reviews, and supports facility and operations change control reviews. This position

is independent from other management positions at the facility to ensure objective HS&E audit,

review, and control activities. The HS&E Manager has the authority to shut down operations if

they appear to be unsafe, and must consult with the Plant Manager with respect to restart of

shutdown operations after the deficiency, or unsatisfactory condition, has been resolved. 

Changes to the facility or to activities of personnel that require prior NRC approval are reviewed
2-3

and approved by the HS&E Manager or designee. The HS&E Manager will have, as a

minimum, a bachelor's degree (or equivalent) in an engineering or scientific field and at least 5

years of responsible nuclear experience in HS&E or related disciplines. The HS&E Manager

will also have at least 1 year of direct experience in the administration of nuclear criticality safety

evaluations and analyses.

The Operations Manager reports to the Plant Manager and has the responsibility of directing the

day-to-day operation of the facility. This includes such activities as ensuring the correct and

safe operation of uranium hexafluoride (UF6) processes, proper handling of UF6, and the

identification and mitigation of any off-normal operating conditions. In case of the absence of

the Plant Manager, the Operations Manager may assume the responsibilities and authorities of

the Plant Manager. The Operations Manager will have, as a minimum, a bachelor's degree (or

equivalent) in an engineering or scientific field and 4 years of responsible nuclear experience.

The Uranium Management Manager reports to the Plant Manager and has the responsibility for

UF6 cylinder management (including transportation licensing) and directing the scheduling of

enrichment operations to ensure smooth production. This includes activities such as ensuring

that proper feed material and maintenance equipment are available for the facility. In case of

the absence of the Plant Manager, the Uranium Management Manager may assume the

responsibilities and authorities of the Plant Manager. The Uranium Management Manager will

have, as a minimum, a bachelor's degree (or equivalent) in an engineering or scientific field and

4 years of responsible nuclear experience.

The Technical Services Manager reports to the Plant Manager and has the responsibility of

providing technical support to the facility. This includes technical support for facility

modifications (including administration of the configuration management system); engineering

support for operations and maintenance; performance; operation of the chemistry laboratory;

maintenance activities; and computer support. In case of the absence of the Plant Manager,

the Technical Services Manager may assume the responsibilities and authorities of the Plant

Manager. The Technical Services Manager will have, as a minimum, a bachelor’s degree (or

equivalent) in an engineering or scientific field and 4 years of responsible nuclear experience.

The Human Resource Manager reports to the Plant Manager and has the responsibility for

community relations; ensuring adequate staffing; ensuring training is provided for facility

employees; providing administrative support services to the facility, including document control;

and for the physical security of the facility. The Human Resource Manager will have as a

minimum, a bachelor's degree in Personnel Management, Business Administration, or related

field, and 3 years of appropriate, responsible experience in implementing and supervising

human resource responsibilities at an industrial facility.

The QA Inspectors report to the QA Manager (via a designated supervisory position, if

applicable) and have the responsibility for performing inspections related to the implementation

of the LES QA Program. QA Inspectors performing QA Level 1 activities will be certified in

accordance with American Society of Mechanical Engineers (ASME) NQA-1 (ASME, 1994) and

ASME NQA-1a (ASME, 1995) inspector qualification requirements.

The QA Auditors report to the QA Manager (via a designated supervisory position, if applicable)

and have the responsibility for performing audits related to the implementation of the LES QA

Program. QA Auditors require certification under the LES QA Program. This certification
2-4

includes training on the LES QA Program, audit fundamentals, objectives and techniques for

performing audits, and on-the-job training.

The QA Technical Support personnel report to the QA Manager (via a designated supervisory

position, if applicable) and have the responsibility for providing technical support related to the

implementation of the LES QA Program. QA Technical Support staff receive QA Indoctrination

Training and training in the specific QA procedures needed to perform their jobs.

The Emergency Preparedness Manager reports to the HS&E Manager and has the

responsibility for ensuring that the facility remains prepared to react and respond to any

emergency situation that may arise. This includes emergency preparedness training of facility

personnel; facility support personnel; the training of, and coordination with, offsite emergency

response organizations; and conducting periodic drills to ensure facility-personnel and offsiteresponse-organization-personnel training is maintained up to date. The Emergency

Preparedness Manager will have a minimum of 5 years of experience in the implementation and

supervision of emergency plans and procedures at a nuclear facility. No credit for academic

training may be taken toward fulfilling this experience requirement.

The Licensing Manager reports to the HS&E Manager and has the responsibility for

coordinating facility activities to ensure that compliance is maintained with applicable NRC

requirements. The Licensing Manager is also responsible for ensuring abnormal events are

reported to NRC in accordance with NRC regulations. The Licensing Manager will have a

minimum of 5 years of appropriate, responsible experience in implementing and supervising a

nuclear licensing program.

The Environmental Compliance Manager reports to the HS&E Manager and has the

responsibility for coordinating facility activities to ensure all local, State, and Federal

environmental regulations are met. This includes submission of periodic reports to appropriate

regulating organizations of effluents from the facility. The Environmental Compliance Manager

will have a minimum of 5 years of appropriate, responsible experience in implementing and

supervising a nuclear environmental compliance program.

The Radiation Protection Manager reports to the HS&E Manager and has the responsibility for

implementing the Radiation Protection program. These duties include: the training of personnel

in use of equipment; control of radiation exposure of personnel; continuous determination of the

radiological status of the facility; and conducting the radiological environmental monitoring

program. During emergency conditions, the Radiation Protection Manager's duties may also

include:

• Providing Emergency Operations Center personnel information and recommendations

concerning chemical and radiation levels at the facility;

• Gathering and compiling onsite and offsite radiological and chemical monitoring data;

• Making recommendations concerning actions at the facility and offsite deemed

necessary for limiting exposures to facility personnel and members of the general public;

and

• Taking prime responsibility for decontamination activities.
2-5

In matters involving radiological protection, the Radiation Protection Manager has direct access

to the Plant Manager. The Radiation Protection Manager will have, as a minimum, a bachelor’s

degree (or equivalent), in an engineering or scientific field, and 3 years of responsible nuclear

experience associated with implementation of a Radiation Protection program. At least 2 years

of experience will be at a facility that processes uranium, including uranium in soluble form.

The Industrial Safety Manager reports to the HS&E Manager and has the responsibility for the

implementation of facility industrial safety programs and procedures. This will include programs

and procedures for training individuals in safety and maintaining the performance of the facility

fire protection systems. The Industrial Safety Manager will have, as a minimum, a bachelor's

degree (or equivalent) in either an engineering or a scientific field and 3 years of appropriate,

responsible nuclear experience associated with implementation of a facility safety program.

Criticality Safety Engineers report to the HS&E Manager (via a designated supervisory position,

if applicable) and are responsible for the preparation or review of nuclear criticality safety

evaluations and analyses, and conducting and reporting periodic nuclear criticality safety

assessments. Nuclear criticality safety evaluations and analyses require independent reviews

by a Criticality Safety Engineer. Criticality Safety Engineers shall have a minimum of 2 years

experience in the implementation of a criticality safety program. These individuals will hold a

bachelor’s degree in an engineering or scientific field and have successfully completed a

training program, applicable to the scope of operations, in the physics of criticality and in

associated safety practices.

Should a change to the facility require a nuclear criticality safety evaluation, an individual who,

as a minimum, possesses the equivalent qualifications of the Nuclear Criticality Engineer will

perform the evaluation or analysis. In addition, this individual will have at least 2 years of

experience performing criticality safety analyses and implementing criticality safety programs. 

An independent review of the evaluation or analysis will be performed by a qualified Criticality

Safety Engineer.

The Chemical Safety Engineer reports to the HS&E Manager (via a designated supervisory

position, if applicable) and is responsible for the preparation or review of chemical safety

programs and procedures for the facility. The Chemical Safety Engineer will have a minimum of

2 years experience in the preparation or review of chemical safety programs and procedures. 

This individual will hold a bachelor's degree (or equivalent) in an engineering or scientific field

and have successfully completed a training program, applicable to the scope of operations, in

chemistry and in associated safety practices.

The Shift Managers report to the Operations Manager and have the responsibility for ensuring

safe operation of enrichment equipment and support equipment. Each Shift Manager directs

assigned personnel, to provide enrichment services in a safe, efficient manner. Shift Managers

will have a minimum of 5 years of appropriate, responsible experience in implementing and

supervising a nuclear operations program.

The Production Scheduling Manager reports to the Uranium Management Manager and has the

responsibility for developing and maintaining production schedules for enrichment services. 

This individual will have a minimum of 3 years of appropriate, responsible experience in

implementing and supervising a continuous production scheduling program.
2-6

The Cylinder Management Manager reports to the Uranium Management Manager and has the

responsibility for ensuring that cylinders of UF6 are received and routed correctly at the facility,

and is also responsible for all transportation licensing. This individual will have a minimum of 3

years of appropriate, responsible experience in implementing and supervising a continuous

production scheduling program.

The Warehouse and Materials Manager reports to the Uranium Management Manager and has

the responsibility for ensuring spare parts and other materials needed for operation of the facility

are ordered, received, inspected, and stored properly. This individual will have a minimum of 3

years of appropriate, responsible experience in implementing and supervising a purchasing and

inventory program.

The Safeguards Manager reports to the HS&E Manager and has the responsibility for ensuring

the proper implementation of the FNMCP. This position is separate from, and independent, of

the Operations, Technical Services, and Human Resources departments, to ensure a definite

division between the safeguards group and the other departments. In matters involving

safeguards, the Safeguards Manager has direct access to the Plant Manager. The Safeguards

Manager will have as a minimum, a bachelor's degree in an engineering or scientific field, and 5

years of experience in the management of a safeguards program for Special Nuclear Material,

including responsibilities for material control and accounting. No credit for academic training

may be taken toward fulfilling this experience requirement.

The Chemistry Manager reports to the Technical Services Manager and has the responsibility

for the implementation of chemistry analysis programs and procedures for the facility. This

includes effluent sample collection, chemical analysis of effluents, comparison of effluent

analysis results to limits, and reporting of chemical analysis of effluents to appropriate

regulatory agencies. The Chemistry Manager will have, as a minimum, a bachelor's degree (or

equivalent) in either an engineering or a scientific field and 3 years of appropriate, responsible

nuclear experience associated with implementation of a facility chemistry program.

The Performance Manager reports to the Technical Services Manager and has the

responsibility for coordinating and maintaining testing programs for the facility. This includes

testing of systems and components to ensure that the systems and components are functioning

as specified in design documents. This individual will have, as a minimum, a bachelor’s degree

(or equivalent) in either an engineering or a scientific field and 4 years of appropriate,

responsible nuclear experience associated with implementation of testing programs. 

The Projects Manager reports to the Technical Services Manager and has the responsibility for

the implementation of facility modifications and for maintaining the configuration management

system. This individual also provides engineering support, as needed, to support facility

operation and maintenance, and support of performance testing of systems and equipment. 

The Projects Manager will have, as a minimum, a bachelor's degree (or equivalent) in an

engineering or scientific field and have a minimum of 5 years of appropriate, responsible

nuclear experience.

The Engineering Manager reports to the Technical Services Manager and has the responsibility

for providing engineering support at the facility. This includes ensuring the safe operation of

enrichment equipment and support equipment, providing maintenance support for equipment

and systems, and developing operating and maintenance procedures for the facility. The
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individual is responsible for the development of all design changes to the plant. The

Engineering Manager will have, as a minimum, a bachelor's degree (or equivalent) in an

engineering or scientific field and have a minimum of 5 years of appropriate, responsible

experience in implementing and supervising a nuclear engineering program.

The Maintenance Manager reports to the Technical Services Manager and has the

responsibility of directing and scheduling maintenance activities to ensure proper operation of

the facility, including preparation and implementation of maintenance procedures. This includes

activities such as repair and preventive maintenance of facility equipment. The Maintenance

Manager also has the responsibility for coordinating and maintaining testing programs for the

facility. This includes testing of systems and components to ensure that the systems and

components are functioning as specified in design documents. This individual will have, as a

minimum, a bachelor's degree (or equivalent) in an engineering or scientific field and 4 years of

responsible nuclear experience.

The Administration Manager reports to the Human Resources Manager and has the

responsibility for ensuring that support functions such as accounting, word processing, and

general office management are provided for the facility. This individual will have a minimum of 3

years of appropriate, responsible experience in implementing and supervising administrative

responsibilities at an industrial facility.

The Community Relations Manager reports to the Human Resources Manager and has the

responsibility for providing information about the facility and LES to the public and media. 

During an abnormal event at the facility, the Community Relations Manager ensures that the

public and media receive accurate and up-to-date information. This individual will have as a

minimum, a bachelor's degree in Public Relations, Political Science, or Business Administration,

and 3 years of appropriate, responsible experience in implementing and supervising a

community relations program.

The Security Manager reports to the Human Resources Manager and has the responsibility for

directing the activities of security personnel to ensure the physical protection of the facility. This

individual is also responsible for the protection of classified matter at the facility and obtaining

security clearances for facility personnel and support personnel. In matters involving physical

protection of the facility or classified matter, the Security Manager has direct access to the Plant

Manager. This individual will have, as a minimum, a bachelor's degree in an engineering or

scientific field and 5 years of experience in the responsible management of physical security at

a facility requiring security capability similar to that required for the facility. No credit for

academic training may be taken toward fulfilling this experience requirement.

The Document Control Manager reports to the Human Resources Manager and has the

responsibility for adequately controlling documents at the facility. This individual will have a

minimum of 3 years of appropriate, responsible experience in implementing and supervising a

document control program.

The Training Manager reports to the Human Resources Manager and has the responsibility for

conducting training and maintaining training records for personnel at the facility. This individual

will have a minimum of 5 years of appropriate, responsible experience in implementing and

supervising a training program.
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The minimum operating shift crew consists of a Shift Manager (or Deputy Shift Manager in the

absence of the Shift Manager); one Control Room operator; one Radiation Protection

technician; one operator for each Cascade Hall and associated UF6 handling systems; and

security personnel. When only one Cascade Hall is in operation, a minimum of two operators

are required.

At least one criticality safety engineer will be available, with appropriate ability to be contacted

by the Shift Manager, to respond to any routine request or emergency condition. This

availability may be offsite if adequate communication ability is provided to allow response as

needed.

The applicant has a program in place to make personnel position descriptions available onsite

for NRC inspections.

2.3.2 Management Control

Section 2.3 of the SAR (LES, 2005) summarizes how the activities that are essential for

implementation of the management measures and other HS&E functions are documented in

formally approved, written procedures, prepared in compliance with a formal document control

program. The mechanism for reporting potentially unsafe conditions or activities to the HS&E

organization and facility management is also summarized. This mechanism involves giving

employees that feel safety or quality is being compromised the responsibility and right to initiate

the “stop work” process to ensure work is returned to safe conditions. Employees also have the

right to access line management, the safety organization, requirements under 10 CFR Part 19,

and the Corrective Action Program to ensure their concerns are addressed.

2.3.2.1 Configuration Management

A Configuration Management program is provided to define and maintain a technical baseline

for the facility and provide a formal process for making changes to that baseline. All changes

made to the facility are made in accordance with the Configuration Management program. 

Section 11.3.1 of this Safety Evaluation Report (SER) discusses the Configuration Management

program.

2.3.2.2 Maintenance

A maintenance program will be implemented, during operations, that will include planned and

scheduled preventive maintenance, surveillance, and performance trending, to ensure that

IROFS are available and reliable to perform their intended functions.

2.3.2.3 Training and Qualifications

The applicant will implement a formal planned training program that will include indoctrination

training for all employees, addressing criticality, radiological, chemical, and industrial safety. 

The level of indoctrination training will depend on the specific jobs to be performed. Continued

or periodic retraining will be established, when applicable, to ensure employee proficiency. 

Further in-depth training will occur in the area of specific job areas. Radiological and criticality

safety retraining will occur annually. Training records will be maintained by the Human
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Resources Manager. Additional information on the applicant’s training program is provided in

Section 11.3 of the SAR (LES, 2005) and in Section 11.3.3 of this SER.

2.3.2.4 Procedures

The applicant will conduct all operations involving licensed material in accordance with

approved, written procedures. These procedures will generally include operating procedures,

administrative procedures, maintenance procedures, and emergency procedures.

2.3.2.5 Audits and Assessments

The applicant will implement a QA Program that requires periodic audits of activities affecting

quality, to ensure that these activities are being conducted in accordance with procedures and

the QA Program requirements. The audits will be identified, scheduled, and performed in

accordance with a written plan. The frequency of audits will depend on the safety significance,

status, and work history of the activity. The SRC and the QA Department will conduct

operational reviews and program audits. Further information on audits is provided in Section

11.5 of the SAR (LES, 2005).

2.3.2.6 Safety Review Committee

Section 2.2.3 of the SAR states that the SRC will report to the Plant Manager and will provide

technical and administrative review and audit of operations that could affect plant worker, public

safety, and environmental impacts. The scope of activities reviewed and audited by the SRC

shall, as a minimum, include the following:

• Radiation protection;

• Nuclear criticality safety;

• Hazardous chemical safety;

• Industrial safety, including fire protection;

• Environmental protection;

• As Low As is Reasonably Achievable (ALARA) policy implementation; and

• Changes in facility design or operations.

The SRC will conduct at least one facility audit per year for the above areas. The SRC will be

composed of at least five members, including the Chairman. Members of the SRC may be from

the LES corporate office or technical staff. The five members will include experts on operations

and all safety disciplines (criticality, radiological, chemical, industrial). The Chairman, members,

and alternate members of the SRC will be formally appointed by the Plant Manager, will have an

academic degree in an engineering or physical science field, and, in addition, will have a

minimum of 5 years of technical experience, of which a minimum of 3 years will relate directly to

one or more of the safety disciplines (criticality, radiological, chemical, industrial). 

The SRC will meet at least once per calendar quarter. 

Review meetings will be held within 30 days of any incident that is reportable to NRC. These

meetings may be combined with regular meetings. After a reportable incident, the SRC will

review the incident's causes, the responses, and both specific and generic corrective actions to

ensure resolution of the problem is implemented. A written report of each SRC meeting and 
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audit will be forwarded to the Plant Manager and appropriate managers within 30 days, and be

retained in accordance with the records management system.

2.3.2.7 Incident Investigations

A Corrective Action Program will be implemented to identify, investigate, analyze, and document

abnormal events that have the potential to threaten or weaken the applicant’s health, safety,

and environmental protection programs. Additional detail is provided in Section 11.3.6 of this

SER.

2.3.2.8 Employee Concerns

The applicant will implement a “stop-work” process for any employee who feels that safety or

quality could be compromised in any work activity. This program is implemented through

general employee indoctrination training and the applicant’s management, the safety

organization, NRC requirements in 10 CFR Part 19, and through the Corrective Action Program.

2.3.2.9 Records Management

The applicant will implement a records management program to control the preparation and

issuance of applicant documents. This document control program will include a formal process 

for preparing, reviewing, approving, and issuing revisions to documents. Further discussion of

the records management program is provided in Section 11.3.7 of this SER.

2.3.2.10 Written Agreements with Offsite Emergency Agencies

The applicant will coordinate emergency actions with appropriate State and local offsite

emergency agencies through written agreements. Further discussion of Emergency

Management is provided in Chapter 8 of this SER.

2.3.3 Transition from Design and Construction to Operations

LES is responsible for the design, QA, construction, testing, initial start-up, operation, and

decommissioning of the facility.

Toward the end of construction, the focus of the organization will shift from design and

construction to initial start-up and operation of the facility. As the facility nears completion, LES

will staff the facility to ensure smooth transition from construction activities to operation

activities. Urenco will have personnel integrated into the LES organization to provide technical

support during startup of the facility and transition into the operations phase.

As the construction of systems is completed, the systems will undergo acceptance testing as

required by procedure, followed by turnover from the construction organization to the operations

organization by means of a detailed transition plan. The turnover will include the physical

systems and corresponding design information and records. After turnover, the operating

organization will be responsible for system maintenance and configuration management. The

design basis for the facility is maintained during the transition from construction to operations

through the configuration management system described in Chapter 11, “Management
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Measures,” of the SAR (LES, 2005).

2.4 EVALUATION FINDINGS

The staff reviewed the organization and administration for the proposed facility according to

Chapter 2 of NUREG-1520 (NRC, 2002). The staff reviewed the applicant’s organization,

management position summaries and qualifications, and management controls. These

organizational and administrative elements describe: (1) clear responsibilities and associated

resources for the design, construction, and operation of the facility; and (2) its plans for

managing and operating the project. The staff reviewed these plans and commitments and

concludes that they provide reasonable assurance that an acceptable organization,

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

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

the facility per 10 CFR 30.33, 10 CFR 40.32, 10 CFR 70.22, 10 CFR 70.23, and 10 CFR

70.62(d).

2.5 REFERENCES

(ASME, 1994) American Society of Mechanical Engineers (ASME). ASME NQA-1, “Quality

Assurance Requirements for Nuclear Facility Applications,” 1994.

(ASME, 1995) American Society of Mechanical Engineers (ASME). ASME NQA-1a, Addenda

to “Quality Assurance Requirements for Nuclear Facility Applications,” 1995.

(LES, 2005) Louisiana Energy Services (LES). “National Enrichment Facility Safety Analysis

Report,” Revision 6, 2005.

(NRC, 2002) U.S. Nuclear Regulatory Commission (NRC). NUREG-1520, “Standard Review

Plan for the Review of a License Application for a Fuel Cycle Facility,” 2002.
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3.0 INTEGRATED SAFETY ANALYSIS (ISA) AND ISA SUMMARY

The purpose of this review is to ensure that the Integrated Safety Analysis (ISA) and ISA

Summary meet the regulatory requirements specified in 10 CFR Part 70, Subpart H, “Additional

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

Material.” The review determines whether appropriate hazards and baseline design criteria

have been addressed. The review also determined whether acceptable Items Relied on for

Safety (IROFS), management measures, and likelihoods and consequences have been

designated for higher-risk accident sequences, and whether, with IROFS, the performance

requirements of 10 CFR 70.61 have been met. For those cases involving nuclear criticality

safe-by-design components, the review determines whether the performance requirements of

10 CFR 70.61 are met through demonstration that failure of those components is highly unlikely. 

The review also determined whether programmatic commitments to maintain the ISA and ISA

Summary are acceptable.

In particular, this review considered information provided by the applicant related to:

1. The use of baseline design criteria for the design of the facility in accordance with 10

CFR 70.64(a).

2. Commitments regarding the applicant’s safety program, including the ISA, pursuant to

the requirements of 10 CFR 70.62; and

3. ISA summaries submitted in accordance with 10 CFR 70.62(c)(3)(ii) and 70.65.

3.1 REGULATORY REQUIREMENTS

The following regulatory requirements are applicable to the ISA and ISA Summary content:

1. 10 CFR 70.62 specifies the requirement to establish and maintain a safety program,

including performance of an ISA that demonstrates compliance with the performance

requirements of 10 CFR 70.61;

2. 10 CFR 70.62(c) specifies requirements for conducting an ISA, including a

demonstration that credible high-consequence and intermediate-consequence events

meet the safety performance requirements of 10 CFR 70.61;

3. 10 CFR 70.64 specifies requirements for baseline design criteria and facility and system

design and facility layout; and

4. 10 CFR 70.65(b) specifies the contents of an ISA Summary.

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 70.61. The safety

program is required to contain three elements: (1) process safety information; (2) an integrated
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safety analysis (ISA); and (3) management measures. The integrated safety analysis must be

conducted and maintained by the applicant and must identify the following:

• Radiological hazards related to possessing or processing licensed material;

• 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;

• The consequence and likelihood of occurrence of each potential accident sequence

identified and the methods used to determine the consequences and likelihood; and

• Each item relied on for safety (IROFS) identified pursuant to 10 CFR 70.61, the

characteristics of its preventive, mitigative, or other safety function and the assumptions

and conditions under which the item is relied upon to support compliance with 10 CFR

70.61.

The regulations, in 10 CFR 70.61, provide that the ISA must evaluate compliance with

performance requirements. Those requirements specify that the risk of each credible

high-consequence event must be limited such that the likelihood of occurrence is highly unlikely

and the risk of each credible intermediate-consequence event must be limited such that the

likelihood of occurrence is unlikely. 

The license application must include a description of the safety program under 10 CFR

70.65(a). In addition, the applicant is required to submit to the NRC an ISA Summary. The

Summary is required to contain:

• 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;

• A description of each process analyzed in the ISA in sufficient detail to understand the

theory of operation and, for each process, the hazards identified in the ISA and a

general description of the types of accident sequences;

• Information that demonstrates compliance with the performance requirements of 10 CFR

70.61, including a description of the management measures, requirements for criticality

monitoring and alarms and the information regarding the baseline design criteria and

defense-in-depth practices set forth in 10 CFR 70.64;

• A description of the team, qualifications, and the methods used to perform the ISA;
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• A list briefly describing each IROFS 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 consequences to

an individual from acute chemical exposure to licensed material or chemicals produced

from licensed material;

• A descriptive list that identifies all IROFS that are the sole item preventing or mitigating

an accident sequence that exceed the performance requirements of 10 CFR 70.61; and

• A description of the definitions of unlikely, highly unlikely, and credible, as used in the

evaluations in the ISA.

3.2 REGULATORY ACCEPTANCE CRITERIA

The acceptance criteria used during the U.S. Nuclear Regulatory Commission’s (NRC’s) review

of the applicant’s ISA and ISA Summary are outlined in Sections 3.4.3.1 and 3.4.3.2 of NUREG1520 (NRC, 2002).

3.3 STAFF REVIEW AND ANALYSIS

The Staff reviewed the safety program as described in the applicant’s ISA Summary (LES,

2005b) in order to assess compliance with the regulatory requirements. This includes

information describing the site, facility, processes, and baseline design criteria. The summary

also details the method used by the applicant to identify hazards associated with the processes

identified, a description of the accident sequences identified, the potential consequences for

each accident, and the identification of applicable IROFS for each accident found to be credible

for which consequences could be classified as intermediate or high. A description of the safety

function of each IROFS was described, along with the means by which the IROFS will be

implemented. 

The staff reviewed the applicant’s Safety Analysis Report (SAR), portions of the Integrated

Safety Analysis (ISA), and ISA Summary. Since the proposed facility has not been constructed,

the staff visited a similar gas centrifuge uranium enrichment facility at Almelo, The Netherlands,

to become familiar with the proposed processes and plant layout. The staff also conducted an

in-office review of the ISA at the AREVA engineering offices in Marlborough, Massachusetts,

and two in-office reviews of criticality, chemical safety, and other related documents at the

applicant’s Washington, D.C., offices. The staff analyzed the applicant’s proposed Safety

Program that includes the elements of process safety information, integrated safety analysis,

and management measures, to determine that the requirements of 10 CFR 70.62 are met. The

staff also conducted detailed, vertical slice reviews of various accident scenarios, selected on a

sampling basis, to confirm that the Safety Program and associated elements are adequately

implemented by the applicant to achieve the performance requirements of 10 CFR 70.61. 

In accordance with the guidance in Section 3.5.2.3 of NUREG-1520 (NRC, 2002), the vertical

slice review examined how the ISA method was applied to a selected subset of facility

processes in order to obtain reasonable assurance that ISA methods would be effective in the
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other processes not sampled by the staff. The staff reviewed the applicant’s HAZOP

methodology and confirmed that it met the guidance in NUREG-1513 and generally acceptable

industry practices (AIChE, 1989 and 1992). The HAZOP technique identifies and evaluates

safety hazards in process plants and the technique requires detailed information concerning the

design and operation of a process, and is typically used, as in this case, during or after the

detailed design phase. Implementation of the technique involves the use of an interdisciplinary

team and systematic approach to identify hazard and operability problems (i.e., accident

sequences). The results of the HAZOP analysis are the team’s findings, which include

identification of the accident sequences and items relied on for safety (IROFS). As a result of

the initial staff review, the applicant added a “safe-by-design” method to the ISA for application

to passive design component features related to nuclear criticality safety. The staff

subsequently determined that use of the HAZOP and “safe-by-design” ISA methods provided

reasonable assurance that the applicant identified all accident sequences that could exceed the

performance requirements of 10 CFR 70.61. 

Accident sequences related to chemical safety, nuclear criticality safety, and fire protection were

selected for a detailed staff “vertical-slice” review based on gas centrifuge uranium enrichment

process knowledge and professional judgement. The vertical-slice review examined how the

ISA methods were applied and examined appropriate safety information not included in the ISA

Summary. The vertical slice review included both high and intermediate consequence accident

scenarios. The purpose of the review was to determine whether accident sequences,

consequences, and likelihoods were reasonably determined, and whether appropriate IROFS

and management measures were selected to limit the risk of the analyzed events (i.e., highconsequence events to “highly unlikely,” and each intermediate-consequence events to

“unlikely”). The results of the staff’s vertical-slice review of a smart sample of accident

sequences in each technical discipline will provide reasonable assurance that, if the methods

described in the SAR, and discussed above, are appropriately applied by the applicant, all

accident sequences and related IROFS will be identified by the applicant. For nuclear criticality

safe-by-design components, the staff review determined whether those components met the

criteria for highly unlikely.

3.3.1 General Information

The staff evaluated information describing the site, facility, processes, baseline design criteria,

the safety program and integrated safety analysis (ISA) to determine whether the performance

criteria of 10 CFR 70.61 are met. The staff’s evaluation considered the applicant’s

implementation of the baseline design criteria for the facility under 10 CFR 70.64. The

development and implementation of the safety program, including the elements of process

safety information, integrated safety analysis and management measures, were reviewed to

confirm that the applicant had established an acceptable methodology for conducting an ISA. 

The results of the applicant’s ISA, contained in the ISA Summary, were further reviewed by the

staff to confirm that it identified appropriate hazards and associated accident sequences that

could exceed the performance requirements of 10 CFR 70.61, including likelihood and

consequence levels, and appropriate items relied on for safety (IROFS) and the measures to

assure that they will be available and reliable to prevent or mitigate those accident sequences.

3.3.1.1 Site Description
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A description of the proposed facility and site is provided in SAR Section 1.3 (LES, 2005a) and

ISA Summary Section 3.2 (LES, 2005b). The general description topics include site geography,

demographics and land use, meteorology, hydrology, geology, seismology and stability of

subsurface materials. External events, such as explosions and aircraft crashes, and natural

phenomena, including tornados, hurricanes, floods, and earthquakes, are assessed to

determine the likelihood of occurrence and their impact on the facility.

The regulations in 10 CFR 70.65(b)(1) require each application to include a general description

of the site in the ISA Summary, with emphasis on those factors that could affect safety (i.e.,

nearby facilities, meteorology, and seismology).

3.3.1.1.1 Site Geography

The proposed site is in Southeastern New Mexico in Lea County, approximately 1.6 km (1 mi)

west of the New Mexico-Texas border on the north side of New Mexico Highway 234. Andrews

County, Texas, lies across the border from the site. The site is about 8 km (5 mi) east of

Eunice, New Mexico, and 32 km (20 mi) south of Hobbs, New Mexico. The site is 220 ha (543

acres) in size and is located within County Section 32, Township 21 South, Range 38 East. The

site is owned by Lea County.

The proposed site is relatively flat with elevations between 1033 and 1045 m (3,390 to 3,430 ft)

above sea level. The site slopes to the southwest, is undeveloped, and is used for domestic

livestock grazing.

3.3.1.1.2 Demographics and Land Use

Information about demographics is provided in Section 1.3.2 of the SAR (LES, 2005a) and

Section 3.2.2 of the ISA Summary (LES, 2005b).

3.3.1.1.2.1 Local Population and Land Use

The proposed site is in Lea County, New Mexico, and is about 1.6 km (1 mi) from the New

Mexico-Texas border. Andrews County, Texas abuts the border on the Texas side. Together,

the counties have a combined population of 68,515, based on the 2000 census. In 1990, the

combined population was 70,130. This decrease is counter to the trends within New Mexico

and Texas, which had state-wide population increases of 20.1 percent and 22.8 percent,

respectively, over that 10-year period. The population decreases in Lea and Andrews Counties

are caused by decrease in petroleum industry jobs since the mid-1980s. It is expected that

population growth in these two counties in the next 30 years will be at a lower rate than the

overall rates in New Mexico and Texas.

Lea County covers 11,378 km2

 (4393 mi2), which is about three times the size of the State of

Rhode Island. Andrews County covers 3,895 km2

 (1504 mi2).

Major population centers near the proposed site include the following:
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Eunice, New Mexico, about 8 km (5 mi) west of the site;

Hobbs, New Mexico, about 32 km (20 mi) north of the site;

Jal, New Mexico, about 37 km (23 mi) south of the site;

Lovington, New Mexico, about 64 km (39 mi) north-northwest of the site;

Andrews, Texas, about 51 km (32 mi) east of the site;

Seminole, Texas (in Gaines County) about 51 km (32 mi) east-northeast of the site; and

Denver City, Texas (in Gaines County) 65 km (40 mi) north-northeast of the site.

Outside of these population centers, population density is very low. The nearest residences are

located about 4.3 km (2.6 mi) west of the proposed facility site.

Within 8 km (5 mi) of the site, land is primarily open land used for cattle grazing. Oil and gas

potentials are absent within this range, although operations are widespread beyond this area. 

Nearby industrial activities include a quarry and a “produced-water” reclamation company. Lea

County operates a county landfill on the south side of New Mexico Highway 234, and about 1.6

km (1 mi) east of the proposed site, Waste Control Specialists (WCS) operates a hazardous

chemical waste disposal facility and has licenses for the treatment and storage of low-level

radioactive and mixed wastes. A natural gas processing plant is located about 6 km (4 mi) from

the site.

3.3.1.1.2.2 Local Public Services

Fire fighting services are provided locally by Eunice Fire and Rescue, which is located 8 km (5

mi) from the proposed site. It is staffed by a full-time fire chief and 34 volunteer firefighters. 

Equipment includes three pumpers, one tanker, and three grass trucks. Eunice Fire and

Rescue also has agreements for mutual assistance with all Lea County fire departments.

Police and law enforcement services are provided by the Eunice Police Department, which has

five full-time officers. The Lea County Sheriff’s Department also has a substation in Eunice. 

Agreements between Lea and Andrews Counties provide mutual support when needed. The

New Mexico State Police can also provide support.

Educational institutions in Eunice include an elementary school, a middle school, a high school,

and a private K-12 school. The nearest other schools are in Hobbs, New Mexico, 32 km (20 mi)

north of Eunice. The nearest schools in Andrews County, Texas, are in Andrews, Texas, about

51 km (32 mi) from the proposed site.

There are two hospitals in Lea County – one is located in Hobbs, New Mexico, 32 km (20 mi)

north of the proposed site, and the other in Lovington, New Mexico, 64 km (39 mi) northnorthwest of the site. The hospital in Hobbs is a 250-bed facility capable of handling acute and

stable chronic care patients. The hospital in Lovington is a full-service, 27-bed facility. The

Eunice clinic is the nearest medical center to the proposed facility site. The nearest nursing

home facilities are in Hobbs, New Mexico.

There are no recreation facilities near the site. The Eunice Golf Course is located

approximately 15 km (9.2 mi) west of the site. A historical marker and picnic area are located

about 3.2 km (2 mi) west of the proposed site at the intersection of New Mexico Highways 234

and 18.
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3.3.1.1.2.3 Water Use

Southeast New Mexico has a semi-arid climate with an average annual precipitation of 33 to 38

cm (13 to 15 in.). The proposed site has no surface water and/or drainage features. Essentially

all precipitation either infiltrates the soil or is evapotranspirated. There are no significant bodies

of water or navigable waterways in the vicinity of the proposed site. There is also no agricultural

activity in the site vicinity although there are various crops grown in Lea and Andrews Counties. 

Cattle grazing does occur at the proposed site and in the nearby vicinity. Dairy farming is

important in Lea County, although none takes place near the site and/or in Andrews County.

Known sources of water near the site include: a man-made pond at the quarry, adjacent to the

proposed site, stocked with fish for private use; Baker Spring, which is an intermittent surfacewater feature, located about 1.6 km (1 mi) northeast of the proposed site; and several cattle

watering holes, where groundwater is pumped by windmills and stored in above-ground tanks.

3.3.1.1.2.4 Nearby Highways

Information concerning public roads is provided in Section 1.3.2.4 of the SAR (LES, 2005a) and

Section 3.2.1.2 of the ISA Summary (LES, 2005b). The New Mexico State Highway 234 passes

along the southern boundary of the proposed facility. Vehicles transporting propane travel this

highway at a relatively high frequency (Snooks, 2003).

The risk associated with the potential hazard of a highway propane explosion to the proposed

facility was analyzed by the applicant (Snooks, 2003). The analysis used the largest volume of

propane transported at one time. This volume was determined based on discussions with

propane operators (Snooks, 2003). During an accident, a large truck with a bounding gross

weight of 4536 kg (10,000 lb) was assumed to be totally crashed. Also, the structures of the

proposed facility for consideration of an explosion impact were assumed to be designed to

withstand 6.9 kPa (1 psi) overpressure, as suggested by Regulatory Guide 1.91 (NRC, 1978b). 

Based on this design-basis overpressure, a safety-significant structure must be at least 0.4 km

(0.24 mi) (381 m (1,251 ft)) from the point of explosion, to avoid damage. This safe-separation

distance is the approximate distance from New Mexico State Highway 234 to the proposed

TSB, a safety-significant structure.

The applicant calculated the likelihood of a propane truck accident, on Highway 234, that could

have an effect on the proposed facility, using information including annual truck accident rate

and miles, annual number of shipments passing the proposed facility, conditional probability of

occurrence of significant incidents from the accidents recorded, and exposure distance of a

structure in miles. The U.S. truck occupancy safety data from 1990 through 2001 were obtained

from the Bureau of Transportation Statistics (Snooks, 2003). However, only data from the most

recent 5-year period (1997–2001) were used in the analysis. The applicant indicated the most

recent data better represent the current and future statistics because of the improvements in

transportation equipment and roadway conditions (LES, 2004a). The same 5-year data for

cargo tank truck incidents were used to calculate the conditional probability of significant

incidents. The number of propane shipments passing the proposed facility on New Mexico

State Highway 234 was obtained by interviewing the local propane operators. The exposure

distance used in the analysis was twice the safe separation distance for conservatism. The

probability of an incident based on the data discussed in this paragraph was determined to be

2.07 × 10!6

 (Snooks, 2003). The applicant defined not credible with a likelihood of occurrence
3-8

less than 10!6

, which is consistent with that suggested in NUREG–1520 (NRC, 2002a), and

accepted by the staff. The applicant concluded that a propane truck explosion was a credible

event. This probability, however, meets the definition of highly unlikely. Therefore, the potential

consequence does not have to be determined because the event sequence is highly unlikely in

accordance with 10 CFR 70.61(b).

Summary

Based on review of the highway propane-explosion-hazard risk assessment, staff concludesI: 

(i) the data used in the analysis are from acceptable sources; and (ii) the approach used in the

analysis for likelihood determination is acceptable because it relied on conservative bounding

assumptions and a design-basis over-pressure suggested by Regulatory Guide 1.91 (NRC,

1978b).

3.3.1.1.2.5 Nearby Gas Pipelines

Information concerning gas pipelines passing through or located near the proposed facility site

is provided in Sections 1.3.2.4 of the SAR (LES, 2005a) and 3.2.2.4 of the ISA Summary (LES,

2005b).

Natural Gas Pipeline

The applicant identified an underground natural gas pipeline located along the south property

line running parallel to New Mexico State Highway 234. This 40.6-cm (16-in.) natural gas

pipeline is a low-pressure line (<345 kPa (<50 psi)) and is located approximately 545 m (1800 ft)

from the proposed TSB, a safety-significant structure. The 40.6-cm (16-in.) natural gas pipeline

is buried approximately 0.9 m (3 ft) below the surface (Thomson, 2004). The natural gas

transported by the pipeline includes 72 percent methane, 11 percent ethane, 7 percent propane,

and less than 1 percent hydrogen sulfide. The pipeline gas flow is between 5.7 × 103

 and 1.4 ×

104

 m3/day (2 × 105 and 5 × 105 ft3/day). A parallel 35.6-cm (14-in.) gas pipeline also is

identified, but is not in use. 

A hazard risk assessment for the 40.6-cm (16-in.) natural gas pipeline was performed by the

applicant (Thomson, 2004). The hazards associated with a natural gas pipeline explosion may

include blast overpressure, missile generation, and thermal radiation. The assessment

performed by the applicant considered all these hazards, and the likelihood of a gas pipeline

explosion causing damage to safety-significant structures of the proposed facility was

determined by summing the probabilities of these hazards.

The applicant (Thomson, 2004) pointed out that the explosion-generated missile hazard

depends on several factors. Because of insufficient information on these factors, the applicant

assumed every natural gas detonation would result in a missile-impact hazard. Therefore, the

probability for a missile hazard was the same as the explosion probability. Staff finds this

assumption conservative and acceptable.

In addressing the potential thermal effects, the applicant stated that the potential thermal effects

might be bounded by a similar analysis involving a natural gas pipeline presented by the

Tennessee Valley Authority (TVA) in its Preliminary Safety Analysis Report for Hartsville

Nuclear Plants (Thomson, 2004). The natural gas pipeline analyzed by TVA was 55.9 cm (22
3-9

in.) in diameter, with a 3861-kPa (560-psi) operating pressure, whereas the natural gas pipeline

in question for the proposed facility is 40.6 cm (16 in.) in diameter, with a 345-kPa (50-psi)

operating pressure. The distance of the natural gas pipeline to the proposed facility is less than

that to the Hartsville Nuclear Plants [545 versus 808 m (1800 versus 2650 ft)]. The applicant

contended, considering the conservatism in the pipeline size and operating pressure, that TVA

results for radiant heat flux would bound the results for the safety-significant structures of the

proposed facility if a detailed analysis were performed. For a worst-case condition, the radiant

heat incidence obtained by TVA was less than 9085 kJ/m2

 (800 Btu/ft2). To cause spontaneous

ignition of wood, a radiant heat incidence of 19,874 kJ/m2

 (1750 Btu/ft2) would be required

(Thomson, 2004, Attachment 9). Therefore, a much higher radiant heat flux exposure would be

necessary to cause damage to concrete structures. As indicated in the ISA Summary (LES,

2005b), the safety-significant structures of the proposed facility will be concrete structures. 

Therefore, the applicant stated (Thomson, 2004) the explosion-induced thermal hazards on the

proposed facility might be neglected. Staff agrees with the applicant’s assessment that the

radiant heat flux resulting from a potential natural gas flame near the safety-significant

structures of the proposed facility would be bounded by the results calculated for the TVA

facility. Consequently, the associated thermal effects on the proposed facility would be

negligible.

In estimating the probability of pipeline explosion-induced blast overpressure hazard, the

applicant considered three parameters: (1) gas pipeline rupture incidents per mile; (2)

conditional probability for significant incidents; and (3) exposure distance in miles. Data from

the Office of Pipeline Safety (OPS) web site (OPS, 2004) were used by the applicant to

calculate gas pipeline rupture incidents per mile. These data included the annual mileage

related to natural gas transmission operations (1984–2003) and the detailed accounts, including

rupture length (mid-1984–present) and telephone records (1987–2001) of the reported

accidents. The data for annual mileage presented in Thomson (Thomson, 2004, Table 1) were

slightly lower than the data currently posted on the OPS web site. Staff does not expect this

slight difference to affect significantly the gas pipeline rupture incidents per mile. The applicant

selected 4 years of data (1998–2001) to calculate gas pipeline rupture incidents per mile. The

applicant (LES, 2004a) pointed out that these 4 years of data are comparable to the number of

gas pipeline rupture incidences for other years and, therefore, representative of the available

rupture incidence data.

Among the rupture incidents in these 4-year data, incidents with a recorded rupture length of

less than 3.1 cm (0.1 ft) were not included in the analysis. The applicant indicated these

incidents would not be a hazard to the safety-significant structures, based on past experience

(Thomson, 2004; LES, 2004a). Fifty incidents were identified to have had a rupture length

greater than 3.1 cm (0.1 ft). The gas pipeline rupture incidents per mile, the ratio of these

rupture incidents to the sum of the annual mileages, was approximately 5.7 × 10!5

 ruptures per

mile (Thomson, 2004). Staff determined that the applicant’s justification for: (I) use of the 4-

year data (1998–2001) for natural gas pipeline hazard analysis is acceptable; and (ii) the use of

a limiting rupture length [3.1 cm (0.1 ft)] to screen out nonconsequential rupture incidents from

further consideration are reasonable.

Not all rupture incidents would involve explosions. The rupture incidents without explosions

would not generate blast overpressure, or missiles, or induce thermal radiation; thus, these

incidents would not be a safety concern. Of the 50-rupture incidents used in the analysis, the

applicant identified seven which were explosion related (Thomson, 2004). These explosion
3-10

incidents represented a fraction of 0.14 (defined as explosion probability, to facilitate discussion)

of the rupture incidents. Gas cloud explosions may be classified as deflagrations or detonations

(NFPA, 1995). Detonations are rapid explosions, generating supersonic pressure waves and

capable of producing blast over-pressures sufficient to cause damage. Deflagrations, on the

other hand, are relatively slow explosions produced by rapid chemical reactions which generate

only subsonic pressure waves. Typically, 28 percent of the explosions are estimated to be

detonation related (NAESC, 1999). The product of the detonation rate and the explosion

probability formed the conditional probability, and this conditional probability was 0.0392. The

applicant used this conditional probability to modify the probability of gas pipeline rupture

incidents per mile, to reflect the fact that only large explosions could produce blast

overpressures sufficient to cause damage to the structure in question.

The exposure distance is the length of a natural gas pipeline segment, with the explosion

resulting from a rupture at any point of this segment posing a safety concern to the structures. 

This exposure distance is a function of the safe-separation distance. To determine the

safe-separation distance, two aspects need to be considered. One is the distance, D1, from a

gas-release location along a direct pathway to the proposed facility to the lower flammable limit

of a gas plume, and D3 to the upper flammable limit. Within these two limits, the flames from an

ignited gas cloud could propagate in a self-sustaining manner (CCPS, 2002). Another aspect is

the distance, D2, estimated from the edge of the gas plume to a safety-significant structure

designed to withstand an overpressure of 6.9 kPa (1 psi).

The applicant used the computer program ALOHA (USEPA, 1999) to estimate D1 and D3. D1

and D3 were estimated to be 1248 and 665 m (4095 and 2181 ft) (Thomson, 2004). In the

calculation, the gas plume was assumed to contain solely methane. This assumption is

acceptable because methane is the major constituent of natural gas. The mass of the released

methane within the flammable range was then determined using D1, D3, and the wind speed in

the ALOHA calculation. After the trinitrotoluene weight of the methane mass was obtained

using the equation suggested by the National Fire Protection Association and Society of Fire

Protection Engineers (NFPA, 1995), the equation in Regulatory Guide 1.91 (NRC, 1978b) was

then used for D2. In calculating D2, two important parameter values (yield and theoretical net

heat of combustion) were obtained from two fire protection handbooks (NFPA, 1995; NFPA,

1991). The staff finds the values used are appropriate and from reliable sources. D2 was

determined to be approximately 448 m (1471 ft) (Thomson, 2004). The minimum distance

required for a safety-significant structure to avoid damage from a pipeline explosion

(safe-separation distance) was D1 + D2 (i.e., 1697 m (5566 ft)). Because the closest location of

the natural gas pipeline of concern was approximately 545 m (1800 ft) from a safety-significant

structure of the proposed facility, hazards associated with a substantial section of the natural

gas pipeline would pose a safety concern for that structure. This pipeline section was located

within a circle centered at the edge of the structure with a radius of D1 + D2 (i.e., 1697 m (5566

ft)). The applicant estimated conservatively the length of the pipeline section (exposure

distance) to be the diameter of the circle (i.e., approximately 3.45 km (2.1 mi)). 

Several parameters were used in the ALOHA calculation (Thomson, 2004)

• Postulated release from a hole with a diameter equal to that of the pipeline. This

assumption is bounding because the amount of gas flow through the pipeline at a given

time is controlled by the pipeline diameter and gas pressure and this amount is the
3-11

maximum that can be released at a given time should there be a natural gas pipeline

rupture.

• Release duration—1 hour. This value was the maximum expected time required to shut

off the gas supply and bleed the system. According to the pipeline operator, the natural

gas pipeline in question is 22.5-24.1 km (14-15 mi) in length with three manual shutoff

valves. Two of the valves are located at the end and one is located in the middle of the

pipeline (Thomson, 2004, Attachment 5: LES, 2004d). To reach, and manually shut off

one of the valves in responding to a natural gas pipeline rupture accident, the pipeline

operator estimated that approximately one hour would be required. Once the gas supply

is shut off and the pipeline is bled, no more natural gas would be available for release. 

Furthermore, the applicant indicated that, in the event of a rupture, a steady-state

concentration would be reached in less than one hour (Thomson, 2004). Consequently,

the use of one hour as the release duration in the ALOHA analyses is reasonable and

acceptable to the staff.

• A stable atmosphere, with minimal dispersion. The wind speed of 1 m/s (3.3 ft/s) used in

the analysis was in the range of stable wind speeds of Pasquill Class F for the proposed

facility site (LES, 2004a; LES, 2005c). This stable wind class occurs 2.2 percent of the

time for the site (LES, 2005c). The selection of wind speed is consistent with Regulatory

Guide 1.78 (NRC, 2001).

With these, and all the parameters which are known, the applicant estimated the probability of

hazards associated with a pipeline explosion-induced blast overpressure to be 4.7 × 10!6

/year

(Thomson, 2004). As indicated earlier, the probability of a missile hazard was assumed to be

equal to the probability of a pipeline explosion. Consequently, the final probability of the natural

gas pipeline hazard affecting the structures was 9.4 × 10!6

/year. The applicant concluded the

natural gas pipeline explosion event was a credible event. This probability, however, meets the

definition of highly unlikely. Therefore, the potential consequences do not have to be

determined because the event sequence is highly unlikely, in accordance with 10 CFR 70.61(b).

Carbon Dioxide Pipeline

An underground 25.4-cm (10-in.)-diameter, high-pressure carbon dioxide pipeline currently runs

across the property. The applicant (2005a) stated that the normal operating pressure of the

pipeline was 13,445 kPa (1950 psi), and the maximum operating pressure was 14,479 kPa

(2100 psi) (Thomson, 2004; LES, 2004a). The ISA Summary (LES, 2005b) indicated this

pipeline will be relocated along the western and southern boundary of Section 32 so the pipeline

will be positioned at least 396.2 m (1300 ft) west of the facility restricted area and is

approximately 945 m (3100 ft) [estimated from Figure 3.2-3 of the ISA Summary (LES, 2005b)]

from the Separations Building Module, a safety-significant structure that houses two cascade

halls. The applicant concluded that, at this distance from the proposed facility, the pipeline was

not a safety concern. Staff agrees with the applicant’s assessment that the carbon dioxide

pipeline is not a safety concern to the Separations Building Module.

Figure 3.2-3 of the ISA Summary (LES, 2005b) shows that a portion of the carbon dioxide

pipeline, at the southern boundary of Section 32, is close to the two natural gas pipelines

located south of the proposed facility site. As noted earlier, the natural gas pipeline is

approximately 545 m (1800 ft) south of the TSB. Even though carbon dioxide is not flammable,
3-12

pipeline ruptures resulting from the high pressure might be credible. According to the pipeline

operator, a clearance requirement of 6- to 9- m (20- to 30-ft) minimum separation is normally

required between parallel running pipelines and pipelines crossing one another for safety

concerns (LES, 2004d). Because of this clearance requirement, the applicant indicated rupture

of the high pressure carbon dioxide pipeline will not affect the hazard probability of the natural

gas pipeline (LES, 2004d). The staff concludes that the applicant’s assessment is acceptable.

Onsite Natural Gas Pipeline

According to the applicant (Harper, 2003a), the proposed facility will include an on-site natural

gas pipeline. This pipeline will be used to provide natural gas for heating the boiler in the

Central Utility Building (CUB).

The potential hazard analyzed for this pipeline was an explosion within the CUB (Harper,

2003a). By examining the rupture incidents per mile (5.7 × 10!5

 ruptures per mile) for the

natural gas pipeline and the likelihood of a rupture leading to an explosion, the applicant

determined that a scenario of a gas explosion in the CUB would be highly unlikely. It would be

even less likely for a detonation to occur with sufficient energy to damage the CUB and the

nearby TSB or Separations Building Modules. 

At the present time, the pipeline to be used for on-site natural gas supply is a 10-cm (4-in.)

diameter, low-pressure [69-kPa (10-psi)] line (LES, 2004a). This pipeline will have an excess

flow valve located at the entrance of the proposed facility site, and this excess flow valve will

automatically shut the gas flow off in case of a pipeline leak. Because the diameter of the onsite natural gas pipeline is small, the operating pressure is low, and the excess flow valve is

used, a potential explosion of the on-site natural gas pipeline that is of sufficient magnitude to

damage the safety-significant structures of the proposed facility is considered not credible (LES,

2004a). 

Staff reviewed the hazard assessments conducted for the natural gas pipelines and concludes

the approach used to assess the natural gas pipeline hazards is acceptable, and the data used

for the assessment were from reliable sources. Staff also concludes that the assessments

performed for the carbon dioxide and the on-site natural gas pipeline hazards are based on data

from reliable sources and based on an acceptable approach.

3.3.1.1.2.6 Nearby Air Transportation

Information concerning potential aircraft crash hazards is provided in Sections 1.3.2.4 of the

SAR (LES, 2005a), and Section 3.2.1.2.4 of the ISA Summary (LES, 2005b). The analysis

concerning the potential hazard of aircraft crash to the proposed facility site was documented in

the aircraft hazard risk determination report (Yeung, 2003). The information, including the

number of operations and holding patterns of six local airports, was analyzed by Yeung (2003). 

These airports include:

• Lea County Regional Airport—40 km (25 statute miles) northwest of the proposed facility

site;

• Eunice Airport—24 km (15 statute miles) west of the proposed facility site;
3-13

• Lea County/Jal Airport—40 km (25 statute miles) south-southwest of the proposed

facility site;

• Andrews County Airport—48 km (30 statute miles) east of the proposed facility;

• Gaines County Airport—48 km (30 statute miles) northeast of the proposed facility; and

• Seminole Spraying Services (private)—48 km (30 statute miles) northeast of the

proposed facility.

The information about the number of operations and holding patterns of each airport was

obtained from the Federal Aviation Administration (FAA) (Yeung, 2003). Military flights are

operated from the Lea County Regional Airport. The number of military operations is included in

the number of operations for the Lea County Regional Airport. In analyzing potential aircraft

crash hazards, the applicant used the proximity criteria provided in Section 3.5.1.6 of NUREG0800 (NRC, 1981) to screen out the hazards associated with some flight activities from further

consideration. According to the proximity criteria, aircraft crash hazards do not have to be

considered if (I) the distance, D, of the facility to the airport is between 5 and 10 statute miles

and the projected number of annual operations is less than 500 D2

, or D is greater than 10

statute miles and the projected number of operations is less than 1000 D2

, (ii) the facility is at

least 5 statute miles from the edge of military training routes, including low-level training routes,

except those associated with a usage greater than 1000 flights per year, or where activities

(such as practice bombing) may create an unusual stress situation, and (iii) the facility is at least

two statute miles beyond the nearest edge of a federal airway, holding pattern, or approach

pattern. Based on the first proximity criterion provided in Section 3.5.1.6 of NUREG–0800

(NRC, 1981), the applicant determined the presence of these airports not to be a safety concern

for the proposed facility. 

Based on information from the FAA (Yeung, 2003), the applicant concluded the holding patterns

for four of the six airports were, in general, more than the 3.2-km (2-statute miles) proximity

criterion (third criterion) provided in Section 3.5.1.6 of NUREG–0800 (NRC, 1981) for airway

distance from the site of interest. The applicant pointed out no specific holding patterns existed

for Eunice or Lea County/Jal Airport (LES, 2004a; LES, 2005a). For the Eunice Airport, the

annual operations are small, approximately 480 flights per year. This number is substantially

smaller than the threshold limit (225,000 annual operations) provided in NUREG–0800 (NRC,

1981). The applicant determined the operation of the Eunice Airport does not pose any aircraft

crash hazard to the proposed facility because of the relatively small number of annual

operations. For the Lea County/Jal Airport, the applicant indicated the airport is more than 32

km (20 mi) away from the proposed facility, and the landing procedure usually will not be

initiated until an aircraft is within 32 km (20 mi) of the airport (LES, 2004a; LES, 2005a). 

Therefore, even if an aircraft is placed in a holding pattern, it will not bring the aircraft near the

proposed facility. Consequently, there is no aircraft crash hazard from holding patterns to the

proposed facility from the operations of the Lea County/Jal Airport. The staff concurs with the

conclusion made by the applicant concerning the potential hazards associated with the random

aircraft holding patterns at the Eunice and Lea County/Jal Airports. 

The applicant identified a low-level federal airway passing within 9 km (-6 statute miles)

northeast of the proposed facility. Using the method provided in Section 3.5.1.6 of

NUREGS0800 (NRC, 1981), the probability of an aircraft on the airway crashing onto the
3-14

proposed facility was estimated to be 3.4 × 10!7

 (Yeung, 2003). This probability makes the

aircraft crash an incredible event or hazard to be considered in either design or integrated

safety analysis. 

Three military training routes were identified in the region. The closest route to the proposed

facility was approximately 26 km (16 statute miles) southwest of the site. Based on the second

proximity criterion provided in Section 3.5.1.6 of NUREG–0800 (NRC, 1981), this military

training route posed no safety threat to the proposed facility.

Based on the review of aircraft-crash risk assessment, staff concludes that: (i) the aircraft

transportation information used in the analysis was obtained from a reliable source; and (ii) the

risk of public exposure was evaluated using the acceptable proximity criteria and method

provided in Section 3.5.1.6 of NUREG–0800 (NRC, 1981).

3.3.1.1.2.7 Demographics and Land Use Summary

The staff reviewed the site demographic information presented by the applicant and concludes

that the applicant has adequately described and summarized general site demographical

information related to local population, identification of population centers, schools, commercial

facilities, land use, and water use. Population information is provided based on the latest

census information.

3.3.1.1.3 Meteorology

3.3.1.1.3.1 Tornado Hazard

Information about the tornadoes and design-basis tornado at the proposed facility is provided in

Sections 1.3.3.3 of the SAR (LES, 2005a), and Section 3.2.3.4.1 of the ISA Summary (LES,

2005b).

There is an average of nine tornadoes a year in New Mexico, and the occurrence of tornadoes

in the vicinity of the proposed facility is rare. Tornadoes are classified using the Fujita Tornado

Damage Scale (F-scale) with classifications ranging from F0-F5 (NOAA, 2005). Eighty-seven

tornadoes of low magnitude (F0 to F2) were reported in Lea County, New Mexico, between

January 1, 1950, and December 31, 2004. Only one additional tornado was reported as F3 on

May 17, 1954. Two tornadoes, one in 1998 and the second in 1999, had a magnitude of F0 and

were located near Eunice. All the reported tornadoes were associated with very light damage

(NCDC, 2005).

The Modified IDR tornado hazard assessment model (McDonald, 1995) is used to quantify the

tornado risks at the facility. The tornado hazard assessment (1) defines the local region that

surrounds the site, (2) determines the occurrence rate and associated confidence limits, (3)

determines the number of tornados per F-scale category, (4) estimates the damage path area of

each F-scale tornado and calculates damage areas with confidence limits, and (5) calculates

the tornado hazard probabilities for each F-scale wind speed category. Due to insufficient

damage path area data for the one-degree square area (4034 square miles) surrounding the

proposed facility, a six-degree area was defined. A linear regression analysis was performed to

obtain a continuous area-intensity function. The total number of recorded tornados in the local

region for the years of 1954 through 1999 were broken down into the respective F-scale and the
3-15

mean damage path was determined from the historical tornado records. The results of the

tornado hazard assessment identified an expected annual probability of an F-3 tornado with 260

km/h (162 mph) wind speed occurring within the one-degree square area where the plant site is

located as 1.21 E-05. When both straight winds and tornado hazards are considered, the

expected wind speed with an annual probability of occurrence of 1.0 E-05 is 302 km/h (188

mph). 

The design parameters for the design-basis tornado are listed in ISA Summary Section

3.3.2.2.2.1 (LES 2005b). All safety-significant structures will be designed to withstand the

design-basis tornado. The development of these tornado design-basis-related parameters was

discussed in SAR Section 1.3.3.3 (LES 2005a), and ISA Summary Section 3.2.3.4 (LES,

2005b). The staff review and acceptance of the development of the design-basis tornado are

discussed in Section 1.3.3.3.1 of this SER. The values of design-basis tornado-related

parameters are listed as follows:

Design-Basis Wind Speed: 302 km/h (188 mph)

Radius of Damaging Winds: 130 m (425 ft)

Atmospheric Pressure Change: !3.83 kg/m2

 (!80 lbs/ft2)

Rate of Atmospheric Pressure Change: !1.44 kg/m2

/s (!30 lbs/ft2/s)

The tornado design-basis is characterized as a 100,000-year tornado, which is the equivalent of

a “F-3" tornado on the Fujita Tornado Scale.

To estimate the tornado hazard, the assessment model proposed by McDonald and Lu

(McDonald, 1995) was used. The local region selected for the assessment was the 1-degree

square area between latitudes 31.94° and 32.94° and longitudes 102.58° and 103.58°. 

According to the applicant (Harper, 2003b), tornadoes in this region were determined to most

likely affect the site.

The historical tornado records from 1954 to 1999 for the selected region were used for the

assessment. Linear regression analysis was performed by the applicant to obtain the

occurrence-intensity relationship. The applicant defined the design-basis tornado for the

proposed facility as the tornado with a return period of 100,000 years. Based on the

occurrence-intensity relationship, the wind speed for this design-basis tornado was determined

to be 302 km/h (188 mph). The design parameters, including atmospheric pressure change,

rate of atmospheric pressure change, and radius of damaging winds, were calculated for the

design-basis tornado using the tornado wind field model proposed by McDonald (NRC, 1983). 

Based on this model, the atmospheric pressure change for the design-basis tornado was 3.83

kPa (80 psf), the rate of atmospheric pressure change was 1.44 kPa/s (30 psf/s), and the radius

of damaging wind was 130 m (425 ft).

The applicant also established the damage area-intensity relationship in terms of the F-Scale for

the proposed facility site. Because the data for the damage path areas associated with the

tornadoes for the selected 1-degree square region were not sufficient to establish a damage

area-intensity relationship, the applicant used the damage path area data related to tornadoes

in a 6-degree square region (bounded by latitude 30°S35° and longitude 100°S105°), containing

the proposed facility site, to determine the damage area-intensity relationship.
3-16

The staff concludes that the design-basis tornado is acceptable because it was estimated based

on an acceptable method and was based on a 100,000-year return period, sufficient to make

the frequency of more damaging tornados highly unlikely.

3.3.1.1.3.2 High Winds and Hurricanes

Information about high winds at the proposed facility is provided in Sections 1.3.3.1 of the SAR

(LES, 2005a); Section 3.2.3.4.2 of the ISA Summary (LES, 2005b), and Section 3.6.1.4 of the

Environmental Report (ER) (LES, 2005c).

According to the SAR (LES, 2005a), no meteorological data were available for the proposed

facility site. Although the measured wind data at Midland–Odessa, Texas, and Roswell, New

Mexico, were discussed in the SAR (LES, 2005a) and ER (LES, 2005c), the Midland–Odessa

annual extreme wind data were used exclusively to estimate the high-wind hazard at the

proposed facility site (LES, 2005b; Harper, 2003b). The annual extreme data used range from

1973 through 1999. The wind speeds were 3-second gust speeds measured at 10 m (32.8 ft)

above ground. The Midland–Odessa weather station is located at the regional airport

approximately 103 km (64 mi) east-southeast of the proposed site, whereas the Roswell station

is approximately 161 km (100 mi) northwest of the proposed site. The climate data for both

locations were collected by the National Oceanic and Atmospheric Administration (NOAA) (LES,

2005c).

The largest wind speed for the annual extreme straight-line winds from 1973 through 1999 at

Midland–Odessa was 140 km/h (87 mph) and the smallest annual extreme straight-line wind

speed was 84 km/h (52 mph) (Harper, 2003b). The mean and standard deviation wind speeds

were 111.5 and 16.6 km/h (69.3 and 10.3 mph).

The high-wind hazard assessment was performed by fitting the annual extreme wind data using

the Fisher–Tippett Type I distribution model. The applicant chose the speed of a wind with an

annual probability of 1.0 × 10!5

 for the design-basis straight-line wind speed for the proposed

facility. This design-basis straight-line wind speed was 252 km/h (157 mph) (LES, 2005b). All

safety significant structures will be designed for this wind speed.

Because the proposed facility is not located near the coastal area (805 km (500 mi) from the

coast), hurricanes affecting the coastal area will have no effect on the performance of the

proposed facility. Consequently, consideration of hurricane hazards on the design of the

proposed facility is not needed.

Based on the review of the information concerning high winds, the staff concludes that highwind hazards and the associated design-basis straight-line winds have been addressed

acceptably because the data used for assessment were from a recognized source and the

method used for assessing high-wind hazards is an accepted method.

3.3.1.1.3.3 Temperature Extremes

Information about the temperature at the proposed facility site is provided in Section 3.6.1.2 of

the ER (LES, 2005c).
3-17

The regional temperatures in Hobbs, New Mexico [32 km (20 mi) north of the proposed facility

site]; Midland–Odessa, Texas; and Roswell, New Mexico, are discussed in the ER. The

discussions are based on 30-year records (from 1971 through 2000). As indicated previously,

NOAA collected the climate data for Midland–Odessa and Roswell. However, the Western

Regional Climate Center collected the climate data for Hobbs (LES, 2005c).

The highest recorded monthly mean maximum temperature was 38.9 °C (102.1 °F), and the

lowest recorded monthly mean minimum temperature was !5.1 °C (22.8 °F) for Hobbs,

New Mexico. No such data were presented in the ER for Midland–Odessa or Roswell. The

highest daily maximum and lowest daily minimum temperatures were 46.7 °C (116.0 °F) and

!23.9 °C (!11.0 °F) for Midland–Odessa, and 45.6 °C (114.0 °F) and !22.8 °C (!9.0 °F) for

Roswell. No such data were presented for Hobbs. As indicated, the highest daily maximum

and the lowest daily minimum temperatures for Midland–Odessa and Roswell were similar.

The staff reviewed the temperature information and concludes that the information is acceptable

because recognized data sources were used and the temperature extremes are properly

determined.

3.3.1.1.3.4 Extreme Precipitation

Section 1.3.3.2 of the SAR (LES, 2005a); and Sections 3.2.3.2 and 3.2.3.4.4 of the ISA

Summary (LES, 2005b) discuss the rainfall precipitation at the proposed facility site. The

precipitation data for Hobbs, Midland–Odessa, and Roswell were listed in Tables 3.2-14 through

3.2-16 of the ISA Summary (LES, 2005b). These data were collected from the Western

Regional Climate Center and NOAA and are based on data from 1971 through 2000

(LES, 2005c). The maximum monthly totals were 35.13 cm (13.83 in.) for Hobbs; 24.6 cm (9.7

in.) for Midland–Odessa; and 17.5 cm (6.88 in.) for Roswell. The minimum monthly totals were

zero for all locations. The highest 24-hour precipitation was 15.2 cm (5.99 in.) for

Midland–Odessa and 12.5 cm (4.91 in.) for Roswell.

According to the SAR (LES, 2005a), the local intense probable maximum precipitation was

estimated from NOAA data (NOAA, 1982). The local intense probable maximum precipitation

was approximately 43.9 cm (17.3 in.) in 1 hour, over a 2.6-km2

 (1-mi2) area.

The staff reviewed the information concerning regional precipitation and local intense probable

maximum precipitation presented in the SAR (LES, 2005a) and the ISA Summary (LES, 2005b)

and concludes that the information is acceptable because recognized data sources, such as

NOAA, were used.

3.3.1.1.3.5 Snow

Section 1.3.3.2 of the SAR (LES, 2005a), and Section 3.2.3.3 of the ISA Summary (LES, 2005b)

discuss the regional snowfall. NOAA collected the snowfall data. The maximum monthly

snowfall/ice pellets were 24.9 cm (9.8 in.) for Midland–Odessa and 53.3 cm (21.0 in.) for

Roswell. The maximum snowfall/ice pellets during a 24-hour period were 12.47 cm (4.91 in.) for

Midland–Odessa and 41.91 cm (16.5 in.) for Roswell. No snowfall information was available for

Hobbs, New Mexico.
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The staff reviewed the information concerning snow precipitation presented in the SAR (LES,

2005a) and the ISA Summary (LES, 2005b) and concludes that the information is acceptable

because recognized data sources, such as NOAA, were used.

3.3.1.1.3.6 Lightning and Thunderstorms

Section 1.3.3.3 of the SAR (LES, 2005a), and Section 3.2.3.4.5 of the ISA Summary

(LES, 2005b) describe the potential of thunderstorms and lightning strikes at the proposed

facility site. The applicant indicated thunderstorms occur every month and are most common in

spring and summer at the proposed facility site.

The applicant estimated the lightning strike frequency at the proposed facility site to be 1.36

flashes per year. The applicant stated in the ISA Summary (LES, 2005b) that the proposed

facility will be designed for lightning protection.

The staff reviewed the information about lightning and concludes that the lightning strike

frequency determined for the site is acceptable and appropriate. Staff further concludes that the

design approach proposed by the applicant to protect the proposed facility from lightning effects

is acceptable.

3.3.1.1.3.7 Sandstorms

Section 1.3.3.3 of the SAR (LES, 2005a) describes the potential of sandstorms at the proposed

facility site. Blowing sand and dust may occur occasionally. Large dust storms with the

potential of covering a large region are rare (DOE, 2003). Staff reviewed the information about

sandstorms presented in the SAR (LES, 2005a) and finds the information sufficient and

acceptable.

Staff concludes that, based on the individual parameter assessments discussed above, the

applicant’s material regarding meteorology presented in the SAR and ISA Summary is sufficient

and acceptable to use in determining appropriate mechanical and thermal loads for the safe

design of principal structures, systems, and equipment.

3.3.1.1.4 Hydrology

Site surface water and groundwater hydrology is discussed in Sections 1.3.4 of the SAR (LES,

2005a) and 3.2.4 of the ISA Summary (LES, 2005b). The applicant obtained hydrological data

principally from previous investigations conducted by WCS, which is located 1.6 km (1 mi) east

of the proposed site. WCS operates a hazardous chemical treatment and disposal facility. The

applicant performed a limited number of geotechnical studies that demonstrate that the WCS

data are applicable to the proposed site.

The proposed site contains no surface water and/or surface water drainage features, with

essentially all precipitation subject to either infiltration or evapotranspiration.

The applicant performed subsurface studies of the alluvial material that overlies the Chinle red

bed clays. These alluvial deposits are 9 to 15 m (30 to 60 ft) thick. The Chinle formation

consists of a low-permeability clay unit having a thickness of 323 to 333 m (1060 to 1092 ft). No
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perched water systems in the alluvial deposits were found, although one well produced water

samples, because of a limited groundwater occurrence.

The low permeability Chinle formation essentially isolates the deep and shallow groundwater

systems. Within the Chinle formation are three distinct groundwater systems, with no

interconnections. The first is a siltstone or silty sandstone unit with some saturation at 65 to 68

m (214 to 222 ft) below the surface. This unit is a low-permeability formation that does not yield

groundwater easily. The second unit is a saturated siltstone layer approximately 30.5 m (100 ft)

thick, at an elevation of 183 m (600 ft) below the surface. The third unit is the Santa Rosa

formation at 340 m (1115 ft) below the surface. The Santa Rosa unit is the first occurrence of a

well-defined aquifer system. However, this system is considered non-potable because of high

concentrations of dissolved solids.

At the quarry site, north of the proposed site, there are shallow groundwater occurrences. 

These shallow perched systems, however, are intermittent and limited and caused by a layer

caliche or caprock at the surface that in places is fractured and can lead to rapid infiltration of

precipitation forming the perched water system. Caprock, however, is not present at the

proposed site, and, therefore, it is not expected that significant perched water systems would be

produced.

Baker Spring is located about 1.6 km (1 mi) northeast of the proposed site. However, this

spring is intermittent and flows only after precipitation events.

Several localized shallow perched groundwater systems exist to the east of the proposed site

and are used to supply water pumped by windmills to tanks for grazing livestock. These

perched systems are located above the Chinle clays, but the volume of water produced is

limited.

Because of the lack of sufficient surface and groundwater supplies, the applicant will not make

withdrawals of groundwater at the site. Instead, the applicant is proposing to obtain water for

plant use from Eunice and Hobbs municipal supplies. These water supplies are obtained from

well fields near Hobbs, New Mexico. The applicant is also not proposing to inject water into

groundwater systems at the site.

Since there are no surface water bodies in the immediate vicinity of the site, flooding is not a

design objective. The only potential flooding at the plant would occur from intense local

precipitation events. Flood protection is provided by establishing building floor levels above the

calculated depth of ponded water caused by intense precipitation events (see Sections

3.3.2.1.3.4 and 3.3.2.1.3.5 of this SER).

The staff reviewed the applicant’s hydrological data and finds that it provides sufficient

information to assess site flooding hazards and ground- and surface water impacts.

3.3.1.1.5 Geology

3.3.1.1.5.1 Seismic Hazard - General
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Seismic hazards are discussed in Section 1.3.5 of the SAR (LES, 2005a); and Sections 3.2.5

and 3.2.6 of the ISA Summary (LES, 2005b).

The following areas concerning the seismic hazard applicable to the safety analysis and design

of the proposed facility were reviewed:

• Seismic source characterization; 

• Ground motion attenuation;

• Seismic hazard calculation;

• Development of site-specific spectra; and

• Surface faulting.

Seismic Source Characterization

Geological and Tectonic Settings

Section 3.2.5 of the ISA Summary (LES, 2005b) provides a description of the regional and local

geological and tectonic settings. The proposed facility site is located within the Central Basin

Platform area. The Central Basin Platform Area is situated between the Midland and Delaware

Basins, all of which are part of the Permian Basin, a 250-million-year-old structure. The

Permian Basin is a downward flexure of a large thickness of originally flat-lying bedded,

sedimentary rock. The base of the Permian Basin sediments extends to approximately 1525 m

(5000 ft) beneath the proposed facility site. The top of the Permian section is approximately 434

m (1425 ft) below ground surface. These sediments are overlain by sedimentary strata of the

Triassic Age Dockum Group. The upper formation of the Dockum Group is the Chinle

Formation, locally overlain by either the Tertiary Ogallala, Gatuña, or Antlers Formations, or

Quaternary alluvium. At the proposed facility site, geotechnical borings identified up to 0.6 m (2

ft) of loose eolian sand underlain by dense to very dense, fine- to medium-grained sand and

silty sand of the Gatuña Formation. The sands of the Gatuña Formation are locally cemented

with caliche. Beneath the Gatuña Formation, the Chinle claystone, a hard and highly plastic

clay, was encountered in geotechnical borings at depths from 10.7 to 12.2 m (35 to 40 ft).

As noted in the ISA Summary (LES, 2005b), the Southeast New Mexico–West Texas area is

presently structurally stable. The Laramide Orogeny (late Cretaceous to Early Tertiary time)

uplifted the region to its present elevation, and there has been no substantial tectonic activity

since this early Tertiary deformation. The Permian Basin has subsided slightly since the

Laramide Orogeny. However, this subsidence is believed to be a result of dissolution of the

Permian evaporite layers by groundwater or possibly compaction from oil and gas extraction. 

As stated in the ISA Summary (LES, 2005b), no active faults have been identified at the site. 

Faulting consists of geologically older subsurface faults in the Permian Basin subregion

related to the development of the Permian Basin and the Laramide Orogeny. The nearest

evidence of Quaternary faulting is 161 km (100 mi) west of the site, in the Basin and Range

tectonic province.

Historical Seismicity

Section 3.2.6.1 of the ISA Summary (LES, 2005b) summarizes the historical seismicity at the

proposed facility site. As stated in the ISA Summary (LES, 2005b), the assessment of historical
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seismicity included earthquakes in the region of interest known from felt or damage records and

from more recent instrumental records (since the early 1960s). The largest earthquake known

to occur within 322 km (200 mi) of the site was the August 16, 1931, earthquake near Valentine,

Texas. This earthquake had an estimated ML (Local Magnitude) of 6.0 to 6.4 and produced a

maximum epicentral intensity of VIII on the Modified Mercalli intensity scale. This earthquake

occurred approximately 237 km (147 mi) from the proposed site location. Within 80 km (50 mi)

of the site, the largest historical earthquake was a ML 5.0 event in 1992, approximately 16 km

(10 mi) southwest of the site. Other significant events between 322 km (200 mi) and 80 km (50

mi) of the proposed facility site ranged in ML from 4.0 to 5.7.

Earthquakes in the region of the proposed facility site include isolated and small clusters of lowto-moderate-magnitude events toward the Rio Grande Valley of New Mexico and in Texas,

southeast of the proposed site. According to the ISA Summary (LES, 2005b), no earthquakes

in the site region are known to be correlated to specific faults. An earthquake catalog based on

the historic seismicity in the region [322-km (200-mi) radius] was presented in the ISA Summary

(LES, 2005b). This catalog was composed of data from: the Advanced National Seismic

System (NCEDC, 2004); University of Texas Institute for Geophysics (UTIG, 2002); New Mexico

Tech Historical Catalog (NMIMT, 2003); and New Mexico Tech Regional catalogs. The catalog

identified a substantial cluster of seismic activity that has occurred on and near the Central

Basin Platform since the mid-1960s. It was suggested by DOE (DOE, 2003) and noted in the

ISA Summary (LES, 2005b) that Central Basin Platform earthquakes are not tectonic in origin

but instead are related to water injection and withdrawal resulting from secondary recovery

operations in oil fields in the Central Basin Platform area. The ISA Summary (LES, 2005b)

noted, however, the January 2, 1992, event was attributed to a tectonic origin because of its

determined focal depth of approximately 12 km (7 mi). This event was likely associated with

pre-existing zones of weakness within the crust that formed in the distant geologic past. These

zones of weakness are characterized by deeply buried and poorly characterized faults, some of

which accomplish a periodic release of strain that builds up continually in the North American

continental plate. At the proposed facility site, postulated earthquakes that could impact safe

operation of the proposed facility are associated with zones of crustal weakness in the Central

Basin Platform and the Basin and Range tectonic province.

The staff concludes that information concerning seismic source characterization presented in

the ISA Summary (LES, 2005b) is acceptable. The information provides a complete summary

of seismicity and potential fault and tectonic sources.

Ground Motion Attenuation

Details of ground motion attenuation functions used to compute the hazard are described in

Section 3.2.6.4.1 of the ISA Summary (LES, 2005b). Several attenuation models were used in

the ISA Summary. The Nuttli attenuation model developed by the U.S. Department of the Army,

Waterways Experiment Station (USDA, 1973) was primarily selected because it was used in the

DOE (DOE, 2003) seismic hazard assessment. The Toro, et al. (Toro, 1997) attenuation model

also was used in the hazard calculations for comparison.

The attenuation models used in the ISA Summary were applicable to locations within the

Central U.S.. The proposed facility site is located at 103° west longitude, slightly east of the

105° west longitude cutoff for Central and Eastern U.S. sites, as specified in Regulatory Guide

1.165 (NRC, 1997). In addition, Frankel, et al. (Frankel, 1996) specified attenuation zones for
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the U.S. in its hazard mapping project. The U.S. Geological Survey (USGS) boundary

separating the Western U.S. and the Central and Eastern U.S. attenuation zones also is located

at approximately 105° west longitude and slightly to the west of the proposed facility site. The

proposed facility site is thus situated within the area in which both the Central and Eastern U.S.

attenuation models are applicable, and the staff concludes that applicant’s evaluation of ground

motion attenuation is acceptable.

Seismic Hazard Calculation

A probabilistic seismic hazard analysis performed for the proposed facility site is discussed in

Section 3.2.6.4.2 of the ISA Summary (LES, 2005b). The method used to calculate the

probabilistic seismic hazard was that of McGuire (McGuire, 1976). The probabilistic seismic

hazard analysis incorporated seismic source zones from both local and distant seismic sources. 

Several alternative distant source zones were used in the hazard calculations. These distant

source zones included the Rio Grande rift and Basin and Range area sources and were taken

directly from DOE (DOE, 2003). Similarly, a suite of alternative local area source zones was

used to account for local seismicity, including the Central Basin Platform area source zone.

The total seismic hazard at the proposed facility site was the sum of ground motion effects from

all distant and local seismic sources. Twelve seismic hazard curves were developed for a

combination of various seismic source zones, attenuation models mentioned previously, bvalues, and upper bound magnitudes. As described in the ISA Summary (LES, 2005b), the

resultant seismic hazard curve was developed through a weighted average of the individual

curves. A total weight of 0.75 was given to the first seismicity model, which includes the Rio

Grande Rift and Basin and Range seismic source zones and the Central Basin Platform,

embedded within a seismic source zone defined by a 161-km (100-mi) radius from the site. A

combined weight of 0.25 was assigned to a second seismicity model that incorporated a seismic

source zone defined by a 322-km (200-mi) radius from the site and a third seismicity model that

incorporated a 161-km (100-mi) radius from the site. These two models did not incorporate the

Central Basin Platform and Rio Grande Rift and Basin and Range seismic source zones and,

therefore, were not as strongly weighted as the first seismicity model. The resulting 250- and

475-year return period peak horizontal ground accelerations were estimated at 0.024 and

0.036g. The respective 10,000- and 100,000-year return period peak horizontal ground

accelerations were estimated at 0.15 and 0.31g.

Comparison of the hazard results for the proposed facility site with USGS National Seismic

Hazard Mapping Project (USGS, 2004) suggests the 10,000-year return period hazard results

may underestimate ground motions at the site. For the 10,000-year return period, the USGS

estimates a peak ground acceleration of 0.40g, with the 0.2–5 second spectral acceleration at

0.97g. These values are 2.6 to 2.8 times higher than the respective values calculated by the

applicant. The applicant provided an evaluation of the USGS results (LES, 2004b). The

applicant concluded that USGS used overly conservative maximum magnitude and activity rate

information in the background source zone.

The staff agrees with the applicant’s assessment that the USGS hazard calculation used larger

maximum magnitudes and seismicity rates and is therefore more conservative. Additionally, the

USGS study was also not site specific because the USGS hazard calculation used a site

spacing of 0.1 degrees latitude and longitude for the Western U.S. and 0.2 degrees for the
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Central and Eastern U.S., which represents a much larger area. The applicant’s calculation is,

on the other hand, site specific and more representative of the site.

In addition, the seismic hazard calculated for facility site is similar to that calculated for the

nearby Waste Isolation Pilot Plant (DOE, 2003). The calculated 10,000-year return period peak

ground acceleration at the Waste Isolation Pilot Plant is slightly less than 0.15 g. Based on all

the information available, the staff concludes that the seismic hazard described in the ISA

Summary (LES, 2005b) is acceptable because it is based on a method that follows current

industry practice and includes available data. 

Development of Site-Specific Spectra

Information on the rock conditions hazard was used to obtain the response spectra at the

proposed facility site. The attenuation model (USDA, 1973) was considered to predict ground

motions at “firm rock” conditions, which are attributed to the Triassic Age claystones (Dockum

Ground including the Chinle Formation) underlying the proposed facility site. As determined by

5 geotechnical borings, above the Triassic claystones is a thickness of 10.7 to 12.2 m (35 to 40

ft) of sands of the Gatuña Formation and loose eolian sands that belong to a soil classification

of C (Dobry, 2000). As outlined in the ISA Summary (LES, 2005b), the firm rock uniform hazard

response spectra were transformed to soil conditions by multiplying the firm rock uniform hazard

response spectra by the appropriate soil-amplification factors specified by Dobry, et al (Dobry,

2000). Vertical component uniform hazard response spectra were determined to be a function

of frequency and were determined using the formulation outlined in NRC Regulatory Guide 1.60

(NRC, 1973). 

The design-basis earthquake for the proposed facility site was selected as the 10,000-year (1.0

× 10!4

 mean annual probability) earthquake. As indicated in the ISA Summary (LES, 2005b),

the applicant proposed the method outlined in DOE–STD–1020 (DOE, 2002) or

ASCE/Structural Engineering Institute (SEI) 43-05, “Seismic Design Criteria for Structures,

Systems, and Components in Nuclear Facilities” (ASCE/SEI, 2005) to demonstrate compliance

to a target performance goal of 1.0 × 10!5

 annual probability by designing to a seismic hazard of

1.0 × 10!4

 annual probability [for further discussion, see Section 3.3.3(C)(v) of this Safety

Evaluation Report (SER)]. The design response spectra for the proposed facility were based on

the 10,000-year uniform hazard response spectra described in the previous paragraph. The

horizontal and vertical design response spectra have standard response spectral shapes (at 5

percent damping) based on the Newmark and Hall amplification factors (NRC, 1978a). The

spectrum is anchored at the 10!4

 per year peak horizontal ground acceleration of 0.151g

determined from the weighted final seismic hazard curve. Amplification factors were then

applied using Dobry, et al. (Dobry, 2000) soil class C definition to account for the local soil

column.

The staff considers the method and results used to develop site-specific spectra acceptable

because they follow current industry practice and are consistent with methods used at other

nuclear facilities (e.g., DOE, 2002). Thus, staff conclude that the site-specific hazard and

response spectra are technically sound.

Surface Faulting
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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 (LES, 2005b), the nearest

recent faulting is located more than 161 km (100 mi) west of the site. Therefore, surface

faulting was not considered a credible disruptive event for the proposed facility.

Recently, a fault was discovered at the nearby Waste Control Specialists (WCS) site. However,

subsequent fault investigations revealed that the faulting is inactive because no faults exist in

formations younger than Triassic age (205 to 240 million years old) (LES, 2005a).

Geotechnical investigation indicated the soil beneath the proposed facility site is a layer of loose

eolian sand underlain by the Gatuña Formation (dense to very dense sand and silty sand). 

Below the Gatuña Formation is the Chinle claystone, a very hard, highly plastic clay. The

Chinle claystone was encountered at depths approximately 10.7 to 12.2 m (35 to 40 ft). For the

top 7.6 m (25 ft) of soils, the blow-count values ranged from 20 to 76. Beneath the 7.6-m (25-ft)

horizon, typical blow-count values were more than 60, with even larger blow-count values for

the Chinle claystone.

Based on the staff’s review of the applicant’s submittals on seismic hazard potential discussed

in the above sections, staff concludes that the potential hazards have been acceptably

addressed by the applicant.

3.3.1.1.5.2 Slope Stability

Section 1.3.1.2 of the SAR (LES, 2005a), and Section 3.2.1.1 of the ISA Summary (LES, 2005b)

describe the topography at the proposed facility site. The SAR (LES, 2005a) and ISA Summary

(LES, 2005b) indicated the site topography is relatively flat, with a gradual elevation increase

from southwest to northeast. The staff site visit on May 27–28, 2004 (NRC, 2004b), confirmed

the area at the proposed facility is relatively flat. Consequently, slope stability is not a safety

concern for this proposed facility.

3.3.1.1.5.3 Liquefaction

Liquefaction potential of soils beneath the proposed facility is discussed in Section 3.2.7.1 of the

ISA Summary (LES, 2005b). According to the ISA Summary (LES, 2005b), except a top layer

of loose sand (up to 0.6 m (2 ft)), the soils at the proposed site are dense to very dense and the

groundwater level is at least 30 m (98 ft) below ground surface. Consequently, the applicant

concluded the potential for liquefaction was remote (LES, 2005b, c).

The staff reviewed the geotechnical investigation information presented in the ISA Summary

(LES, 2005b, c) and concurs with the applicant that the potential for liquefaction of soils at the

site may not be a safety concern for the proposed facility. The applicant committed to perform

additional geotechnical investigations at the site to confirm that liquefaction is not a safety

concern for the proposed facility. Additional site testing will be evaluated in accordance with

NRC Regulatory Guide 1.198, “Procedures and Criteria for Assessing Seismic Soil Liquefaction

at Nuclear Power Plant Sites” (NRC, 2003b).

3.3.1.1.5.4 Settlement
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Settlement of foundations for the proposed facility is discussed in Section 3.2.7 of the ISA

Summary (LES, 2005b). In its ISA Summary, the applicant stated that only five borings were

drilled at the proposed facility site to determine the suitability of the site. The applicant

recognized the geotechnical results obtained from the five borings were not sufficient for final

design purposes. The applicant states that due consideration will be given to settlement and

differential settlement during final design, and final design details will be based on a more

comprehensive geothechnical investigation to be undertaken when additional project details

become available. Allowable soil bearing pressures will be evaluated in accordance with Naval

Facilities Engineering Command Design Manual NAVFAC DM-7.02, “Foundations and Earth

Structures” (NAVFAC, 1996). Building settlement analyses will be performed in accordance

with Naval Facilities Engineering Command Design Manual NAVFAC DM-7.01, “Soil

Mechanics” (NAVFAC, 1986) and Winterkorn and Fang, “Foundation Engineering Handbook”

(Winterkorn, 1975).

The staff reviewed the information presented concerning differential settlements and find the

applicant’s commitment to perform additional geotechnical investigations using acceptable

geotechnical standards for final facility design to be acceptable.

Conclusion

Staff finds that, based on the individual parameter assessments discussed above, the

applicant’s material regarding the geological information presented in the SAR and ISA

Summary is sufficient and acceptable to use in determining appropriate criteria for the safe

design of principal structures, systems, and equipment.

3.3.1.1.6 Site Description Conclusion

Based on the above, the Staff concludes that sufficient site information has been provided to

support identification of those factors that could affect safety, including related internal and

external hazards and associated accident sequences that could exceed the performance

requirements of 10 CFR 70.61. The staff concludes that ISA Summary site description meets

the requirements of 10 CFR 70.65(b)(1).

3.3.1.2 Facility Description

A description of the proposed facility arrangement, buildings and major components, and

structural design criteria is provided in SAR Chapter 1 (LES, 2005a) and ISA Summary Section

3.3 (LES, 2005b). The facility location and the distance from the boundary in all directions are

provided in ISA Summary Table 3.3-1 (LES 2005b). The distance to the nearest resident is

identified in ISA Summary Section 3.2 (LES 2005b), as about 4.3 km (2.6 mi) west of the site. 

The arrangement and location of buildings on the facility site, and identification of the site

boundary (controlled area) and controlled access area boundary, are shown in ISA Summary

Figure 3.3-1 (LES 2005b).

3.3.1.2.1 Buildings and Major Components
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This section contains a review of the civil structures of the proposed facility. According to the

SAR (LES, 2005a), the proposed facility consists of the following structures:

• Three Separations Building Modules

• Technical Services Building

• Cylinder Receipt and Dispatch Building

• Centrifuge Assembly Building

• Blending and Liquid Sampling Area

• Uranium Byproduct Cylinder Storage Pad

• Central Utilities Building

• Administration Building

• Visitor Center

• Site Security Building

Among these structures, the Separations Building Modules, Technical Services Building,

Cylinder Receipt and Dispatch Building, the centrifuge test facility in the Centrifuge Assembly

Building, and the Blending and Liquid Sampling Area are determined to be safety significant,

based on the ISA performed by the applicant (LES, 2005b). These structures are required to be

designed to withstand design-basis natural phenomena hazards and external hazards as

required by the baseline design criteria of 10 CFR 70.64(a)(2).

Each Separations Building Module includes two cascade halls. Each Separations Building

Module has a uranium hexafluoride handling area and a process services area. A Separations

Building Module is 170 m (557.75 ft) long, 67.9 m (222.75 ft) wide, and 13 m (42.7 ft) high, and

has 12,703 m2

 (137,025 ft2) of space. The Technical Services Building is a two-story building

with 9,192 m2

 (98,942 ft2) of space. The Cylinder Receipt and Dispatch Building is 246.2 m

(807.75 ft) long, 45.9 m (150.6 ft) wide, and 13 m (42.7 ft) high, and has 11,300 m2

 (121,638 ft2)

of space. The Centrifuge Assembly Building is 195.5 m (641.4 ft) long, 50.9 m (167 ft) wide,

and 11 to 16 m (36.1 to 52.5 ft) high, and has 11,364 m2

 (122,322 ft2) of space. The Blending

and Liquid Sampling Area is 33.5 m (109.9 ft) long, 45.9 m (150.6 ft) wide, and 10 m (32.8 ft)

high, and has 1,538 m2

 (16,555 ft2) of space.

3.3.1.2.2 Structural Design Criteria

Structural design criteria for the proposed facility is discussed in Section 3.3 of the SAR (LES,

2005a) and Section 3.3.2.1 of the ISA Summary (LES, 2005b). Specifically, structural design

loads for the proposed facility are discussed in Section 3.3.2.2 of the ISA Summary (LES,

2005b).

3.3.1.2.2.1 Codes and Standards

A list of the codes and standards for the structural design of the proposed facility are provided in

Section 3.3.2.1 of the SAR (LES, 2005a). These codes and standards included guidance for

general structural design, concrete design, precast concrete design, steel construction, and

testing and material selections.

Staff reviewed the cited codes and standards to be used for the design and construction of the

proposed facility structures and conclude that the cited codes and standards conform with

standard engineering practice and are reasonable and acceptable.
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3.3.1.2.2.2 Structural Design Loads

Straight Wind Load

ISA Summary Section 3.3.2.2.1, (LES, 2005b), identifies the design-basis straight-line wind as

252 km/h (157 mph) for all safety-significant structures. This design-basis wind is characterized

as a 100,000-year return period wind. The applicant will use American Society of Civil

Engineers (ASCE) ASCE 7-98 (ASCE, 1998) for the determination of wind loads and the design

for wind loads for all safety-significant structures and components exposed to winds. For

structures that are not safety significant, the design-basis wind speed [130 km/h (80 mph)] will

be that of a 50-year return period, and the Uniform Building Code (ICBO, 1997) will be followed. 

Staff concludes that the method used to calculate wind loads is based on an acceptable

standard.

Because the design-basis wind is characterized as a 100,000-year wind, the event sequences

associated with high winds are highly unlikely events based on the definition given in ISA

Summary Section 3.1, “General ISA Information” (LES 2005b) (see SER Section 3.3.3.c.1.3 for

the definition of highly unlikely).

Tornado Wind Load

The values of design-basis tornado-related parameters are listed as follows:

Design-Basis Wind Speed: 302 km/h (188 mph)

Radius of Damaging Winds: 130 m (425 ft)

Atmospheric Pressure Change: !3.83 kg/m2

 (!80 lbs/ft2)

Rate of Atmospheric Pressure Change: !1.44 kg/m2

/s (!30 lbs/ft2/s)

The tornado design-basis is characterized as a 100,000-year tornado, which is the equivalent of

a “F-3" tornado on the Fujita Tornado Scale. The applicant concluded a tornado was a credible

event. This probability, however, meets the definition of highly unlikely. Therefore, the potential

consequences for an event of this magnitude or greater do not have to be determined because

the event sequence is highly unlikely, in accordance with 10 CFR 70.61(b).

In ISA Summary Sections 3.3.2.1 and 3.3.2.2, the applicant indicated that ASCE 7-98 (ASCE,

1998) will be used for the determination of tornado wind loads and the design for tornado wind

loads for all safety-significant structures and components exposed to tornado winds. Use of

ASCE 7-98 (ASCE, 1998) for tornado load calculation and design of structures and components

for tornado loads is acceptable to the staff.

Tornado-Generated Missile Loads

Three types of missiles associated with the design-basis tornado were selected for design of

safety-significant structures. As indicated in ISA Summary Sections 3.2.3.4.1 and 3.3.2.2.3.2

(LES, 2005b), these missiles included (I) a 6.8-kg (15-lb), 10.2 × 30.5-cm (2 × 4-in) timber

plank, (ii) 34-kg (75-lb), 7.6-cm (3-in)-diameter steel pipe, and (iii) 1,361-kg (3,000-lb)

automobile. The associated vertical and horizontal impact velocities of each missile also were

provided in the same sections of the ISA Summary.
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The tornado-generated missile effects on structures will be considered in two aspects. For the

localized impact, depth of penetration and scabbing thickness will be determined for reinforced

concrete targets. The applicant will use the formulas provided in “Structural Analysis and

Design of Nuclear Plant Facilities” (ASCE, 19080) for depth of penetration and scabbing

thickness calculations. Staff finds the use of these formulas acceptable because these formulas

are commonly used for assessing local penetration and scabbing effects. The applicant will use

the requirements in Section C.7.2.2 of the American Concrete Institute (ACI) ACI 349-90 (ACI,

1990) for determining minimum concrete thickness to resist hard missiles such as steel pipes

and the requirements in Section C.7.2.3 of ACI 349-90 (ACI, 1990) to check against punching

shear resulting from soft missiles such as timbers. The use of ACI 349-90 (ACI, 1990) is

acceptable to staff because this standard is developed specifically for nuclear safety-related

concrete structures. For steel targets, the formula provided in “Structural Analysis and Design

of Nuclear Plant Facilities” (ASCE, 1980) will be used to estimate the perforation thickness. The

requirement in this document will be used to establish the required steel thickness. Staff

considers the use of this ASCE document acceptable because it is a design code for nuclear

plant facilities.

For design consideration of tornado-generated missile effects on overall structural response, all

missile momentum is assumed to transfer into the target during impact, and the target is

assumed to behave elasto-plastically. An equivalent static load acting at the impact area will be

calculated using the procedure outlined in “Structural Analysis and Design of Nuclear Plant

Facilities” (ASCE, 1980). This equivalent will be used in combination with other design loads

using conventional design methods. The applicant may also employ other formulations such as

presented in ACI 349-90 (ACI, 1990) for design to resist the overall effects of missile impact. 

Staff find the use of either of these methods acceptable because they represent standards and

codes for nuclear facilities.

Flood Design

The grade level of the proposed facility is above the maximum foreseeable flood level (LES,

2005a). The only potential flood may result from intense local precipitation. Two floodprotection design features will be used. First, the floor level of the proposed facility will be 0.15

m (0.5 ft) higher than the finished grade elevation, which will be at least 0.45 m (1.5 ft) above all

roads. Second, an earth berm and intercept trench will be constructed uphill of the building

structures to divert water flow. NRC staff has concluded that the flood-protection design

features are adequate to protect building structures from water damage.

Seismic Loads

Development of a design-basis earthquake for the proposed facility site is discussed in Section

3.2.6 of the ISA Summary (LES, 2005b). The site-specific, design-basis earthquake is

discussed in ISA Summary Sections 3.2.6.5 and 3.3.2.2.5.2 (LES, 2005b). The staff review and

acceptance of the selected design-basis ground acceleration are discussed in

Section 3.3.1.1.7.a of this SER.

The design-basis earthquake for the proposed facility is based on an earthquake with a return

period of 10,000 years (10!4

 annual probability). The corresponding peak horizontal and vertical

ground accelerations were 0.15g. In ISA Summary Section 3.3.2.2.5.1 (LES, 2005b), the

applicant indicated the seismic design loads will be calculated using the method outlined in
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Chapter 16, Division IV, of the Uniform Building Code (ICBO, 1997) for all buildings and

structures, including items such as equipment supports, for the proposed facility. The proposed

facility site is located in seismic zone 1 based on International Conference of Building Officials

(ICBO, 1997). In considering soil amplification, suggested soil amplification factors for Soil

Type C defined by Dobry, et al. (Dobry, 2000) will be used. These amplification factors are

listed in Section 3.2.6.4.3 of the ISA Summary (LES, 2005b). The applicant will verify the

appropriateness of the soil amplification factors for the proposed facility during final design. The

staff considers this method acceptable because it follows modern practices and is consistent

with methods used at other nuclear facilities [e.g., DOE–STD–1020–2002 (DOE, 2002)].

The applicant states that structures designed to a seismic hazard of 10!4

 annual probability can

meet the performance goal of 10!5

 annual probability so the seismic hazards for structures

relied on for safety satisfy the definition of highly unlikely. As a result, according to

10 CFR 70.61(b), the potential consequences resulting from seismic hazards do not have to be

assessed in the ISA. This additional performance may be achieved because of conservatism in

the design including conservative specifications of material strength and elastic design

approach. The applicant will use either the method outlined in DOE–STD–1020–2002

(DOE, 2002) or ASCE Standard Seismic Design Criteria (ASCE, 2003) for such demonstration. 

Specifically, the applicant intends to show that even though the design code allowables are

exceeded for the targeted structural performance goal, the ultimate capability of the design is

not exceeded. Confirmatory seismic performance calculations for structures relied on for safety

will be conducted during detailed design.

Staff reviewed the approach proposed for seismic design-load calculation and concludes that

the method presented in the Uniform Building Code (UBC) (ICBO, 1997) is acceptable for

determining seismic design loads because the UBC (ICBO, 1997) was developed with

emphasis on earthquake effects. Staff also concludes that the applicant’s approach in

demonstrating that they meet the performance goal of 10!5

 annual probability is reasonable

because of the conservatism in the conventional design approach.

Snow Loads

ISA Summary Section 3.2.3 (LES, 2005b) describes the design basis values for snow or ice

loads. According to the SAR (LES, 2005a), the design-basis snow load for the proposed facility

was developed based on the method provided in the NRC “Site Analysis Branch Position for

Winter Precipitation Loads” (NRC, 1975). The design-basis snow load was a combination of

100-year snowfall load at the site and the weight of the 48-hour probable maximum winter

precipitation for the month corresponding to the selected snowfall.

According to Figure 7-1 of American Society of Civil Engineers (ASCE) ASCE 7-98 (ASCE,

1998), the ground snow load with a 2-percent annual probability of being exceeded (i.e., a 50-

year mean recurrence interval) at the proposed facility site was approximately 0.48 kPa (10 psf). 

The applicant used this value to determine the 100-year return period snow load using the

method described in ASCE 7-98 (ASCE, 1998). The 100-year return period snow load was

determined to be 0.56 kPa (12 psf). The applicant (Harper, 2003c) estimated the 48-hour

probable maximum winter precipitation using the method described in Hydrometeorological

Report No. 33 (USWB, 1956). The 48-hour probable maximum winter precipitation was

determined to be 48.3 cm (19.0 in.), which corresponds to a loading of 0.95 kPa (19.8 psf). The

design-basis ground snow load was 1.53 kPa (32 psf). The design-basis snow load for
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buildings not safety significant will be determined in accordance with the UBC (ICBO, 1997)

based on a mean return period of 50 years.

The staff reviewed the information about design-basis snow load presented in the SAR (LES,

2005a) and the ISA Summary (LES, 2005b) and concludes that the information is acceptable

and appropriate because acceptable standards and acceptable methods were used to

determine the 100-year return period snow load and the 48-hour probable maximum winter

precipitation.

Rain Loads

In Section 3.3.2.2.6.2 of the ISA Summary (LES 2005b), the applicant states the rain loads on

roofs may be related to a 100-year return period rainfall and a localized intense rainfall. This

100-year return period rainfall was established based on a storm with a duration of 1 hour. 

According to the International Plumbing Code (ICC, 2000), the rainfall rate of this storm in

Hobbs, New Mexico, was 7.6 cm/h (3 in/h). Staff concludes that use of the International

Plumbing Code (ICC, 2000) in determining the 100-year return period rainfall for the site is

acceptable. For the 100-year return period rainfall, the roofs will be designed, assuming the

primary roof drains are blocked, to prevent water from accumulating in excess of the normal roof

design live load.

For the localized intense rainfall, the rain loads on roofs assuming a total blockage of the roof

drainage system will be accounted for as an individual load. This rain load is for safetysignificant structures only. Including a rain load determined assuming loss of roof drainage

capacity in the design of roofs for safety-significant structures is reasonable and acceptable

to the staff.

Process and Equipment-Derived Loads

The process and equipment-derived loads for the proposed facility include equipment; piping;

heating, ventilation, and air conditioning; and electric tray and conduit loads. Equipment heavier

than 454 kg (1,000 lb) will be treated as individual dead loads. Weights for other equipment will

be accounted for as appropriate uniform dead loads in the particular area.

Piping loads were estimated at 2.39 kg/m2

 (50 lbs/ft2) based on combined dead and live loads. 

Heating, ventilation, and air conditioning duct loads were 1.44 kg/m2

 (30 lbs/ft2) and were

estimated based on combined dead and live loads. Electric tray and conduit loads were

104 kg/m (70 lb/ft) of tray and conduit and a 91-kg (200-lb) concentrated load at midspan of

the tray. The electric tray and conduit loads were estimated based on combined dead and

live loads.

The applicant states (LES, 2005a) that the equipment; piping; heating, ventilation, and air

conditioning; and electric tray and conduit loads provided in the ISA Summary represent the

minimum loads expected to be used. These minimum loads will be assumed as live loads and

treated as such in load combinations. During final design, individual loads (dead versus live

loads) will be identified and accounted for in the design accordingly by applying appropriate load

factors defined in the load combinations. The resulting actual load will be compared to the load

obtained using the minimum load value treated as a live load. The larger of the two loads will

be used for design. The staff concludes that the approach used for considering loads created
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by equipment; piping; heating, ventilation, and air conditioning; and electric trays and conduit in

the load combination is conservative and acceptable.

Settlement-Induced Loads

Settlement of foundations for the proposed facility is discussed in Section 3.2.7 of the ISA

Summary (LES, 2005b). In its ISA Summary, the applicant stated that only five borings were

drilled at the proposed facility site to determine the suitability of the site. The applicant

recognized the geotechnical results obtained from the five borings were not sufficient for final

design purposes. The applicant states in SAR Chapter 3 (LES, 2005a) that additional

geotechnical investigations will be performed to obtain further data for the final design of the

proposed facility. Allowable soil bearing pressures will be evaluated in accordance with Naval

Facilities Engineering Command Design Manual NAVFAC DM-7.02, “Foundations and Earth

Structures” (NAVFAC, 1996). Building settlement analyses will be performed in accordance

with Naval Facilities Engineering Command Design Manual NAVFAC DM-7.01, “Soil

Mechanics” (NAVFAC, 1986) and Winterkorn and Fang, “Foundation Engineering Handbook,”

(Winterkorn, 1975).

NRC reviewed the information presented concerning differential settlements and concludes that

the applicant’s commitment to perform additional geotechnical investigations using acceptable

geotechnical standards for final facility design is acceptable.

Explosion Overpressure

In assessing potential hazards of propane truck accidents and a nearby natural gas pipeline

explosion, the applicant (Snooks, 2003; Thomson, 2004) used the assumption that the safetysignificant structures of the proposed facility would be designed to withstand 6.9 kPa (1 psi)

overpressure. This design-basis overpressure was included in Section 3.3.2.2.8.2 of the ISA

Summary (LES, 2005a) for design of safety-significant structures. Staff concludes that treating

the explosion overpressure as an extreme environmental load in the load combination

is acceptable.

Load Combinations

The design loads considered in the load combination include two parts: normal and extreme

environmental loads. The normal loads were listed in Section 3.3.2.2.8.1 of the ISA Summary

and the extreme environmental loads were listed in Section 3.3.2.2.8.2 (LES, 2005b).

In ISA Summary Section 3.3.2.2.8 (LES, 2005b), the applicant states that all concrete structures

will be designed using the ACI strength method (ACI, 1999). Load combinations for the design

of safety-significant concrete structures will follow those provided in ACI 349-90 (ACI, 1990),

while load combinations for other structures will follow ASCE 7-98 (ASCE, 1998). Specific load

combinations for both safety-significant and non-safety-significant concrete structures and

components are listed in Section 3.3.2.2.8.3 of the ISA Summary (LES, 2005b). The rain load

caused by localized intense rainfall will be evaluated using the load combination A.10 in Section

3.3.2.2.8.3 by replacing the design-basis tornado with this rain load (LES, 2005b).
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All structural steel will be designed using the American Institute of Steel Construction (AISC)

Allowable Stress Method (AISC, 1989). Specific load combinations for design of structure steel

are listed in Section 3.3.2.2.8.3 of the ISA Summary (LES, 2005b).

The load combinations for foundations are provided in Section 3.3.2.2.8.3, “Combined Load

Applications,” of the ISA Summary (LES, 2005b). The minimum factor of safety of foundations

against overturning was 1.5 and sliding was 2.0.

Staff reviewed the information presented in the ISA Summary regarding the design loads and

approaches used to develop seismic design loads for the safety-significant structures and

components and find these design loads were developed based on site characteristics. Staff

concludes that the approaches for determining the design loads followed acceptable codes and

standards. Staff also concludes that the approach proposed by the applicant to demonstrate

the structures designed to a seismic hazard of 10!4

 annual probability, meeting the performance

goal of 10!5

 annual probability, is acceptable because the conventional design approach

contains adequate conservatism. The design loads and the approaches described satisfy the

requirements of 10 CFR 70.64(a)(2), 70.64(a)(4), and 70.65(b)(2).

3.3.1.2.2.3 Foundations

Foundations will be shallow concrete footings. The applicant anticipated the footings for all

safety-significant buildings will be founded on firm and dense sandy soils. The allowable

bearing pressure is 335 kPa (7,000 lbs/ft2

). In areas where the footings bear in existing or new

fill areas, the allowable bearing pressure is 144 kPa (3,000 lbs/ft2

). The allowable pressure may

be higher in areas where the fill material is entirely rock.

3.3.1.2.3 Facility Description and Design Basis Conclusion 

The staff concludes that sufficient information has been provided describing those areas that

could affect safety, including identification of the controlled area boundaries. The facility

description is adequate to allow identification of the general features, including internal and

external hazards and associated accident sequences that could exceed the performance criteria

of 10 CFR 70.61. The ISA Summary facility description meets the requirements of 10 CFR

70.65(b)(2). The staff also concludes that appropriate design loads and associated codes and

standards were identified to meet 10 CFR 70.64(a)(2) and (4). 

3.3.1.3 Processes

The enrichment processes and equipment analyzed as part of the ISA process are described in

the ISA Summary, Section 3.4 (LES, 2005b). The enrichment systems are comprised of the

Uranium Hexafluroide (UF6) Feed System, Cascade System, Product Take-off System, and

Tails Take-off System. Major support systems include the Product Blending System, Product

Liquid Sampling System, and Contingency Dump System. Systems used to support the

enrichment process and the handling of UF6 are the Gaseous Effluent Vent System (GEVS),

Centrifuge Test Facility and Centrifuge Post Mortem Facility, and Material Handling. A

subsection containing information pertaining to the functional description, major components,

design description, interfaces, design and safety features, operating limits, instrumentation, and

IROFS is provided for each of the 10 enrichment and supporting systems.
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The process function is to enrich the amount of the 235U isotope in UF6 from 0.711 weight

percent up to a maximum of 5.0 weight percent through a mechanical centrifuge separation

process. Naturally occurring uranium will be received from a conversion facility in the form of

UF6 shipped in Type 48Y or 48X cylinders qualified to American National Standard Institute

(ANSI) Standard N14.1 (ANSI, 1995). The UF6 will be in a solid state under vacuum. The UF6

feed system will heat the cylinder to 53°C (127°F) to sublime the UF6 into a gas. The feed

purification system removes light gas components from the feed to a specified level prior to

admittance to the cascade in order to protect the centrifuges and enhance efficiency. At the

feed purification Low Temperature Take-off Stations, Type 48X or 48Y cylinders are cooled to -

25°C (-13°F). Gaseous UF6 enters the cylinders and desublimes into the solid phase. UF6 cold

traps, aluminum oxide (Al2O3) traps, and vacuum pumps are used to transfer residual light gas

to the GEVS. The traps remove any UF6 or hydrogen fluoride (HF) from the effluent stream. 

From purification, UF6 is transferred to the cascades where enrichment occurs. The cascades

are operated under a significant vacuum (about 65 mbar or 0.09 psia) to assure the UF6 does

not desublime back into a solid state at ambient temperatures.

The cascade systems separates the UF6 feed stream into a product stream and a tails stream. 

The product stream is routed to the Product Take-off System, where it is transported by vacuum

pumps to the Product Low Temperature Take-off Station. These stations are operated at -25°C

(-13°F) and the UF6 desublimes into the solid form inside a Type 30B or 48Y cylinder. The Type

30B cylinder is used for final product, while the Type 48Y cylinder is used for future blending

operations. Any light gas impurities are purged through cold traps followed by product vent

vacuum pump/chemical trap sets.

The Tails Take-off System withdraws the depleted UF6 stream and provides a means to

withdraw UF6 from the centrifuge cascades under abnormal conditions. The UF6 is routed to a

Low Temperature Take-off Station operated at -25°C (-13°F). The gaseous UF6 is desublimed

into a solid form inside a Type 48Y cylinder.

The Product Blending system is used to provide a specific enrichment of 235U by blending UF6

at two different enrichment levels. The donor stations can handle Type 30B and 48Y product

cylinders. The UF6 is sublimed back into a gas and transported to a Blending Receiver Station

containing an empty Type 30B cylinder operated at -25°C (-13°F), where the UF6 desublimes

back into a solid form.

The only system at the facility that changes solid UF6 into liquid UF6 is the Product Liquid

Sampling System. A filled Type 30B cylinder is placed in an autoclave that is heated to 70°C

(158°F) by electric heaters. When the pressure reaches +2.5 bar (36.3 psia), the temperature is

stabilized for about 16 hours to allow homogenization prior to sampling. The sample bottles are

connected to the cylinder via a header, all located within the confines of the autoclave. The

main safety feature of the autoclave is to provide a secondary confinement barrier in the event

of a UF6 leak. All Type 30B cylinders are required to meet ANSI N14.1 (ANSI, 1995)

requirements, which include a cylinder design pressure of 1380 kPa (200 psi) and testing to

2760 kPa (400 psi).

Section 3.5 (LES, 2005b) of the ISA Summary emphasizes capacities, redundancies, and other

provisions for coping with routine and non-routine events. The system descriptions include

functional requirements, design capacities, system interfaces, and descriptions of major

components. Operational characteristics and safety considerations are also described. The
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applicant asserts that the health and safety of the public are protected such that a failure or

inadvertent operation of any utility system would not result in the release of hazardous

quantities of chemicals or radiation. These systems include the Building Ventilation System,

Electrical System, Compressed Air System, Water Supply, Cooling Water Systems, Septic

Systems, Communication and Alarm Annunciation Systems, Fire Protection, Control Systems,

Standby Diesel Generator System, Nitrogen System, Liquid Effluent Collection and Treatment

System, Solid Waste Collection System, Decontamination Workshop, Fomblin Oil Recovery

System, Laundry System, Ventilated Room, Chemical Laboratory, and the Cylinder Preparation

Room. Many of these systems do not contain IROFS. However, codes and standards to which

these systems will be designed are identified in Table 3.3-8 of the SAR (LES, 2005a) and Table

3.5-1 of the ISA Summary (LES, 2005b).

The applicant did not identify any essential utility service systems that are required to support

the IROFS safety function. However, upon completion of the design, the IROFS boundaries will

be defined by using LES procedure, “IROFS Boundary Definition” (LES, 2005a). Any essential

utility service systems that are required to support the safety function of the IROFS will fall

within the boundary of the IROFS per above procedure. The applicant states that if any

essential utility service systems are within the IROFS boundary per final design of the IROFS,

then these essential utility service systems will be designed to withstand environmental stresses

caused by environmental and dynamic service conditions under which failure could prevent the

satisfactory accomplishment of the IROFS safety function. Also, if an essential utility is

identified in the future, its continued operation will be addressed at that time through the

applicant’s design change process.

Instrumentation and Control (I&C)

Process schematics (LES, 2005a) show instrumentation component functions in sufficient detail

to understand the I&C systems at a functional level, and the results of the ISA based on those

functional descriptions. The detailed information on the physical components will be developed

when the final design is complete. Once the final design is complete, the applicant will maintain

I&C system design documentation onsite as part of the ISA (see Section 3.8 of the ISA

Summary (LES, 2005b)). The design documentation will include hardware design details,

identification of essential support utilities, operating ranges and limits for measured process

variables, and safety limits and safety margins, as applicable. The applicant has also

committed to following good engineering design practices established by the ISA and current

applicable codes and standards (NRC, 2002a and LES, 2005a).

Measured process variable operating ranges and limits (i.e., trip setpoints) are discussed in

Section 3.4 of the ISA Summary (LES, 2005b). Section 3.8.1 of the ISA Summary addresses

IROFS. For hardware IROFS with instrumentation providing automatic prevention or mitigation

actions, setpoint calculations will be performed in accordance with a setpoint method consistent

with the applicable guidance provided in Regulatory Guide 1.105, “Setpoints for Safety-Related

Instrumentation,” Revision 3, dated December 1999 (NRC, 1999a).

The Process Control System (PCS) is essentially a distributed control system interconnecting

the various major components over a plant control network. The PCS and its software will not

have any direct, hardwired interface to systems or components responsible for the actuation,

control, or reset of safety systems and components (LES, 2005a and NRC, 2002a).
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Since the proposed design, which is not complete at the time of this review, does not include

IROFS that use software, firmware, microcode, PLCs, and/or any digital device, including

hardware devices which implement data communication protocols, the staff finds the current

applicant commitments satisfactory. However, should the applicant choose to implement design

changes to include any of the preceding features, prior NRC approval will be necessary (see

License Condition No.1 in Section 3.4 of this SER).

In the facility SAR, the applicant provided preliminary design basis information for I&C systems

that it identified as IROFS for the facility. The design information is at the system functional

level. Individual components and vendors had not yet been selected. Based on the staff’s

review of the SAR, supporting information provided by the applicant, and the applicant’s

commitments to the industry standards and guidance cited in the sections above for I&C

systems, the staff finds that the preliminary design meets the requirements of 10 CFR 70.61 and

70.64(a)(10). 

Given that these conclusions are based on preliminary design information and the possibility

that the applicant may choose to implement design changes as discussed in the previous

section on I&C, the staff is imposing a license condition to ensure that the final design is

adequate and acceptable to the staff. Specifically, the following condition will be included in the

license:

“Currently, there are no IROFS that have been specified as using software, firmware,

microcode, PLCs, and/or any digital device, including hardware devices which

implement data communication protocols (such as fieldbus devices and Local Area

Network controllers), etc. Should the design of any IROFS be changed to include any of

the preceding features, the licensee shall obtain Commission approval prior to

implementing the change(s). The licensee’s design change(s) shall comply with

accepted best practices in software and hardware engineering, including software

quality assurance controls as discussed in the Quality Assurance Program Description

(LES, 2005a) throughout the development process and the applicable guidance of the

following industry standards and regulatory guides as specified in SAR Chapter 3 (LES,

2005a):

1. American Society of Mechanical Engineers (ASME) NQA-1-1994, Part II, subpart

Part 2.7, “Quality Assurance Requirements of Computer Software for Nuclear

Facility Applications,” as revised by NQA-1a-1995 Addenda of NQA-1-1994 and

ASME NQA-1-1994, Part 1, Supplement 11S-2, “Supplementary Requirements

for Computer Program Testing.” (Refer to SAR Chapter 11, Appendix A, Section

3.)

2. Electric Power Research Institute (EPRI) NP-5652, “Guideline for the Utilization

of Commercial Grade Items in Nuclear Safety Grade Applications,” June 1988.

3. EPRI Topical Report (TR) -102323, “Guidelines for Electromagnetic Interference

Testing in Power Plants,” Revision 1, December 1996.

4. EPRI TR-106439, “Guideline on Evaluation and Acceptance of Commercial

Grade Digital Equipment for Nuclear Safety Applications,” October 1996.
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5. Regulatory Guide 1.152, “Criteria for Digital Computers in Safety Systems in

Nuclear Power Plants,” Revision 1, January 1996.

6. Regulatory Guide 1.168, “Verification, Validation, Reviews, and Audits for Digital

Software Used in Safety Systems of Nuclear Power Plants,” Revision 1, February

2004.

7. Regulatory Guide 1.169, “Configuration Management Plans for Digital Computer

Software Used in Safety Systems of Nuclear Power Plants,” September 1997.

8. Regulatory Guide 1.170, “Software Test Documentation for Digital Computer

Software Used in Safety Systems of Nuclear Power Plants,” September 1997.

9. Regulatory Guide 1.172, “Software Requirements Specifications for Digital

Computer Software Used in Safety Systems of Nuclear Power Plants,”

September 1997.

10. Regulatory Guide 1.173, “Developing Software Life Cycle Processes for Digital

Computer Software Used in Safety Systems of Nuclear Power Plants,”

September 1997.

If any above changes result in IROFS requiring operator actions, the licensee shall

conduct a human factors engineering review of the human-system interfaces using the

applicable guidance in NUREG-0700, “Human-System Interface Design Review

Guidelines,” Revision 2, dated May 2002 (NRC, 2002d), and NUREG-0711, “Human

Factors Engineering Program Review Model,” Revision 2, dated February 2004.” 

Contingent on the above provision, the staff finds the above I&C safety-related commitments

acceptable.

Electrical

In ISA Summary, Section 3.5.2 (LES, 2005b), the applicant states that the overall electrical

power system is designed with a high level of redundancy to maintain a reliable power supply to

the process equipment for investment protection. Total loss of electrical power does not have

any safety implications. Furthermore, the applicant states in ISA Summary Section 3.8.1 (LES,

2005b), that IROFS systems will be designed to be fail-safe.

The applicant also describes in Section 3.5 of this ISA Summary (LES, 2005b) how the various

support functions would be performed during a loss of power. Included in the discussion were

descriptions of the Standby Diesel Generators, Uninterruptible Power Supplies, and defense-indepth approach. The applicant will provide two standby diesel generators to power short break

loads and uninterruptible power to process instruments, plant control system, emergency

lighting, site communications system, and environmental monitoring to allow safe shutdown of

the facility. The applicant will mitigate the effects of hydrogen ignition in areas containing

batteries as part of final design through implementation of the National Fire Protection

Association (NFPA) 70 E, “Standard for Electrical Safety Requirements for Employee

Workplaces,” and American National Standards Institute (ANSI) C2, “National Electrical Safety

Code,” if ventilation control of hydrogen gas is required.
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The staff concludes that the descriptions of the overall electrical power system is adequate to

allow evaluation of the completeness of the hazard and accident identification tasks and the

likelihood and consequences of accidents. The descriptions covered basic process function

and theory, and the general arrangement, function, and operation of the major components in

the process. Process design, equipment, and instrumentation were sufficiently discussed.

Human Factors

The staff’s analysis included the process used to conduct the human factors engineering review

of the Communication and Alarm Annunciation System, the Central Control Room, and the

Criticality Accident Alarm System (CAAS) as it applies to IROFS requiring operator actions.

The applicant describes the Control Room in ISA Summary, Section 3.3.1.2.2.17 (LES, 2005b),

the Central Control System in ISA Summary Section 3.5.9.2.1 (LES, 2005b), the

Communication and Alarm Annunciation System in ISA Summary, Section 3.5.7 (LES, 2005b),

and the CAAS in SAR, Section 5.3 (LES, 2005a) and ISA Summary, Section 3.1.5 (LES,

2005b).

The applicant described (LES, 2005a) how it used the design, maintenance, and operating

experience of Urenco to derive best practices for all elements of the plant. Incorporated

implicitly in the Core Plant Design (CPD) are all the human factors engineering enhancements

from over 30 years of experience. The Urenco SP5 plant design, on which the proposed facility

is based, is based on this CPD. Additionally, every element of the design is subject to HAZOP

analysis and design review. Operations experienced personnel are mandatory members during

such HAZOPs and design reviews. Urenco design review guidelines include design review

topics of functionality, operability, maintenance, layout, and orientation. Urenco engineering

design safety principles also address the human factors engineering issues. Specifically these

principles state that the design of all interfaces between operating personnel and the plant

should follow good human factors and ergonomics practice. These principles also note that

analysis of the safety function tasks requires determination of the demands on personnel in

order to evaluate the feasibility of the tasks and provide input to the design interfaces. The

design of tasks and equipment are fully compatible with training arrangements for operations

personnel, proposing staffing levels, and the development of operating procedures.

In ISA Summary Section 3.8.1 (LES, 2005b), the applicant states that for those IROFS requiring

operator actions, a human factors engineering review of the human-system interfaces will be

conducted using the applicable guidance in NUREG-0700, “Human-System Interface Design

Review Guidelines,” Revision 2, dated May 2002 (NRC, 2002d), and NUREG-0711, “Human

Factors Engineering Program Review Model,” Revision 2, Dated February 2004 (NRC, 2004a).

The staff concludes that appropriate human factors guidance has been identified for conducting

reviews of the human-system interfaces for those IROFS requiring operator actions, and the

engineering review of the Communication and Alarm Annunciation System, the Central Control

Room, and the CAAS.

Process Conclusion

The operating ranges and limits for measured variables for processes that are controlled by

IROFS are included in the ISA Summary or the classified portion of the ISA. The staff

concludes that this information is sufficient to permit an understanding of the theory of operation
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and assess compliance with the performance requirements of 10 CFR 70.61. The staff

concludes that an adequate description was provided for each process analyzed in the ISA to

understand the theory of operation; and for each process, to allow evaluation of the

completeness of the hazards that were identified and the accident sequences identified. 

3.3.2 Baseline Design Criteria and Defense-In-Depth

ISA Summary, Section 3.1.7 (LES, 2005b), addresses compliance with 10 CFR 70.64. The ISA

accident sequences for credible high and intermediate consequence events define the design

basis events. The IROFS and safety parameter limits for these events are available in the ISA

documentation. Together, the IROFS and safety parameter limits ensure that the associated

baseline design criteria (BDC) are met. Specific details on how the facility design or operation

conforms to the BDC are located in the individual sections of the SAR. Staff review of external

events, including natural phenomena hazards, is in SER Section 3.3.1.1.

The applicant will ensure that the structures, systems, and components (SSCs) that are

determined to have safety significance are designed, fabricated, erected, and tested in

accordance with the criteria set forth in 10 CFR Part 50, Appendix B, “Quality Assurance Criteria

for Nuclear Power Plants and Fuel Reprocessing Plants.” Appropriate records of the design,

fabrication, erection, procurement and testing of SSCs which are determined to have safety

significance are maintained throughout the life of the plant.

The design addresses natural phenomena hazards, fire protection, environmental and dynamic

effects, chemical protection, emergency capability, utility services, inspection, testing and

maintenance, criticality safety, instrumentation and controls, and defense-in-depth. For a more

detailed review of the individual BDC, see SER Sections 3.3.1.1 and 3.3.1.2 for natural

phenomena hazards, environmental and dynamic effects; Section 3.3.1.3 for utility services;

Chapter 5 for nuclear criticality safety; Chapter 7 for fire protection; Section 3.3.1.3 and Chapter

6 for chemical safety; Chapter 8 for emergency management; and Chapter 11 for management

measures for surveillance, maintenance and testing.

The applicant states that IROFS components and systems will be qualified using the applicable

guidance in the Institute of Electrical and Electronics Engineers, Inc. (IEEE) Standard IEEE323, 1983, “IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating

Stations.” The qualification of each IROFS component and system will demonstrate that the

IROFS will perform their safety function under the environmental and dynamic service

conditions in which they will be required to function and for the length of time the function is

required. Additionally, non-IROFS components and systems will be qualified to withstand

environmental stress caused by environmental and dynamic service conditions under which

their failure could prevent satisfactory accomplishment of the IROFS safety functions. 

Furthermore, IROFS components and systems will be designed, procured, installed, tested, and

maintained using the applicable guidance in Regulatory Guide 1.180, “Guidelines for Evaluating

Electromagnetic and Radio-Frequency Interference in Safety-Related Instrumentation and

Control Systems,” Revision 1, dated October 2003 (NRC,2003).

In Section 3.1.7.J of the ISA Summary (LES, 2005b), the applicant states that I&C systems will

be provided to monitor variables and operating systems that are significant to safety over

anticipated ranges for normal operation, for abnormal operation, for accident conditions, and for

safe shutdown. These systems will ensure adequate safety of process and utility service
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operations in connection with their safety function. Controls will be provided to maintain these

variables and systems within the prescribed operating ranges under all normal conditions. I&C

systems will be designed to fail into a safe state or to assume a state demonstrated to be

acceptable if conditions such as disconnection, loss of energy or motive power, or adverse

environments are experienced.

The status and operation of hardware IROFS involving instrumentation that provides automatic

prevention or mitigation of events will be monitored by the plant control system (PCS) by means

of an alarm. This alarm will be provided by an isolated, hardwired digital signal from the

associated IROFS to the PCS programmatic logic controller (PLC). This signal will only be

directed from the associated IROFS to the PCS PLC. The required separation and isolation will

be provided at the IROFS hardware interface in the process equipment for the connections to

the PCS PLC. Consistent with ANSI/IEEE-279-1971, “Criteria for Protection Systems for

Nuclear Power Generating Stations” (ANSI/IEEE, 1971), the isolation devices will be classified

as part of the IROFS boundary and will be designed such that no credible failure at the output of

the isolation device will prevent the associated IROFS from meeting its specified safety function. 

Additionally, the applicant has committed to the criteria contained in IEEE Standard 603-1998,

"IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations“ (IEEE,

1998), for separation and isolation of IROFS systems and components.

The staff conducted an on-site review of both the classified and non-classified portions of the

ISA and concludes that the applicant has appropriately addressed the baseline design criteria

for new facilities, as required by 10 CFR 70.64.

Defense-in-Depth

The applicant based the facility and system designs on defense-in-depth practices. The design

incorporates a preference for engineered controls over administrative controls, and to

incorporate features that enhance safety by reducing challenges to IROFS. Facility and system

IROFS are identified in ISA Summary Section 3.8 (LES, 2005b), and are described in ISA

Summary Section 3.4 (LES, 2005b). The utility and support systems are described in ISA

Summary Section 3.5 (LES, 2005b). The system descriptions identify their associated IROFS

and additional design and safety features that provide defense-in-depth. The staff reviewed the

identified safety features for each of the process systems described in the ISA Summary and

determined that the applicant had established a preference for the selection of engineered

controls over administrative controls to increase overall system reliability and reduce challenges

to IROFS.

The staff concludes that the applicant has appropriately addressed the defense-in-depth

practices for new facilities, as required by 10 CFR 70.64(b).

Baseline Design Criteria and Defense-In-Depth Conclusion

Based on the above, the staff concludes that the applicant has provided sufficient information in

the ISA Summary to demonstrate that the baseline design criteria was applied in an acceptable

manner to the design of the gas centrifuge uranium enrichment facility, and the facility and

system design and facility layout will be based on defense-in-depth practices.
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3.3.3 Safety Program and Integrated Safety Analysis

3.3.3.1 Safety Program

The staff reviewed the applicant’s proposed safety program commitments identified in SAR

Section 3.0 (LES, 2005a) to determine that the three elements of process safety information,

integrated safety assessment, and management measures demonstrates compliance with the

performance requirements of 10 CFR 70.61, that records are established and maintained to

demonstrate compliance with 10 CFR 70.62(b) through (d), and that records are established

and maintained documenting each discovery that an IROFS or management measure has failed

to perform on demand or has degraded such that the performance requirements of 10 CFR

70.61 are not satisfied.

3.3.3.1.1 Process Safety Information

The applicant has developed process safety information program requirements in Section 3.0.1

of the Safety Analysis Report (SAR) (LES, 2005a). Program commitments require that up-todate documentation of process safety information pertaining to the hazards of all materials used

or produced in the process, the technology of the process, and the process equipment are

compiled. The program commitments also require that process safety information be

maintained up-to-date in accordance with the program elements described in SAR Section 11.1

(LES 2005a). SER Chapter 11 discusses the staff review and evaluation of the proposed

program elements needed to maintain the process safety information documentation. 

Therefore, the staff concludes that if the program elements are followed, there is reasonable

assurance that the applicant will compile and maintain process safety information up-to-date.

The applicant has committed to develop procedures and criteria for changing the ISA that meet

the requirements of 10 CFR 70.72 in accordance with procedure development and

implementation requirements contained in SAR Section 11.4 (LES, 2005a). Also, the ISA will

be maintained by personnel with the appropriate experience and expertise in engineering and

process operations. The ISA Team for the various processes consists of individuals who are

knowledgeable in the ISA method(s) and the operation, hazards, and safety design criteria of

the particular process. Training and qualifications of individuals responsible for maintaining the

ISA are described in SAR Sections 2.2, 3.2, and 11.3 (LES, 2005a).

3.3.3.1.2 Integrated Safety Analysis

In SAR Section 3.0.2 (LES, 2005a), the applicant identifies ISA program commitments that were

used to establish the ISA process. Those commitments include the performance of an ISA for

each process that identifies the radiological hazards, chemical hazards that could increase

radiological risk, facility hazards that could increase radiological risk, 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. Implementation of the Configuration Management Program

described in SAR Section 11.1 (LES, 2005a) will maintain the ISA and supporting

documentation accurate and up-to-date. The ISA update process will account for any changes

to the facility or its processes. The update process will also verify that initiating event

frequencies and IROFS reliability values assumed in the ISA remain valid. Management

policies, organizational responsibilities, revision time frame, and procedures to perform and
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approve revisions to the ISA are outlined in SAR Section 11.0 (LES, 2005a). Personnel used to

update and maintain the ISA and ISA Summary will be trained and qualified, as described in

SAR Section 11.3 (LES 2005a). Proposed changes to the facility will be evaluated using the

ISA method(s). All IROFS will be maintained available and reliable, and unacceptable

performance deficiencies are addressed. The adequacy of existing IROFS and associated

management measures are promptly evaluated to determine if they are impacted by changes to

the facility. Unacceptable performance deficiencies associated with IROFS are addressed that

are identified through updates to the ISA. Written procedures will be maintained onsite in

accordance with the requirements of SAR Section 11.4 (LES, 2005a).

3.3.3.1.3 Management Measures

In SAR Section 3.0.3 (LES 2005a), the applicant describes management measures that

comprise the principal mechanism by which the reliability and availability of each IROFS is

ensured. General requirements applicable to each IROFS for configuration management,

maintenance, training and qualifications, procedures, audits and assessments, incident

investigation, records management, and other quality assurance elements, are discussed. Any

management measures deviating from these general requirements, which are consistent with

the performance requirements assumed in the ISA documentation, are discussed in the ISA

Summary (LES, 2005b). Incident investigations are conducted within the applicant’s Corrective

Action Program. Incidents associated with IROFS, and any items that may affect the function of

IROFS, include: processes that behave in unexpected ways; procedural activities not

performed in accordance with an approved procedure; discovered deficiency, degradation, or

non-conformance of an IROFS; or, any items that may affect the function of IROFS. Feedback

from the results of incident investigations are used, as appropriate, to modify management

measures to provide continued assurance of the availability and reliability of IROFS, in order to

meet the performance requirements of 10 CFR 70.61. All records associated with IROFS, and

any item that may affect the function of IROFS, will be managed and controlled in a systematic

manner in order to provide identifiable and retrievable documentation. The management

measures are further detailed in SAR Chapter 11 (LES 2005a), and evaluated in Chapter 11 of

this Safety Evaluation Report (SER). All IROFS are Quality Level 1 items per the applicant’s

approved Quality Assurance Program Description.

3.3.3.1.4 Safety Program Conclusions

Based on the above information, the staff concludes that the applicant meets the requirements

of 10 CFR 70.62(a)(1) through (3) to establish and maintain a safety program that includes

process safety information, integrated safety analysis and management measures, and

appropriate safety program records. The staff also concludes that the applicant has an

appropriate program to establish and maintain records of IROFS failures that will be retrievable

for NRC inspection.

3.3.3.2 Integrated Safety Analysis and ISA Summary

The staff reviewed the ISA Summary to confirm that it contained appropriate information

pertinent to: (1) a general description of the site, (2) a general description of the facility, (3) a

description of each process in sufficient detail to understand the theory of operation and the

hazards that were identified in the ISA and a general description of the accident sequences; (4)

information demonstrating compliance with the performance requirements of 10 CFR 70.61,
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including a description of management measures, the requirements for criticality monitoring and

alarms, and the baseline design criteria; (5) a description of the team, qualifications, and the

methods used to perform the ISA; (6) a list briefly describing each IROFS in sufficient detail to

understand its function in relation to the performance requirements of 10 CFR 70.61; (7) a

description of the proposed quantitative standards used to assess the consequences of

chemical exposures; (8) a list of sole IROFS, and (9) a description of the definitions of “not

unlikely,” “unlikely,” and “highly-unlikely,” as used in the evaluations in the ISA.

The staff also reviewed selected portions of the applicant’s Integrated Safety Analysis (ISA) to

confirm that the applicant performed an ISA of appropriate detail for the proposed gas

centrifuge uranium enrichment process, that identifies: radiological hazards, chemical hazards

of licensed material and hazardous chemicals produced from licensed material; facility hazards;

potential accident sequences caused by credible internal and external events, including natural

phenomena; consequences and likelihood of occurrence of each potential accident sequence;

each item relied on for safety (IROFS), as required by 10 CFR 70.62(c).

3.3.3.2.1 ISA Summary Content

ISA Summary Section 3.2 (LES, 2005b) contains the site description, and includes information

on site geography, demographics and land use, meteorology, hydrology, geology, seismology,

and stability of subsurface materials. Based on the review conducted in Section 3.3.1.1 of this

SER, the staff has determined that the information provided is sufficient to comply with 10 CFR

70.65(b)(1).

ISA Summary Section 3.3 (LES, 2005b) contains the facility description, and includes

information on the buildings and major components, and structural design criteria. Based on

the review conducted in Section 3.3.1.2 of this SER, the staff has determined that the

information provided is sufficient to comply with 10 CFR 70.65(b)(2).

ISA Summary Sections 3.4 and 3.5 (LES, 2005b) contain descriptions of the process and utility

and support systems, Sections 3.3 describes the facility hazards and Sections 3.1 and 3.7

describe the radiological, and chemical hazards and accident sequences. Based on the review

conducted in Section 3.3.1.3 of this SER, the staff has determined that the information provided

is sufficient to comply with the requirements of 10 CFR 70.65(b)(3).

ISA Summary Section 3.1 (LES, 2005b) contains information regarding the ISA methods used

to demonstrate compliance with the performance requirements of 10 CFR 70.61, a description

of the management measures, the requirements for criticality monitoring and alarms, and

baseline design criteria. The demonstration of meeting the performance requirements is

contained in Sections 3.7 and 3.5 of the ISA Summary (LES, 2005b). Based on the review

conducted in Section 3.3.1.1 of this SER and additional information in Chapters 4, 5, 6, and 7 of

this SER, the staff has determined that this information is sufficient to comply with 10 CFR

70.65(b)(4).

ISA Summary Section 3.1 (LES, 2005b) contains information regarding the description of the

ISA Team, qualifications, and the methods used to perform the ISA. Based on the review

conducted in Section 3.3.3.2.2 of this SER, the staff has determined that this information is

sufficient to comply with 10 CFR 70.65(b)(5).
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ISA Summary Section 3.8, Table 3.8-1 (LES, 2005b) contains a list briefly describing each

IROFS in sufficient detail to understand their functions in relationship to the performance criteria

of 10 CFR 70.61. Based on the review conducted in Section 3.3.3.2.4 of this SER, the staff has

determined that this information is sufficient to comply with 10 CFR 70.65(b)(6).

ISA Summary Section 3.1, and Tables 3.1-3 and 4 (LES, 2005b) contains information regarding

the selection of quantitative standards used to assess the consequences to an individual from

acute chemical exposure to licensed material or chemicals produced from licensed material. 

Based on the review conducted in Section 3.3.3.2.2.2 of this SER, the staff has determined that

this information is sufficient to comply with 10 CFR 70.65(b)(7).

ISA Summary Section 3.8, Table 3.8-2 (LES, 2005b) contains a list of sole IROFS used for

preventing or mitigating an accident sequence that exceeds the performance requirements of 10

CFR 70.61. Based on the review conducted in Section 3.3.3.2.5 of this SER, the staff has

determined that this information is sufficient to comply with 10 CFR 70.65(b)(8).

ISA Summary Section 3.1 and Table 3.1-5 (LES, 2005b) contains a description of the definitions

of unlikely, highly unlikely and credible as used in the evaluations in the ISA. Based on the

review conducted in Section 3.3.3.2.2.2 of this SER, the staff has determined that this

information is sufficient to comply with 10 CFR 70.65(b)(9).

ISA Summary Content Conclusion

Based on the staff’s review of the ISA Summary, the staff concludes that the ISA Summary

content complies with the requirements of 10 CFR 70.65(b).

3.3.3.2.2 ISA Introduction

The staff reviewed information concerning the ISA methodology located in SAR Section 3.1

(LES, 2005a) and ISA Summary Section 3.1.1 (LES, 2005b). Two ISA teams were employed, a

classified ISA team to deal with the sensitive classified processes and a non-classified ISA team

to analyze the remaining processes. The ISA was performed in accordance with a procedure

developed to guide the conduct of the ISA, and several participants were employed on both ISA

teams to ensure consistency of the results. The applicant used a semi-quantitative risk method

consistent with "Example Procedure for Accident Sequence Evaluation," Appendix A to Chapter

3 of NUREG-1520 (NRC, 2002a). An integration checklist is used as a guide to facilitate the

integrated review process. Included in the applicant’s review is the identification of IROFS

functions that may be simultaneously beneficial and harmful with respect to different hazards,

and interactions that might not have been considered in previous subsets. The review is

intended to ensure that the designation of one IROFS does not negate the preventive or

mitigation function of another IROFS. 

3.3.3.2.2.1 Integrated Safety Analysis Team Qualifications

In SAR Section 3.2 (LES, 2005a), the applicant describes the two ISA teams (classified and

non-classified). To facilitate consistency of results, several members were employed for each

team. The ISA was performed by teams with expertise in various technical disciplines and

included personnel with experience and knowledge specific to each process or system being

evaluated. The applicant states that the ISA Team Leader was trained and knowledgeable in
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the ISA method used. The ISA Manager is responsible for the overall direction of the ISA. The

ISA Team Leader has an adequate understanding of the process, but is not the responsible

cognizant engineer or enrichment process expert. Staff review of the team’s credentials

confirmed that they are consistent with NUREG-1520 (NRC, 2002a). The staff also finds that

the above responsibilities and authorities are consistent with the guidance provided in NUREG1520 (NRC, 2002a), and are, therefore, acceptable.

Based upon the above, the staff concludes that the applicant’s ISA team composition and

membership meet the requirements of 10 CFR 70.62(c)(2).

3.3.3.2.2.2 Hazard and Accident Sequence Identification

The applicant used a process hazard method (HAZOP) to identify potential hazards related to

UF6 process systems and Technical Services Building Systems and potential accident

sequences. The HAZOP method is a structured technique commonly used in the chemical

industry that is well suited to analyze processes during or after a detailed design stage. The

method uses an interdisciplinary team to identify hazards and operability problems resulting

from deviations from the process’s design intent that could lead to undesirable consequences. 

In SAR Section 3.1.2 (LES, 2005a), the applicant states that implementation of the HAZOP

method was accomplished by either validating the Urenco HAZOPs for the NEF design or

performing a new HAZOP for systems where there were no existing HAZOPs. For cases where

there was an existing HAZOP, the ISA team developed a new HAZOP through the validation

process. The format of the HAZOP tables used in the ISA documentation is provided. The

“Fire” and “External Event” guidewords are handled on a facility-wide basis. The applicant

states that the HAZOPs were revised/updated as necessary to be consistent with the

requirements of 10 CFR Part 70 and as described in NUREG-1513 (NRC, 2001) and NUREG1520 (NRC, 2002a). The process hazards are described in the ISA Summary Section 3.6 (LES,

2005b), and the associated accident sequences are identified in ISA Summary Section 3.7

(LES, 2005b).

The HAZOP method is widely used in the chemical processing industry because it is suitable for

performing a detailed analysis of a wide range of hazards to identify potential accident

sequences. NUREG-1513, “Integrated Safety Analysis Guidance Document,” Appendix A,

“Flowchart for Selecting a Hazards Analysis Technique” (NRC, 2001), identifies the HAZOP

technique as an acceptable approach. Therefore, the staff concludes that the process hazard

analysis method used by the applicant is acceptable for the identification of potential

radiological, chemical and facility hazards, and potential accident sequences caused by process

deviations or other events internal to the facility and credible external events, including natural

phenomena that could lead to a loss of UF6 confinement or a criticality.

The hazard identification process used by the applicant documented materials that are

radioactive, fissile, flammable, explosive, toxic, and reactive, and identified potentially

hazardous conditions. Hazards were assessed individually for the potential impact on discrete

components. However, hazards related to fires and external events were assessed on a facilitywide basis. The Fire Hazard Analysis (discussed in SAR Section 7.2 (LES, 2005a)) was

consulted in order to place reasonable and conservative bounds on the fire scenarios. External

events evaluated included seismic, tornado, tornado missile and high wind, snow and ice,

flooding, local precipitation, transportation and nearby facility accidents, aircraft, pipelines,

highway, railroad, and internal flooding from above ground storage tanks.
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The applicant considered common mode failures and common cause situations, support system

failures, divergent impacts of IROFS, non-IROFS impacts on system performance, multiple

impact scenarios, system interactions and interdependence, and major hazards or events which

tend to be common cause situations that could lead to interactions between processes,

systems, buildings, etc.

SAR, Chapter 6, Tables 6.1-1 through 6.1-6 (LES, 2005a), identifies the hazardous properties of

all chemicals used on site, and the inventories and locations of chemicals of concern (including

UF6). Potential interactions involving UF6 and any reaction products are identified in SAR,

Section 6.1.2 (LES, 2005a). These chemicals include HF and UO2F2. This section identifies the

physical properties, reactivity, toxicological properties, and flammability of these chemicals. 

Chemical reactions and interactions involving UF6 and water, Fomblin oil, chemical trap

materials, and other materials used in the process are described in SAR, Section 6.2.1 (LES,

2005a). SAR Section 7.2 describes the fire hazards analysis for those facility areas containing

licensed material. ISA Summary Chapter 3 identifies the external hazards. Other general types

of process hazards are briefly discussed in ISA Summary, Section 3.6 (LES, 2005b). The

applicant identified either a loss of confinement (of UF6) or a criticality as the hazard of concern

in ISA Summary, Section 3.1.4 (LES, 2005b). The staff agrees that all accident sequences at a

gas centrifuge uranium enrichment facility that could exceed the performance requirements of

10 CFR 70.61 are the direct result of either a loss of confinement (of UF6) or a criticality. 

Potential accident sequences that could result in a UF6 release or criticality of high or

intermediate consequence are discussed in detail in ISA Summary Section 3.7 (LES, 2005b). 

Based on the process hazard analysis method used to identify hazards, a review of the

information in the application and ISA Summary and the staff’s visit to a similar gas centrifuge

uranium enrichment facility in Almelo, The Netherlands, the AREVA engineering offices in

Marlborough, Massachusetts, and the applicant’s offices in Washington, D.C., the staff has

reasonable assurance that the applicant has identified all of the potential internal and external

hazards and accident sequences that could affect the safety of licensed materials, and has

identified the materials and conditions that could result in a hazardous condition, including the

expected inventory amounts and locations. Potential interactions between materials or

conditions that could result in hazardous conditions were appropriately considered.

Risk Matrix Development

The applicant utilized a risk index method in order to categorize accident sequences in terms of

likelihood of occurrence and consequences. This evaluation identifies the accident sequences

which could exceed the performance requirements of 10 CFR 70.61 and, therefore, require

designation of Items Relied on for Safety (IROFS). The staff reviewed the methods used by the

applicant to determine the consequences and likelihood of each potential accident sequence

identified through the HAZOP process. In SAR Section 3.1.3.3 (LES, 2005a), the applicant

describes the consequence verses likelihood risk matrix. The risk matrix and computed index

values are shown in SAR Table 3.1-6 (LES, 2005a). Use of the risk matrix is demonstrated in

ISA Summary Section 3.1 (LES, 2005b).

Consequence Analysis Method

In SAR Section 3.1.3.1 (LES, 2005a), the applicant used the dispersion method discussed in

SAR Section 6.3.2 (LES, 2005a), for determining the chemotoxic exposure to HF and UO2F2. 
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The radiological and chemical consequence severity levels are provided in SAR Table 3.1-3

(LES, 2005a). Information on the chemical dose limits specific to NEF are found in SAR Table

3.1-4 (LES, 2005a). The applicant developed credible accident scenarios, and the dispersion

analysis and chemical/radiological dose assessment associated with each accident sequence in

accordance with NUREG/CR-6410, “Nuclear Fuel Cycle Facility Accident Analysis Handbook”

(NRC, 2002b). The consequences of an inadvertent criticality were conservatively assumed to

be high for both the public and the workers. For accident sequences postulated to result in an

inadvertent criticality, IROFS are specified to ensure subcriticality under all normal and credible

abnormal conditions.

Based on the above, the staff finds that applicant’s method for consequence determination is

consistent with the guidance described in NUREG-1513 (NRC, 2001) and NUREG-1520 (NRC,

2002a), and therefore, acceptable.

Consequence Categories

Accident sequences identified by the applicant as a result of the process hazards analysis are

categorized as either a “high” consequence event, an “intermediate” consequence event, or are

considered a low consequence event in accordance with the performance criteria of 10 CFR

70.61. ISA Summary Table 3.1-3 (LES, 2005b) identifies the consequence severity categories

for workers, offsite public and the environment for defined radiation and chemical doses. The

values proposed by the applicant for both radiological dose and chemical dose are consistent

with the guidance in NUREG-1520, Appendix A, Table A-5 (NRC, 2002a). All of the identified

high-consequence and intermediate-consequence events are listed in the ISA Summary (LES,

2005b).

Quantitative Standards for Chemical Consequences

The applicant identifies the radiological and chemical dose consequence categories for workers,

offsite public, and environment in ISA Summary Table 3.1-3 (LES, 2005b), and SAR Table 6.3-2

(LES, 2005a). The chemical dose consequence categories are consistent with the guidance

contained in NUREG-1520, Table A-1 (NRC, 2002a). SAR Tables 6.3-3 and 6.3-4 (LES,

2005b) contain the AEGL values for HF and UF6, respectively. The applicant proposed a

worker-exposure strategy that incorporates 10-minute AEGL values for HF, as used in NUREG1391 (NRC, 1991). The staff finds the use of AEGL standards to be acceptable because these

are unambiguous quantitative standards developed by the National Advisory Committee for

Acute Guideline Levels for Hazardous Substances (AEGLs) that are used nationally in a broad

application for emergency planning, response, and prevention in the community, the workplace,

transportation and remedial action. Application of the consequence categories is discussed in

Section 6.3.2.1 of the SAR. The quantitative standards that the consequence categories are

based on are in accordance with 10 CFR 70.65(b)(7) and are consistent with the standards in

10 CFR 70.61.

Likelihood Determination

The likelihood evaluation method is discussed in SAR Section 3.1.3.2, (LES, 2005a). The

definitions of “not unlikely,” “unlikely,” and “highly unlikely” are in SAR Table 3.1-5 (LES, 2005a). 

“Not unlikely” is defined as “Likelihood Category 3" with a probability of occurrence of more than

10-4 per event per year. “Unlikely” is defined as “Likelihood Category 2" with a probability of
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occurrence of between 10-4 and 10-5 per event per year. “Highly unlikely” is defined as

“Likelihood Category 1" with a probability of occurrence of less than 10-5 per event per year. 

The staff concludes the definitions are acceptable because they are consistent with the

guidance provided in NUREG-1520 (NRC, 2002a).

Not-Credible Determination

In SAR Section 3.1.3.2, (LES, 2005a) the applicant states that the definition of “not credible” is

taken from NUREG-1520 (NRC, 2002a), and any one of the following independent acceptable

sets of qualities could define an event as not credible:

a. An external event for which the frequency of occurrence can conservatively be estimated

as less than once in a million years;

b. A process deviation that consists of a sequence of many unlikely human actions or

errors for which there is no reason or motive (In determining that there is no reason for

such actions, a wide range of possible motives, short of intent to cause harm, must be

considered. Necessarily, no such sequence of events can ever have actually happened

in any fuel cycle facility.)

c. Process deviations for which there is a convincing argument, given physical laws that

they are not possible, or are unquestioningly extremely unlikely.

The staff concludes that the above definition of “not credible” is consistent with the regulatory

guidance and suitable for use in determining the likelihood performance requirements of 10

CFR 70.61.

Likelihood Evaluation Method

ISA Summary Section 3.1.1.3.2 (LES, 2005b) discuss the definitions of “Not Unlikely,”

“Unlikely,” and “Highly Unlikely,” and ISA Summary, Table 3.1-5 (LES, 2005b), cross references

the three likelihood categories with a probability of occurrence based on approximate order of

magnitude ranges. The proposed values are consistent with NUREG-1520, Chapter 3,

Appendix A, Table A-6 (NRC, 2002a). Implementation of the evaluation method is discussed in

ISA Summary Section 3.1.1.3 (LES, 2005b). ISA Summary Tables 3.7-1 through 3.7-4 (LES,

2005b), show how each designated IROFS acts to prevent or mitigate the consequences of an

accident sequence. When multiple IROFS are designated, IROFS interactions are considered. 

The likelihood of failure was qualitatively evaluated for each IROFS, often based on the

operational history of similar facilities. Each sequence was evaluated as either “Not Unlikely,”

“Unlikely,” or “Highly Unlikely.”

Safe-By-Design Evaluation Method for Nuclear Criticality Safety

In SAR Section 3.1.3.2 (LES, 2005a), the applicant described a safe-by-design ISA method for

selected equipment for nuclear criticality safety (NCS) used to identify safe-by-design

components, the failure of which would be highly unlikely. The applicant described the

connection between subcriticality and the safe-by-design ISA process for NCS. Using the safeby-design ISA process, there are no accident sequences and, hence, IROFS are not identified
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because it is highly unlikely these components would fail. Those safe-by-design components

are considered items which may affect IROFS.

A qualitative determination of highly unlikely can apply to passive design component features of

the facility that do not rely on human interface to perform the NCS function. Safe-by-design

components are those components that by their physical size or arrangement have been shown

to have a keff < 0.95. The definition of safe-by-design components encompasses two different

categories of components. The first category includes those components that are safe-byvolume, safe-by-diameter, or safe-by-slab thickness (i.e., favorable geometry components). A

set of generic, conservative NCS calculations has determined the maximum volume, diameter,

or slab thickness that would result in a keff < 0.95. A favorable geometry component has a

volume, diameter, or slab thickness that is less than the associated value for keff < 0.95. The

components in the second category (i.e., non-favorable geometry components) require a more

detailed NCS Analysis to demonstrate keff < 0.95. For the non-favorable geometry components,

the design configuration is not bounded by the results of the generic, conservative NCS

calculations for maximum volume, diameter, or slab thickness that would result in a keff < 0.95.

For failures of these passive safe-by-design components (i.e., both favorable geometry

components and non-favorable geometry components) to be considered highly unlikely, those

components must also meet the criterion that the only potential means to effect a change that

might result in a failure to function would be to implement a design change (i.e., no potential

failure mode exists). The evaluation of the potential to adversely impact the safety function of

these design features includes consideration of potential mechanisms to cause bulging,

corrosion, or breach of confinement/leakage and the subsequent accumulation of material. The

evaluation further includes consideration of adequate controls to ensure that the double

contingency principle is met. For each of these passive design components (i.e., both favorable

geometry components and non-favorable geometry components), it must be concluded that

there is no credible means to effect a geometry change that might result in a failure of the safety

function and that significant margin exists.

For favorable geometry components, significant margin is defined as a margin of at least 10

percent, during both normal and upset conditions, between the actual design parameter value of

the component and the value of the corresponding critical design attribute. For non-favorable

geometry components, significant margin is defined as keff < 0.95, where keff = kcalc + 3σcalc. This

calculation of keff conservatively assumes the components are full of uranic breakdown material

at maximum enrichment, with worst credible moderation, and with worst credible reflection.

These passive, safe-by-design features (i.e., both favorable geometry components and nonfavorable geometry components) are considered items that may affect IROFS. As a result,

Quality Level 1 requirements apply to these features. Also, the configuration management

program required by 10 CFR 70.72 ensures the maintenance of the safety function of these

features and assures compliance with both the double contingency principle and the defense-indepth criterion of 10 CFR 70.64(b).

Additionally, the applicant commits to apply the guidelines of the ASME QA standard NQA-1-

1994, “Quality Assurance Program Requirements for Nuclear Plants,” including supplements as

revised by the ASME NQA-1-a-1995 Addenda (ASME, 1994), to all IROFS and any item,

structures, systems, controls, and administrative controls that are determined to affect the
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functions of the IROFS, and, in general, to items required to satisfy regulatory requirements. All

IROFS are Quality Level 1 components. 

Safe-By-Design Highly-Unlikely Determination for NCS

SER Chapter 5 provides the staff’s review of the applicant’s demonstration for meeting highly

unlikely for safe-by-design components.

Risk Index Evaluation

The staff reviewed the applicant’s risk index evaluation description in SAR Section 3.1.4 (LES,

2005a). The results of the ISA are summarized in tabular form. The risk index evaluation

process cross references an accident sequence consequence with its likelihood to first

determine the total risk, and then to demonstrate that with the application of the selected

IROFS, the performance requirements of 10 CFR 70.61 are met. The ISA results are

summarized in tabular form, with columns for the initiating event and IROFS. For sequences

that place the system in a vulnerable state, the duration of the vulnerable state is considered,

and a duration index assigned. The values of all index numbers in a sequence are added to

obtain a total likelihood index, T. Accident sequences are then assigned to one of three

likelihood categories, depending on the value of the index in accordance with SAR Table 3.1-8

(LES, 2005a). The criteria of SAR Tables 3.1-9 through 3.1-11 are used to assign index

numbers to accident sequences. The staff reviewed the risk index evaluation for selected

accident scenarios in the fire, chemical, and criticality safety areas as part of the “vertical-slice”

review and determined that the applicant’s evaluation process was appropriately applied. The

staff concludes that the risk evaluation methodology described by the applicant is consistent

with the guidance in NUREG-1520 (NRC, 2002a) and suitable for determining which accident

sequences require IROFS, and the level of risk reduction provided the IROFS to comply with 10

CFR 70.61. 

3.3.3.2.3 Compliance Item Commitments

In SAR Sections 3.3.1 through 3.3.5 (LES, 2005a), the applicant committed to: (1) provide

justification for having Initiating Event Frequency (IEF) index numbers for certain accident

sequences or else revise the ISA Summary; (2) provide justification for having Failure

Probability Index Numbers (FPINs) when using Administrative Controls with ‘use of’ a

component or device or else revise the ISA Summary; (3) provide justification for having FPINs

when using Administrative Controls with ‘verification of’ a state or condition or else revise the

ISA Summary; (4) use three of four criteria when using Administrative Controls with

‘independent sampling;’ and (5) meet all criteria when using Enhanced Administrative Controls,

which require ‘independent verification’ of a safety function.

3.3.3.2.4 Descriptive List of IROFS

ISA Summary, Table 3.8-1 (LES, 2005b), lists all IROFS identified in ISA Summary Section 3.7

(LES, 2005b). The IROFS address hazards that could result in a loss of UF6 confinement

and/or possible criticality events. Safety functions addressed include radiological safety,

chemical safety, criticality safety, and fire protection. The staff notes that no IROFS are

identified as a result of using the alternate “safe-by-design” ISA method because “highly
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unlikely” is achieved with a significant margin. The staff concludes that the applicant has

provided IROFS for all identified high and intermediate consequence accident sequences.

ISA Summary Table 3.8-1 (LES, 2005b) also describes the safety function of each IROFS and

the reliability management measures. More detailed descriptions of each IROFS are contained

in ISA Summary Sections 3.4 and 3.5 (LES, 2005b). Hardware control descriptions contain

sufficient information to permit evaluation of its reliability. Safety limits on key parameters are

identified. Administrative control IROFS are adequately described to permit an understanding

that adherence to it should be reliable. Additionally, the applicant commits to apply the

guidelines of the ASME QA standard NQA-1-1994, “Quality Assurance Program Requirements

for Nuclear Plants,” including supplements as revised by the ASME NQA-1-a-1995 Addenda

(ASME, 1994), to all IROFS and any item, structures, systems, controls, and administrative

controls that are determined to affect the functions of the IROFS, and, in general, to items

required to satisfy regulatory requirements. All IROFS are Quality Level 1 components. 

The applicant states that it will ensure that IROFS components and systems will be able to

perform their required safety functions under normal and accident conditions (e.g., pressure,

temperature, humidity, seismic motion, electromagnetic interference, and radiofrequency

interference) as required by the ISA. The applicant states that IROFS will be qualified using the

applicable guidance in Institute of Electrical and Electronics Engineers (IEEE) Standard IEEE323, 1983, “IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating

Stations” (IEEE, 1983). Furthermore, IROFS components and systems will be designed,

procured, installed, tested, and maintained using the applicable guidance in Regulatory Guide

1.180, “Guidelines for Evaluating Electromagnetic and Radio-Frequency Interference in SafetyRelated Instrumentation and Control Systems,” Revision 1, dated October 2003 (NRC, 2003a). 

Redundant IROFS systems will be separate and independent from each other. IROFS systems

will be designed to be fail-safe. Process control system failures will not affect the ability of the

IROFS systems to perform their required safety functions. Required testing and calibration of

IROFS will be consistent with the assumptions of the ISA and setpoint calculations, as

applicable. For hardware IROFS involving instrumentation that provides automatic prevention

or mitigation of events, setpoint calculations are performed in accordance with a setpoint

method, which is consistent with the applicable guidance provided in Regulatory Guide 1.105,

“Setpoints for Safety-Related Instrumentation,” Revision 3 (NRC, 1999a).

In SAR Section 3.3.6 (LES, 2005a), the applicant states that upon completion of the final

design, the IROFS boundaries will be defined. In defining the boundaries of each IROFS, LES

procedure, "IROFS Boundary Definitions," will be used. The procedure requires identification of

each support system and component necessary to ensure the IROFS is capable of performing

its specified safety function. The procedure also requires identification of the management

measures necessary to support the IROFS availability and reliability ISA assumptions. Because

historical failure data was not used to derive all of the indices in ISA Summary Table 3.7-1, once

the facility is operating, failure data will be trended and the impact of this failure data on the

values assumed in the ISA will be evaluated to validate those assumptions. By completing this

procedure for each IROFS, the applicant will more specifically: (1) identify and establish

engineered and administrative IROFS to protect against the credible accident sequences; (2)

identify and establish management measures for specific IROFS to protect against the credible

accident sequences; and (3) confirm that the credible accident sequences are highly unlikely,

and that credible intermediate accident sequences are unlikely. In conjunction with the
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applicant’s configuration management program, the IROFS boundary provides assurance that

IROFS will be maintained available and reliable over the life of the facility.

SER Chapter 11 discusses the management measures that the applicant will apply to IROFS

and items that could affect IROFS to assure that they are reliable and available when called

upon to perform their intended safety functions.

The staff reviewed the IROFS for selected accident sequences involving chemical hazards,

criticality safety and fire protection, as described in SER Chapters 5 through 7.

Under 10 CFR 70.72(d)(3), for all changes that affect the ISA Summary, the licensee must

submit to NRC annually revised pages of the ISA Summary. Any changes to IROFS would be

included in the annual revisions to the ISA Summary.

The staff concludes, pursuant to 10 CFR 70.23(a), that the preliminary design bases of the

IROFS evaluated in this SER section will provide reasonable assurance of protection against

natural phenomena and the consequences of potential accidents. Given that these conclusions

are based on preliminary design information, the staff is imposing a license condition to ensure

that the final design is adequate and acceptable to the staff. Specifically, the following condition

will be included in the license:

“The applicant shall utilize its procedure, "IROFS Boundary Definitions," to define the

boundaries of each IROFS. Completed IROFS boundaries for all IROFS shall be

available for inspection at the time of the operational readiness review.”

The applicant has committed to conforming to the applicable guidance of the codes and

standards and regulatory guidance listed in SAR Section 3.3.7 (LES, 2005a), for the design,

procurement, installation, testing, and maintenance of I&C related IROFS to comply with the

performance requirements of 10 CFR 70.61.

The staff concludes that the applicant has adequately described the management measures

applied to IROFS (including safety grading), characteristics of its preventive, mitigative, or other

safety function, and assumptions and conditions, under which the item is relied on to support

compliance with the performance requirements of 10 CFR 70.61.

3.3.3.2.5 List of Sole IROFS

ISA Summary, Table 3.8-2 (LES, 2005b), identifies each sole IROFS and cross references it to

the specific accident sequences for which it is applicable. The accident sequences in turn

provide a clear and unambiguous reference to the process to which the item applies. The list

includes a descriptive title of the IROFS, and a clear and traceable reference to the description

of the item as it appears in the full list of all IROFS. Therefore, the staff concludes that the

applicant has adequately described all IROFS that are the sole item for preventing or mitigating

an accident sequence identified by the applicant.

3.3.3.3 Summary of ISA Results

3.3.3.3.1 Hazards
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The applicant’s ISA uses the HAZOP method for identifying the hazards for UF6 process

systems, the Technical Services Building systems, the Centrifuge Assembly Building systems,

and the Uranium Byproduct Storage Pad. The HAZOP analysis was applied to discrete process

components. Radiological hazards identification considered the characteristics of uranium

enriched to 5 percent (see Chapter 4 of this SER). Criticality hazards identification was

performed for areas of the facility where fissile material is expected to be present (see Chapter

5 of this SER). Chemical hazards identification included those from licensed material and

chemicals produced from licensed material, including chemical interactions (see Chapter 6 of

this SER). Fire hazards identification considered in-situ and transient combustible sources and

the use of fire barriers (see Chapter 7 of this SER). External hazards were considered at the

site and facility level (see Chapter 3 of the SER). The applicant identified and listed hazards

that could result in accident sequences exceeding the performance requirements of 10 CFR

70.61 in ISA Summary Tables 3.7-1 through 3.7-4. The HAZOP method is an acceptable

hazard identification method per the guidance provided in NUREG-1513 (NRC, 2001).

Based on the above, the staff has reasonable assurance that the applicant used an acceptable

hazard identification technique and identified all of the radiological hazards relating to

possessing or processing licensed material; the chemical hazards of licensed material and

hazardous chemicals produced from licensed material; facility hazards that could affect the

safety of licensed material, and thus, present an increased radiological risk; and hazards related

to process deviations or other events internal to the facility and credible external events,

including natural phenomena.

3.3.3.3.2 Accident Sequences

In ISA Summary Tables 3.7-1 through 3.7-4, the applicant lists and describes the identified

accident sequences and external and fire events for which the consequences could exceed the

10 CFR 70.61 performance requirements. ISA Summary Table 3.8-1 identifies each IROFS

listed in the ISA Summary and how it protects against each accident sequence. The information

is sufficient to determine how each accident sequence that could exceed the performance

requirements of 10 CFR 70.61 is protected against by IROFS. The staff performed a review of

selected high-consequence and intermediate-consequence events in the areas of chemical

safety, criticality safety, and fire protection to confirm that the applicant had properly identified

and analyzed accident sequences and the related consequences. The staff identified one

unanalyzed accident sequence with consequences that exceeded the performance

requirements that were overlooked by applicant. The accident sequence involves either a feed

or depleted uranium cylinder shipment truck fire, as discussed in SER Section 7.3.2.2. The

applicant addressed the accident sequence and identified additional IROFS.

Based on the above, the staff has reasonable assurance that the applicant has identified and

analyzed all accident sequences that could potentially exceed the 10 CFR 70.61 performance

requirements by resulting in high or intermediate consequences.

3.3.3.3.3 IROFS and Management Measures 

The only accident sequence types that can potentially result in intermediate or high

consequences at this facility are loss of confinement events (i.e., caused by process upsets,

human error, natural phenomena, fires, and external events) and criticality accidents (i.e., which

are assumed to have a high consequence to the worker). Management measures are identified
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to ensure the IROFS are available and reliable when needed to perform their safety function

(see SER Chapter 11). All IROFS are designated Quality Level 1 items under the applicant’s

approved Quality Assurance Program Description. Safe-by-design components do not have

IROFS identified because a failure is considered highly unlikely (i.e., there is no credible way to

effect a geometry change that might result in a failure of the safety function). However, safe-bydesign components are considered items which may affect IROFS, and the configuration

management system required by 10 CFR 70.72, which is implemented by the facility

Configuration Management Program, ensures the maintenance of the safety function of these

features. See SER Chapter 5 for a discussion of IROFS and safe-by-design components

related to criticality safety, Chapter 6 for chemical safety IROFS, and Chapter 7 for fire safety

IROFS. The staff conducted a sample review of selected IROFS related to chemical safety,

criticality safety and fire protection for selected accident sequences. The staff review identified

the need for additional IROFS related to a cylinder shipment truck fire that were omitted by the

applicant, as discussed in SER Section 7.3.2.2. The applicant subsequently identified

appropriate IROFS for this accident scenario.

Based on the above, the staff concludes that the applicant has identified suitable IROFS to

make intermediate consequence events unlikely and high consequence events highly unlikely,

and suitable management measures are identified to make the IROFS available and reliable

when called upon to perform their safety function.

3.3.3.3.4 Meeting the Performance Requirements of 10 CFR 70.61

Based on the above information regarding hazards, accident sequences, IROFS, and

management measures, the staff concludes that the applicant has: (1) identified hazards

related to this type of facility; (2) identified credible events that could exceed the performance

requirements of 10 CFR 70.61 through the application of appropriate accident identification

method in accordance with NUREG-1513 (see SER Chapter 3); (3) identified appropriate

chemical dose and radiological dose values for determining intermediate consequence and high

consequence events (see SER Section 3.3.3.b.2); (4) determined the consequences in

accordance with the fuel facility accident analysis guidance in NUREG-6410 (see SER Chapters

3 and 6)(NRC, 2002b); (5) established appropriate definitions for likelihood; and (6) applied

those definitions in an acceptable manner to demonstrate that intermediate consequence

events are unlikely and high consequence events are highly unlikely. For safe-by-design

components, the staff concludes that the applicant has identified hazards and demonstrated

that failure of safe-by-design components will be highly unlikely.

3.4 EVALUATION FINDINGS

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 failure of the safe-by-design components will be
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highly unlikely and IROFS, management measures, and the applicant’s programs will, if

properly implemented, make all credible intermediate consequence accidents unlikely, and all

credible high consequence accidents highly unlikely. NRC staff finds that the applicant has

adequately demonstrated that safe-by-design components for NCS are highly unlikely, as

required by the regulations.

In the facility SAR, the applicant provided preliminary design basis information for I&C systems

that it identified as IROFS for the facility. The design information is at the system functional

level. Individual components and vendors had not yet been selected. Based on the staff’s

review of the SAR, supporting information provided by the applicant, and the applicant’s

commitments to the industry standards and guidance cited in the sections above for I&C

systems, the staff finds that the preliminary design meets the requirements of 10 CFR 70.61 and

70.64(a)(10). 

Given that these conclusions are based on preliminary design information and the possibility

that the applicant may choose to implement design changes as discussed in the previous

section on I&C, the staff is imposing a license condition to ensure that the final design is

adequate and acceptable to the staff. Specifically, the following condition will be included in the

license:

“Currently, there are no IROFS that have been specified as using software, firmware,

microcode, PLCs, and/or any digital device, including hardware devices which

implement data communication protocols (such as fieldbus devices and Local Area

Network controllers), etc. Should the design of any IROFS be changed to include any of

the preceding features, the licensee shall obtain Commission approval prior to

implementing the change(s). The licensee’s design change(s) shall adhere to accepted

best practices in software and hardware engineering, including software quality

assurance controls as discussed in the in the Quality Assurance Program Description

(LES, 2005a) throughout the development process and the applicable guidance of the

following industry standards and regulatory guides as specified in SAR Chapter 3 (LES,

2005a):

a. American Society of Mechanical Engineers (ASME) NQA-1-1994, Part II, subpart

Part 2.7, “Quality Assurance Requirements of Computer Software for Nuclear

Facility Applications,” as revised by NQA-1a-1995 Addenda of NQA-1-1994 and

ASME NQA-1-1994, Part 1, Supplement 11S-2, “Supplementary Requirements

for Computer Program Testing.” (Refer to SAR Chapter 11, Appendix A, Section

3.)

b. Electric Power Research Institute (EPRI) NP-5652, “Guideline for the Utilization

of Commercial Grade Items in Nuclear Safety Grade Applications,” June 1988.

c. EPRI Topical Report (TR) -102323, “Guidelines for Electromagnetic Interference

Testing in Power Plants,” Revision 1, December 1996.

d. EPRI TR-106439, “Guideline on Evaluation and Acceptance of Commercial

Grade Digital Equipment for Nuclear Safety Applications,” October 1996.
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e. Regulatory Guide 1.152, “Criteria for Digital Computers in Safety Systems in

Nuclear Power Plants,” Revision 1, January 1996.

f. Regulatory Guide 1.168, “Verification, Validation, Reviews, and Audits for Digital

Software Used in Safety Systems of Nuclear Power Plants,” Revision 1, February

2004.

g. Regulatory Guide 1.169, “Configuration Management Plans for Digital Computer

Software Used in Safety Systems of Nuclear Power Plants,” September 1997.

h. Regulatory Guide 1.170, “Software Test Documentation for Digital Computer

Software Used in Safety Systems of Nuclear Power Plants,” September 1997.

I. Regulatory Guide 1.172, “Software Requirements Specifications for Digital

Computer Software Used in Safety Systems of Nuclear Power Plants,”

September 1997.

j. Regulatory Guide 1.173, “Developing Software Life Cycle Processes for Digital

Computer Software Used in Safety Systems of Nuclear Power Plants,”

September 1997.

If any above changes result in IROFS requiring operator actions, the licensee shall

conduct a human factors engineering review of the human-system interfaces using the

applicable guidance in NUREG-0700, “Human-System Interface Design Review

Guidelines,” Revision 2, dated May 2002 (NRC, 2002d), and NUREG-0711, “Human

Factors Engineering Program Review Model,” Revision 2, dated February 2004.”

The staff concludes, pursuant to 10 CFR 70.23(a), that the preliminary design bases of the

IROFS evaluated in SER Section 3.3.3.2.4 will provide reasonable assurance of protection

against natural phenomena and the consequences of potential accidents. Given that these

conclusions are based on preliminary design information, the staff is imposing a license

condition to ensure that the final design is adequate and acceptable to the staff. Specifically,

the following condition will be included in the license:

“The applicant shall utilize its procedure, "IROFS Boundary Definitions," to define the

boundaries of each IROFS. Completed IROFS boundaries for all IROFS shall be

available for inspection at the time of the operational readiness review.”

3.5 REFERENCES

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Safety Related Concrete Structures.” Farmington Hills, Michigan: American Concrete Institute.

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(ACI, 1999) American Concrete Institute (ACI). ACI 318, “Building Code Requirements for

Structural Concrete.” Farmington Hills, Michigan: American Concrete Institute. 1999.

(AIChE, 1989) American Institute of Chemical Engineers (AIChE). “Guidelines for Chemical

Process Quantitative Risk Analysis.” Center for Chemical Process Safety, New York. 1989

(AIChE, 1992) American Institute of Chemical Engineers (IChE). “Guidelines for Hazard

Evaluation Procedures, Second Edition with Worked Examples.” Center for Chemical Process

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(AISC, 1989) American Institute of Steel Construction (AISC). “Allowable Stress Method,”

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(ANSI, 1995) American National Standard Institute (ANSI). ANSI N14.1, “American National

Standard for Nuclear Materials - Uranium Hexafluoride - Packaging for Transport,” 1995.

(ANSI/IEEE, 1971) American National Standard Institute/Institute of Electrical and Electronics

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Generating Stations,” 1971.

(ASCE, 1980) American Society of Civil Engineers (ASCE). ASCE Manuals and Reports on

Engineering Practice No. 58, “Structural Analysis and Design of Nuclear Plant Facilities.” 

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(ASCE, 1998) American Society of Civil Engineers (ASCE). ASCE 7, “Minimum Design Loads

for Buildings and Other Structures.” 1998.

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(Harper, 2003a) Harper, G.A. “Assessment of Other External Event Hazards at NEF for ISA

and Design Basis.” Document Identifier 51-2400552-00, Framatome ANP, Inc., 2003.

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Loads at NEF for ISA and Design Basis—Attachment 2.” Document Identifier 51-2400548-00,

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(Harper, 2003c) Harper, G.A. “Assessment of Winter Precipitation Loads at NEF for ISA and

Design Basis.” Document Identifier 51-2400547-00, Framatome ANP, Inc., 2003.

(ICBO, 1997) International Conference of Building Officials, Uniform Building Code, 1997.

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(LES, 1993) Louisiana Energy Services (LES). “Safety Analysis Report for Claiborne

Enrichment Center (CFC)” (1993)

(LES, 2004a) Louisiana Energy Services (LES), letter to U.S. Nuclear Regulatory Commission,

“Clarifying Information Related to External Hazards Analysis,” August 31, 2004.

(LES, 2004b) Louisiana Energy Services (LES). Letter to U.S. Nuclear Regulatory

Commission, “Clarifying Information Related to Probabilistic Seismic Hazard Analysis,”

November 22, 2004.

(LES, 2005a) Louisiana Energy Services (LES). “National Enrichment Facility Safety Analysis

Report,” Revision 6, 2005.

(LES, 2005b) Louisiana Energy Services (LES). “Integrated Safety Analysis Summary,”

Revision 4, 2005.

(LES, 2005c) Louisiana Energy Services (LES). “National Enrichment Facility Environmental

Report,” Revision 5, 2005.
3-58

(McDonald, 1995) McDonald, J.R. and D. Lu. “A Methodology for Tornado Hazard Probability

Assessment.” Proceeding for the 5th Department of Energy Natural Phenomena Hazard

Mitigation Conference, Denver, Colorado, November 13S14, 1995. 1995.

(McGuire, 1976) McGuire, R.K., “EQRISK, Evaluation of Earthquake Risk to a Site,” U.S.

Geologic Survey Open File Report 76-67, 1976.

(NAESC, 1999) North Atlantic Energy Service Corporation (NAESC). “Seabrook Station

Updated Final Safety Analysis Report,” Docket No. 50-443, 1999.

(NAVFAC, 1986) Naval Facilities Engineering Command Design Manual NAVFAC DM-7.01,

“Soil Mechanics,” 1986.

(NAVFAC, 1996). Naval Facilities Engineering Command Design Manual NAVFAC DM-7.02,

“Foundations and Earth Structures,” 1996.

(NCDC, 2005) National Climatic Data Center (NCDC). "Storm Events."

<http://www4.ncdc.noaa.gov/cgi-win/wwcgi.dll?wwevent~storms> (Accessed 3/17/05).

(NCEDC, 2004) Northern California Earthquake Data Center (NCEDC). “Composite

Earthquake Catalog,” 2004, <http://quake.geo.berkeley.edu/anss/>

(NFPA, 1991) National Fire Protection Association (NFPA). Fire Protection HandbookTM, 7th

Edition, 1991.

(NFPA, 1995) National Fire Protection Association (NFPA) and Society of Fire Protection

Engineers. Society of Fire Protection Engineers Handbook of Fire Protection Engineering, 2nd

Edition, 1995.

(NMIMT, 2003) New Mexico Institute of Mining and Technology (NMIMT). “Earthquake

Catalogs for New Mexico and Bordering Areas: 1869–1998,” 2003.

(NOAA, 1982) National Oceanic and Atmospheric Administration (NOAA). “Application of

Probable Maximum Precipitation Estimates—United States East of the 105th Meridian,”

Hydrometeorological Report No. 52, 1982.

(NOAA, 2005) National Oceanic and Atomospheric Administration. "Tornadoes."

<http://www.noaa.gov/tornadoes.html> (Accessed 3/17/05).

(NRC, 1973) U.S. Nuclear Regulatory Commission (NRC). Regulatory Guide 1.60, “Design

Response Spectra for Seismic Design of Nuclear Power Plants,” 1973.

(NRC, 1975) U.S. Nuclear Regulatory Commission (NRC). “Site Analysis Branch

Position—Winter Precipitation Loads,” 1975.

(NRC, 1978a) U.S. Nuclear Regulatory Commission (NRC). NUREG/CR–0098, “Development

of Criteria for Seismic Review of Selected Nuclear Power Plants,” 1978.
3-59

(NRC, 1978b) U.S. Nuclear Regulatory Commission (NRC). Regulatory Guide 1.91,

“Evaluation of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power

Plants,” Rev. 1, 1978.

(NRC, 1981) U.S. Nuclear Regulatory Commission (NRC). NUREG–0800, “Standard Review

Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,” 1981.

(NRC, 1983) U.S. Nuclear Regulatory Commission (NRC). NUREG/CR–3058, “A Methodology

for Tornado Hazard Probability Assessment,” 1983.

(NRC, 1991) U.S. Nuclear Regulatory Commission (NRC). NUREG-1391, “Chemical Toxicity

of Uranium Hexafluoride Compared to Acute Effects of Radiation.” NRC: Washington, D.C. 

February 1991.

(NRC, 1994) U.S. Nuclear Regulatory Commission (NRC). NUREG-1491, “Safety Evaluation

Report for the Claiborne Enrichment Center, Homer, Louisiana.” NRC: Washington, D.C. 

January 1994.

(NRC, 1997) U.S. Nuclear Regulatory Commission (NRC). Regulatory Guide 1.165,

“Identification and Characterization of Seismic Sources and Determination of Safe Shutdown

Earthquake Ground Motion,” 1997.

(NRC, 1999a) U.S. Nuclear Regulatory Commission (NRC). Regulatory Guide 1.105,

“Setpoints for Safety-Related Instrumentation,” Revision 3, dated December 1999.

(NRC, 1999b) U.S. Nuclear Regulatory Commission (NRC). Regulatory Guide 1.145,

“Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear

Power Plants,” Revision 1, dated November 1982.

(NRC, 2001) U.S. Nuclear Regulatory Commission (NRC). NUREG-1513, “Integrated Safety

Analysis Guidance Document.” NRC: Washington, D.C. May 2001.

(NRC, 2002a) U.S. Nuclear Regulatory Commission (NRC). NUREG-1520, “Standard Review

Plan for the Review of a License Application for a Fuel Cycle Facility.” NRC: Washington, D.C. 

March 2002.

(NRC, 2002b) U.S. Nuclear Regulatory Commission (NRC). NUREG-6410, “Nuclear Fuel

Cycle Facility Accident Analysis Handbook.” NRC: Washington, D.C. March 1998.

(NRC, 2002c) U.S. Nuclear Regulatory Commission (NRC). NUREG-0700, “Human-System

Interface Design Review Guidelines,” NRC: Washington, D.C. Revision 2, 2002.

(NRC, 2003a) U.S. Nuclear Regulatory Commission (NRC). Regulatory Guide 1.180,

“Guidelines for Evaluating Electromagnetic and Radio-Frequency Interference in Safety-Related

Instrumentation and Control Systems,” Revision 1, 2003.

(NRC, 2003b) U.S. Nuclear Regulatory Commission (NRC). Regulatory Guide 1.198,

“Procedures and Criteria for Assessing Seismic Soil Liquefaction at Nuclear Power Plant Sites,”

2003.
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2, 2004.

(NRC, 2004b) Graves, H.L., U.S. Nuclear Regulatory Commission (NRC). Summary of

Meeting with Louisiana Energy Services on Structural, Seismic, High Winds, and Tornado

Safety Analysis. June 29, 2004.

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Document Identifier 32-2400573-00, Framatome ANP, Inc., 2003.

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Document Identifier 32-400572-00, Framatome ANP, Inc., 2004.

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Motions from Earthquakes in Central and Eastern North America: Best Estimates and

Uncertainties,” Seismological Research Letters, Vol. 68, No. 1, pp. 41S5, 1997.

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Earthquake for the Central United States,” Miscellaneous Paper S–731–1, 1973.

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Atmospheric Administration. “ALOHA—(Areal Locations of Hazardous Atmospheres) User’s

Manual,” 1999.

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2004. <http://eqhazmaps.usgs.gov/>

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Precipitation East of the 105th Meridian for Areas from 10 to 1,000 Square Miles and Duration of

6, 12, 24, and 48 Hours,” Hydrometeorological Report No. 33, 1956.

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Earthquakes,” 2002.

(Winterkorn, 1975) Winterkorn and Fang, “Foundation Engineering Handbook,” 1975.
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4.0 RADIATION PROTECTION

The purpose of this review is to determine whether the applicant’s radiation protection (RP)

program is adequate to protect the radiological health and safety of workers and to comply with

the associated regulatory requirements in 10 CFR Parts 19, 20, 30, 40, and 70. Public and

environmental protection is discussed in Chapter 9 of this Safety Evaluation Report.

4.1 REGULATORY REQUIREMENTS

4.1.1 Radiation Protection Program Implementation

Regulations applicable to establishment of an RP program are presented in Part 20, Subpart B,

“Radiation protection programs.”

4.1.2 As Low As is Reasonably Achievable (ALARA) Program

Regulations applicable to the ALARA program are presented in 10 CFR 20.1101, “Radiation

protection programs.”

4.1.3 Organization and Personnel Qualifications

Regulations applicable to the organization and qualifications of the radiological protection staff

are presented in 10 CFR 30.33, 10 CFR 40.32, and 10 CFR 70.22, “Contents of applications.”

4.1.4 Written Procedures

The regulations applicable to RP procedures and Radiation Work Permits (RWPs) are

presented in 10 CFR 30.33, 10 CFR 40.32, and 10 CFR 70.22, “Contents of applications.”

4.1.5 Training

The following regulations apply to the radiation safety training program:

1. 10 CFR 19.12 “Instructions to workers”

2. 10 CFR 20.2110 “Form of records”

4.1.6 Ventilation and Respiratory Protection Programs

Regulations applicable to the ventilation and respiratory protection programs are presented in

Part 20, Subpart H, “Respiratory protection and controls to restrict internal exposure in restricted

areas.”
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4.1.7 Radiation Survey and Monitoring Programs

The following NRC regulations in Part 20 are applicable to radiation surveys and monitoring

programs:

1. Subpart F “Surveys and Monitoring”

2. Subpart C “Occupational Dose Limits”

3. Subpart L “Records”

4. Subpart M “Reports”

4.1.8 Additional Program Requirements

The following Part 20 regulations are applicable to the additional program requirements:

1. Subpart L “Records”

2. Subpart M “Reports”

3. Section 70.61 ”Performance requirements”

4. Section 70.74 ”Additional reporting requirements”

4.2 REGULATORY ACCEPTANCE CRITERIA

The acceptance criteria for NRC’s review of the RP program are outlined in 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 “Standard Review Plan for the

Review of a License Application for a Fuel Cycle Facility,” NUREG-1520 (NRC, 2002).

4.3 STAFF REVIEW AND ANALYSIS

4.3.1 Radiation Protection Program Implementation

In Section 4.1 of the Louisiana Energy Services’ (LES) Safety Analysis Report (SAR) (LES,

2005a), the applicant describes the proposed RP program for the proposed facility. The RP

program is developed, documented, and will be implemented commensurate with the risk posed

by a uranium enrichment operation that will meet the requirements of 10 CFR Part 20, Subpart

B.

The RP program’s organizational structure and the responsibilities of key program personnel

are outlined in Section 4.1.1 of the SAR (LES, 2005a). The Plant Manager will be responsible

for the protection of all persons against radiation exposure resulting from facility operations and

material, and for compliance with applicable NRC regulations and the facility license. The RP

Manager will be responsible for implementing the RP program, and in matters involving RP, will 
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have direct access to the Plant Manager. The RP Manager and his staff will also be

responsible for:

• Establishing the RP program;

• Generating and maintaining procedures associated with the program;

• Assuring that ALARA is practiced by all personnel;

• Reviewing and auditing the efficacy of the program in complying with NRC and other

governmental regulations and applicable Regulatory Guides;

• Modifying the program, based on experience and facility history;

• Adequately staffing the RP group to implement the RP program;

• Establishing and maintaining the ALARA program;

• Establishing and maintaining a respirator usage program;

• Monitoring worker doses, both internal and external;

• Complying with the radioactive materials possession limits for the facility;

• Handling of radioactive wastes when disposal is needed;

• Calibration and quality assurance of all radiological instrumentation, including verification

of required Lower Limits of Detection or alarm levels;

• Establishing and maintaining a radiation safety training program for personnel working in

Restricted Areas;

• Performing audits of the RP program on an annual basis;

• Establishing and maintaining the radiological environmental monitoring program; and

• Posting the Restricted Areas, and within these areas, posting: Radiation, Airborne

Radioactivity, High Radiation, and Contaminated Areas, as appropriate; and developing

occupancy guidelines for these areas as needed.

The responsibilities of the Health, Safety, and Engineering (HS&E) Manager, the Operations

Manager, and the facility personnel are outlined in Sections 4.1.1.2, 4.1.1.4, and 4.1.1.5 of the

SAR (LES, 2005a), respectively.

The applicant will staff the facility with suitably trained RP personnel. The staff will be trained

and qualified consistent with the guidance provided in American National Standards

Institute/American Nuclear Society (ANSI/ANS) Standard 3.1, “Selection, Qualification and

Training of Personnel for Nuclear Power Plants” (ANSI/ANS, 1993). The applicant will provide

sufficient resources in terms of staffing and equipment to implement an effective RP program.
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The applicant will ensure that the RP program will remain independent of the facility’s routine

operations, and that it maintains its objectivity and is focused only on implementing sound RP

principles necessary to achieve ALARA goals.

The applicant will review the content and implementation of the RP program at least annually, in

accordance with 10 CFR 20.1101(c). In addition, constraints on atmospheric releases are

established to ensure compliance with 10 CFR 20.1101(d).

As described above, the applicant will maintain the RP program in accordance with the

acceptance criteria in NUREG-1520 (NRC, 2002).

4.3.2 ALARA Program

The ALARA program will be implemented using written policies and procedures, to ensure

occupational radiation exposures are maintained ALARA and that such exposures are

consistent with the requirements of 10 CFR 20.1101.

The applicant states that the RP Manager will be responsible for implementing the ALARA

program and preparing an ALARA program evaluation report annually, to review:

• Radiological exposure and effluent release data for trends;

• Audits and inspections;

• Use, maintenance, and surveillance of equipment used for exposure and effluent control;

and

• Other issues, as appropriate, that may influence the effectiveness of the radiation

protection/ALARA programs.

This report will be submitted to the Plant Manager and the Safety Review Committee (SRC).

Radiological zones, such as “Radiation Area,” “Airborne Radioactivity Area,” and “Contaminated

Area,” will be established, as necessary, to minimize the spread of contamination and reduce

unnecessary radiation exposure to personnel.

Goals of the ALARA program include maintaining occupational exposures, as well as

environmental releases, as far below regulatory limits as is reasonably achievable. This is

accomplished by minimizing the size and number of areas with higher dose rates and designing

areas to maintain the lowest dose rates reasonably achievable, particularly where facility

personnel spend significant amounts of time.

The applicant will establish an SRC, which meets at least quarterly and within 30 days of any

NRC-reportable incident, and will fulfill the duties of the ALARA Committee. The SRC will have

at least five members, to include experts in operations, criticality safety, radiological safety,

chemical safety, and industrial safety.

Responsibilities of the SRC include: (1) reviewing the effectiveness of the ALARA program; (2)

determining if exposures, releases, and contamination levels are in accordance with the ALARA
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concept; (3) ensuring that the occupational radiation exposure dose limits of Part 20 are not

exceeded under normal operations; and (4) identifying any upward trends in personnel

exposures, environmental releases, and facility contamination levels.

The applicant will use the ALARA program to facilitate interaction between RP and operations

personnel. The SRC reviews and revises the program goals and objectives to incorporate,

when appropriate, new approaches, technologies, operating procedures, or changes that could

cost-effectively reduce potential radiation exposure.

The applicant will maintain an ALARA program in accordance with the acceptance criteria as

described above.

4.3.3 Organization and Personnel Qualifications

The applicant will employ only suitably trained RP personnel at the facility, by following the

guidance in Regulatory Guides 8.2 and 8.10 (NRC, 1973 and NRC, 1997). Further information

on personnel qualifications and training is provided in Sections 2.24, 2.3.3, and 11.3 of the SAR

(LES, 2005a). The applicant will assign the responsibility of implementing the RP program

functions to the RP program staff.

The RP Manager will be responsible for establishing and implementing the RP program, which

includes training personnel in use of equipment, control of radiation exposure of personnel,

continuous determination and evaluation of the radiological status of the facility, and conducting

the radiological environmental monitoring program. The RP Manager will have direct access to

the Plant Manager regarding all matters involving RP; will be skilled in the interpretation of RP

data and regulations; will be familiar with the operation of the facility and RP concerns of the

site; and will be a resource in radiation safety management decisions.

The RP Manager will have, as a minimum, a bachelor’s degree (or equivalent) in an engineering

or scientific field and 3 years of responsible nuclear experience associated with the

implementation of an RP program, to include at least 2 years experience at a facility that

processes uranium, including uranium in soluble form. The RP staff will be trained and qualified

consistent with guidance provided in ANSI/ANS Standard 3.1 (ANSI/ANS, 1993).

The applicant will organize and staff an RP program in accordance with the acceptance criteria.

4.3.4 Written Procedures

Written procedures will be used for all operations involving licensed materials. The procedures

are prepared, reviewed, and approved to carry out activities related to the RP program, and are

used to ensure RP activities are conducted in a safe, effective, and consistent manner. The RP

procedures are reviewed and revised as necessary, to incorporate any facility or operational

changes to the facility’s Integrated Safety Analysis (ISA).

RP procedures are prepared by RP staff; reviewed by members of the facility staff, personnel

with enrichment plant operating experience, and other staff members, as appropriate; and

approved by the RP Manager, or designee.
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The applicant will perform all work in Restricted Areas in accordance with an RWP. The

applicant will also issue RWPs for activities involving licensed materials not covered by

operating procedures, and where radioactivity levels are likely to exceed airborne radioactivity

limits, or whenever deemed as necessary by the RP Manager.

RWPs provide a description of the work or authorized activities; summary results of dose rate,

contamination, and airborne radioactivity surveys, etc.; and precautions, which may include

personnel protective equipment, stay-times or dose limits, record-keeping requirements, and

coverage by an RP technician.

RWPs require approval by the RP Manager or designee, and have a predetermined period of

validity with a specified expiration or termination time. The designee must meet the

requirements of Section 4.1.2 of the SAR (LES, 2005a).

The following criteria are used for RWPs:

• Planned activities or changes to activities inside Restricted Areas or work with licensed

materials are reviewed by the RP Manager or designee, for the potential to cause

radiation exposures to exceed action levels or to produce radioactive contamination;

• RWPs include requirements for any necessary safety controls, personnel monitoring

devices, protective clothing, respiratory protective equipment, and air sampling

equipment, and the attendance of RP technicians at the work location;

• RWPs are posted at access points to Restricted Areas, with copies of current RWPs

posted at the work area location;

• RWPs clearly define and limit the work activities to which they apply. An RWP is closed

out when the applicable work activity for which it was written is completed and

terminated; and

• RWPs are retained as a record at least for the life of the facility.

The applicant will prepare written procedures and RWPs in accordance with these acceptance

criteria.

4.3.5 Training

An RP training program is designed and implemented to provide training to all personnel and

visitors, unless provided with trained escorts, who enter Restricted Areas or Controlled Areas,

commensurate with the radiological hazard to which they may be exposed. The level of training

is based on the potential radiological health risks associated with the individual’s work

responsibilities.

The applicant has incorporated the provisions of 10 CFR 19.12 into the radiation training

program, as outlined in Section 4.5.1 of the SAR (LES, 2005a). The requirements in 10 CFR

19.12 address required health physics information the applicant must make available to workers

likely to receive exposures greater than 1 mSv (100 mrem) per year.
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Retraining is performed for personnel requiring unescorted access to Restricted Areas on an

annual basis, and as necessary, to address changes in policies, procedures, requirements, and

the facility ISA. The HS&E Manager and RP Manager review and update the content of the

formal RP training program at least every 2 years, to ensure that the programs are current and

adequate.

The applicant will evaluate the effectiveness and adequacy of the training program curriculum

and instructors through audits performed by operational area personnel responsible for criticality

safety and RP.

The applicant will train its employees in RP in accordance with the acceptance criteria.

4.3.6 Ventilation and Respiratory Protection Programs

The design criteria, including flow velocity at openings, for the ventilation systems, are

described in Sections 3.4.9 and 3.5.1 of the ISA Summary (LES, 2005b). Filters to be used in

the systems include pre-filters for dust removal, High-Efficiency Particulate Air (HEPA) filters for

removal of uranyl fluoride (UO2F2) aerosols, and activated carbon filters for hydrogen fluoride

(HF) removal. Differential pressure across HEPA filters in potentially contaminated exhaust

systems is either checked monthly or automatically monitored and alarmed, and filters are

replaced when differential pressure exceeds the manufacturers’ ratings or fail to function

properly.

The containment of uranium hexafluoride (UF6) is maintained by the process equipment, and

the enrichment process, with the exception of liquid sampling, is maintained under a partial

vacuum, to ensure any leaks will not result in a release of UF6 into the work areas.

Air-flow rates at system openings (hoods, exhausted enclosures, close-capture point

equipment, or ventilation systems serving these barriers) will be sufficient to preclude escape of

airborne uranium and minimize the potential for worker intake, and will be checked monthly

when in use and after any modifications.

To meet the respiratory protection requirements in 10 CFR Part 20, Subpart H, the applicant will

prepare written procedures for the selection, fitting, issuance, maintenance, testing, training of

personnel, monitoring, and record-keeping for individual respiratory protection equipment, in

accordance with 10 CFR 20.1703(c)(4). The applicant has established a respiratory protection

program that meets the requirements of Part 20, Subpart H.

The applicant will revise respiratory protection procedures as necessary, whenever changes are

made to the facility, process, or equipment. The records of the respiratory protection program,

including training for respirator use and maintenance, are maintained in accordance with the

facility records management program, which is described in Section 11.7 of the SAR (LES,

2005a).

The applicant has established ventilation and respiratory protection programs in accordance

with the acceptance criteria, and satisfies the regulatory requirements of Part 20, Subpart H.
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4.3.7 Radiation Survey and Monitoring Programs

The applicant has a radiation survey and monitoring program using prepared written procedures

that will include an outline of the program objectives; sampling procedures; data analysis

methods; types of equipment and instrumentation to be used; frequency of measurements;

record-keeping and reporting requirements; and actions to be taken when measurements

exceed Part 20 occupational dose limits, or the administrative levels established by the

applicant.

Thermoluminescent dosimeters (TLDs) that are sensitive to beta, gamma, and neutron radiation

- supplied by a vendor holding dosimetry accreditation from the National Voluntary Laboratory

Accreditation Program - will be required to be worn by all personnel who enter Restricted Areas. 

The TLDs are evaluated at least quarterly.

The applicant has established an annual administrative limit for Total Effective Dose Equivalent

of 10 millisieverts (mSv) [1000 millirem (mrem)]. If 2.5 mSv (250 mrem) are exceeded in any

quarter, an investigation will be performed and documented to determine what factors

contributed to that exposure, including procedural reviews, efficiency studies of the air-handling

system, cylinder storage protocol, and work practices.

All personnel wearing external dosimetry devices will be evaluated for internal exposures via

direct bioassay, indirect bioassay, or an equivalent technique. These doses will be evaluated at

least annually. Bioassay will be performed for all personnel who are likely to have had an intake

of 1 milligram (mg) of uranium during a week (10 percent of the 10 mg in-a-week NRC limit)

based on air-sample monitoring data. Bioassay will be performed using urinalysis having a

sensitivity of 5 micrograms per liter (μg/l), assuming proper timing and amount of the sample. 

Workers will be restricted from activities that could result in internal exposures to soluble

uranium, as long as uranalysis results are $15 μg/l.

The applicant will sum the internal and external exposure values in accordance with 10 CFR

20.1202, using procedures based on the guidance in Regulatory Guides 8.7 (NRC, 1992a) and

8.34 (NRC, 1992c).

The applicant will perform air sampling consistent with the guidance provided in Regulatory

Guide 8.25 (NRC, 1992b); NUREG-1400 (NRC, 1991); and ANSI/Health Physics Society (HPS)

Standard 13.1 (ANSI/HPS, 1999). Regulatory Guide 8.25 (NRC, 1992b) and NUREG-1400

(NRC, 1991) provide guidance on evaluating the need for air sampling; locating air samplers;

demonstrating that air sampling is representative of inhaled air; adjustments to derived air

concentrations; measuring the volume of air sampled; and evaluating sample results, including

the calculation of minimum detectable activity for samples. ANSI/HPS 13.1 (ANSI/HPS, 1999)

provides guidance on sampling location, sampling system design, and quality assurance and

quality control.

Airborne activity levels in the Restricted Areas of the facility will be continuously monitored, with

permanent air monitors designed to detect alpha emitters. Monitors located in Airborne

Radioactivity Areas will have alarms, which will be set with consideration of both toxicity and

radioactivity. Portable air samplers will also be available for use.
4-9

Airflow and radioactivity measurement instruments will be calibrated at least annually or

according to manufacturers’ recommendations, after failing an operability check, after

modifications or repairs that could affect proper response, or when believed to have been

damaged. The applicant will use calibration sources traceable to the National Institute of

Standards and Technology, or equivalent, and which are ±5 percent of the stated values.

The applicant will provide routine contamination survey monitoring in all UF6 process areas, with

routine, periodic checks of non-UF6 process areas, including those areas that are normally free

of contamination. Restricted Areas will be surveyed at least weekly, and lunch rooms and

change rooms will be surveyed daily. Monitoring will include measurements of fixed and

removable surface contamination, with extent and frequency based on the potential for

contamination in each area and operational experience. Removable surface contamination will

be considered to be uranium that can be transferred to a dry smear paper with moderate

pressure.

The applicant has defined a contaminated area as an area where removable contamination

levels are above 0.33 Bq/100 cm2

 (20 dpm/100 cm2) alpha, or 16.7 Bq/100 cm2 (1000 dpm/100

cm2

) beta/gamma, and will initiate clean-up of contamination within 24 hours of detection, if

levels exceed 83.3 Bq/100 cm2

 (5000 dpm/100 cm2) alpha or beta/gamma for removable

contamination, and 4.2 kBq/100 cm2

 (250,000 dpm/100 cm2) alpha or beta/gamma for fixed

contamination.

The facility corrective action process, which is described in Section 11.6 of the SAR (LES,

2005a), will be implemented if:

1. Personnel dose-monitoring results or personnel contamination levels exceed the

administrative personnel limits;

2. An incident results in airborne occupational exposures exceeding the administrative

limits; or

3. The dose limits in Part 20, Appendix B, or 10 CFR 70.61 are exceeded.

The applicant has a radiation survey program consistent with the guidance contained in

Regulatory Guide 8.24, “Health Physics Surveys During Enriched Uranium-235 Processing and

Fuel Fabrication” (NRC, 1979), which includes the selection of instruments that should be

operable and capable of measuring, at or below the required level, the types of radiation that

will be encountered. These instruments will be calibrated as discussed previously in this

section.

For transfer of material and equipment to unrestricted areas and release from the facility for

unrestricted use, the applicant will meet surface contamination guidelines prescribed in

“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,

1993a).

The applicant will leak-test sources in accordance with the following NRC Branch Technical

Positions:
4-10

1. “License Condition for Leak-Testing Sealed Byproduct Material Sources” (NRC, 1993b);

2. “License Condition for Leak-Testing Sealed Source Which Contains Alpha and/or BetaGamma Emitters” (NRC, 1993c); and

3. “License Condition for Leak-Testing Sealed Uranium Sources” (NRC, 1993d).

Sources containing plutonium will not be used at the facility.

The applicant has an access control program that ensures that: (a) signs, labels, and other

access controls are properly posted and operative; (b) restricted areas are established to

prevent the spread of contamination and are identified with appropriate signs; and (c) step-off

pads, change facilities, protective clothing facilities, and personnel monitoring instruments are

provided in sufficient quantities and locations.

The applicant has established action levels of 2.5 Bq/100cm2

 (150 dpm/100cm2) alpha or

beta/gamma contamination (corrected for background) at points of egress from Restricted Areas

and any additional designated areas within the Restricted Area.

The applicant has established radiation survey and monitoring programs in accordance with the

acceptance criteria.

4.3.8 Additional Program Requirements

The applicant has established a program to maintain records of the RP program, radiation

survey results, and results of corrective action program referrals, RWPs, and planned special

exposures.

The applicant will report, to NRC, any event that results in an occupational exposure to radiation

exceeding the dose limits in Part 20, within the time specified in 10 CFR 20.2202, 10 CFR

30.50, 10 CFR 40.60, and 10 CFR 70.74. The applicant will prepare and submit, to NRC, an

annual report of the results of individual monitoring, as required by 10 CFR 20.2206(b).

The applicant will refer to the facility’s corrective action program any radiation incident that

results in an occupational exposure that exceeds the dose limits in Part 20, Appendix B, or is

required to be reported per 10 CFR 30.50, 10 CFR 40.60, and 10 CFR 70.74, and to reporting,

to NRC, both the corrective actions taken (or planned) to protect against a recurrence, and the

proposed schedule to achieve compliance.

The staff reviewed the ISA Summary (LES, 2005b) and performed an on-site review of the ISA,

and agree with the applicant that, as stated in Section 3.6 of the ISA Summary (LES, 2005b),

the hazards from radioactivity were evaluated in the ISA and found to be low consequence.

4.4 EVALUATION FINDINGS
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The applicant has established and will maintain an acceptable RP program that includes:

1. An effective documented program to ensure that occupational radiological exposures are

ALARA;

2. An organization with adequate qualification requirements for the RP personnel;

3. Approved written RP procedures and RWPs for RP activities;

4. RP training for all personnel who have access to restricted areas;

5. A program to control airborne concentrations of radioactive material with engineering

controls and respiratory protection;

6. A radiation survey and monitoring program that includes requirements for controlling

radiological contamination within the facility and monitoring of external and internal

radiation exposures; and

7. Other programs to maintain records, report to NRC in accordance with Parts 20 and 70,

and correct for upsets at the facility.

The applicant’s RP program meets the requirements of Parts 19, 20, 30, 40, and 70. 

Conformance to the license application will ensure safe operation.

4.5 REFERENCES

(ANSI/HPS, 1999) American National Standards Institute/Health Physics Society (ANSI/HPS). 

ANSI/HPS 13.1, “Sampling and Monitoring Releases of Airborne Radioactive Substances From

the Stacks and Ducts of Nuclear Facilities,” 1999.

(ANSI/ANS, 1993) American National Standards Institute/American Nuclear Society

(ANSI/ANS). ANSI/ANS 3.1, “Selection, Qualification, and Training of Personnel for Nuclear

Power Plants,” 1993.

(LES, 2005a) Louisiana Energy Services (LES). “National Enrichment Facility Safety Analysis

Report,” Revision 6, 2005.

(LES, 2005b) Louisiana Energy Services (LES). “National Enrichment Facility Integrated

Safety Analysis Summary,” Revision 4, 2005.

(NRC, 1973) U.S. Nuclear Regulatory Commission (NRC). Regulatory Guide 8.2, “Guide for

Administrative Practice in Radiation Monitoring,” 1973.

(NRC, 1977) U.S. Nuclear Regulatory Commission (NRC). Regulatory Guide 8.10, “Operating

Philosophy for Maintaining Occupational Radiation Exposures As Low As Is Reasonably

Achievable,” 1977.
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(NRC, 1979) U.S. Nuclear Regulatory Commission (NRC). Regulatory Guide 8.24, “Health

Physics Surveys During Enriched Uranium-235 Processing and Fuel Fabrication,” 1979.

(NRC, 1991) U.S. Nuclear Regulatory Commission (NRC). NUREG-1400, “Air Sampling in the

Workplace,” 1991.

(NRC, 1992a) U.S. Nuclear Regulatory Commission (NRC). Regulatory Guide 8.7, 

“Instructions for Recording and Reporting Occupational Radiation Exposure Data,” 1992.

(NRC, 1992b) U.S. Nuclear Regulatory Commission (NRC). Regulatory Guide 8.25, “Air

Sampling in the Workplace,” 1992.

(NRC, 1992c) U.S. Nuclear Regulatory Commission (NRC). Regulatory Guide 8.34,

“Monitoring Criteria and Methods to Calculate Occupational Radiation Doses,” 1992.

(NRC, 1993a) U.S. Nuclear Regulatory Commission (NRC). Branch Technical Position,

“Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted

Use or Termination of Licenses for Byproduct, Source, or Special Nuclear Material,” 1993.

(NRC, 1993b) U.S. Nuclear Regulatory Commission (NRC). Branch Technical Position,

“License Condition for Leak-Testing Sealed Byproduct Material Sources,” 1993.

(NRC, 1993c) U.S. Nuclear Regulatory Commission (NRC). Branch Technical Position,

“License Condition for Leak-Testing Sealed Source Which Contains Alpha and/or Beta-Gamma

Emitters,” 1993.

(NRC, 1993d) U.S. Nuclear Regulatory Commission (NRC). Branch Technical Position,

“License Condition for Leak-Testing Sealed Uranium Sources,” 1993.

(NRC, 2002) U.S. Nuclear Regulatory Commission (NRC). NUREG-1520, “Standard Review

Plan for the Review of a License Application for a Fuel Cycle Facility,” 2002.
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5.0 NUCLEAR CRITICALITY SAFETY

The purpose of this review is to determine whether the applicant’s nuclear criticality safety

(NCS) program is adequate to support safe design, construction, and operation of the facility, as

required by 10 CFR Part 70. In addition, the purpose of this review is to determine whether the

Integrated Safety Analysis (ISA) and ISA Summary meet the regulatory requirements specified

in 10 CFR Part 70, Subpart H, “Additional Requirements for Certain Licensees Authorized to

Possess a Critical Mass of Special Nuclear Material,” for NCS.

The NCS programmatic review will determine whether: (1) the applicant has provided for the

appropriate management of the NCS program; (2) the applicant has identified, and committed

to, the responsibilities and authorities of individuals for developing and implementing the NCS

program; (3) the facility management measures described in 10 CFR 70.62 have been

committed to and will support implementing and maintaining the NCS program; and (4) an

adequate NCS program is described, which includes identifying and committing to the NCS

methods and NCS technical practices used to ensure the safe operation of the facility, as

required by Part 70.

The NCS ISA review was performed to determine whether: (1) the ISA program is acceptable

for NCS; (2) the ISA has been acceptably performed and will be maintained for NCS; and 

(3) the ISA Summary contains necessary information, such that the NCS accident sequences

are “highly unlikely” or else failures of NCS safe-by-design components meet “highly unlikely”

with significant margin.

5.1 REGULATORY REQUIREMENTS

The NCS review of the applicant’s NCS program should verify if the information the applicant

provided (LES, 2004; LES, 2005a; LES, 2005b; and LES, 2005c) meets the requirements of 

10 CFR 70.22 and 70.65, which, respectively, specify the general and additional content of a

license application. In addition, the NCS review should verify compliance with the regulatory

requirements in 10 CFR 70.24; 70.52; 70.61; 70.62; 70.64; 70.65; 70.72; and Appendix A to

Part 70.

The NCS review of the applicant’s ISA program and ISA Summary should verify if the

information the applicant provided (LES, 2005a; LES, 2005c) meets the requirements of 

10 CFR 70.62 and 70.65, which, respectively, specify: (1) the requirements for establishing and

maintaining a safety program (10 CFR 70.62), including an ISA program that addresses NCS;

(2) requirements for conducting and maintaining an ISA (10 CFR 70.62(c)) for NCS; and 

(3) requirements for the contents of an ISA Summary (10 CFR 70.65(b)) for NCS.

5.2 REGULATORY ACCEPTANCE CRITERIA
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The acceptance criteria for the NCS review of the applicant’s NCS program are outlined in

Sections 5.4.3.1, 5.4.3.2, 5.4.3.3, and 5.4.3.4 of NUREG-1520 (NRC, 2002). This includes the

commitment to use NRC NCS Regulatory Guide 3.71 (NRC, 1998), which modified the use of

the American National Standards Institute/American Nuclear Society (ANSI/ANS) Series-8 NCS

standards.

The acceptance criteria used for the NCS review of the applicant’s ISA program and ISA

Summary are outlined in Sections 3.4.3.1 and 3.4.3.2 of NUREG-1520 (NRC, 2002).

5.3 STAFF REVIEW AND ANALYSIS

5.3.1 Possession Limits

In Chapter 1 of the Safety Analysis Report (SAR) (LES, 2005a), the applicant described the

NCS-important special nuclear material possession limits of the facility, as shown in Section

1.2.3.4 of this Safety Evaluation Report (SER), at an enrichment level of no greater than 

5.0 weight (wt) percent, in the chemical form of uranium hexafluoride (UF6), uranium

tetrafluoride (UF4), and uranyl fluoride (UO2F2) in the physical form of solid, liquid, and gas.

5.3.2 NRC NCS Regulatory Guide 3.71 and ANSI/ANS Series 8 NCS Standards

In Chapter 5 of the SAR (LES, 2005a), the applicant committed to following NRC NCS

Regulatory Guide 3.71 (NRC, 1998) and to use the following ANSI/ANS Series-8 NCS

standards:

• ANSI/ANS-8.1, "Nuclear Criticality Safety in Operations with Fissionable Materials

Outside Reactors" (ANSI/ANS, 1988a);

• ANSI/ANS-8.1, "Nuclear Criticality Safety in Operations with Fissionable Materials

Outside Reactors" (ANSI/ANS, 1998a);

• ANSI/ANS-8.3, “Criticality Accident Alarm System” (ANSI/ANS, 1997a);

• ANSI/ANS-8.5, “Use of Borosilicate-Glass Raschig Rings as a Neutron Absorber in

Solutions of Fissile Material” (ANSI/ANS, 1996a);

• ANSI/ANS-8.7, “Guide for Nuclear Criticality Safety Criteria in the Storage of Fissile

Materials” (ANSI/ANS, 1998b);

• ANSI/ANS-8.6, “Safety in Conducting Subcritical Neutron-Multiplication Measurements

In Situ” (ANSI/ANS, 1995a);

• ANSI/ANS-8.10, “Criteria for Nuclear Criticality Safety Controls in Operations with

Shielding and Confinement” (ANSI/ANS, 1988b);

• ANSI/ANS-8.12, “Nuclear Criticality Control and Safety of Plutonium-Uranium Fuel

Mixtures Outside Reactors“ (ANSI/ANS, 1993);
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• ANSI/ANS-8.15, “Nuclear Criticality Control of Special Actinide Elements” (ANSI/ANS,

1995c);

• ANSI/ANS-8.17, “Criticality Safety Criteria for the Handling, Storage, and Transportation

of LWR Fuel Outside Reactors” (ANSI/ANS, 1997b);

• ANSI/ANS-8.19, “Administrative Practices for Nuclear Criticality” (ANSI/ANS, 1996b);

• ANSI/ANS-8.20, “Nuclear Criticality Safety Training” (ANSI/ANS, 1991);

• ANSI/ANS-8.21, “Use of Fixed Neutron Absorbers in Nuclear Facilities Outside

Reactors” (ANSI/ANS, 1995d);

• ANSI/ANS-8.22, “Nuclear Criticality Safety Based on Limiting and Controlling

Moderators” (ANSI/ANS, 1997c); and

• ANSI/ANS-8.23, “Nuclear Criticality Accident Emergency Planning and Response”

(ANSI/ANS, 1997d).

The applicant took exception to the ANSI/ANS-8.9 standard (ANSI/ANS, 1995b), in that piping

configurations containing aqueous solutions of fissile material will be evaluated in accordance

with the 1998 version of ANSI/ANS-8.1 (ANSI/ANS, 1998a), using validated methods to

determine subcritical limits.

In addition, the applicant used a newer version of the ANSI/ANS-8.1 standard (ANSI/ANS,

1998a) than the version of the ANSI/ANS-8.1 standard (ANSI/ANS, 1988a) that the NRC

endorsed, with exception, in NRC NCS Regulatory Guide 3.71 (NRC, 1998). NRC staff

reviewed the differences between the two versions of ANSI/ANS-8.1, along with the NRC

endorsement with exception. Since NRC’s intent did not change, but the standard did change,

the applicant also committed to the following, concerning validation using ANSI/ANS-8.1-1998

(ANSI/ANS, 1998a): “In addition, the details of validation should state computer codes used,

operations, recipes for choosing code options (where applicable), cross-section sets, and any

numerical parameters necessary to describe the input.”

The applicant also used a newer version of ANSI/ANS-8.7 (ANSI/ANS, 1998b) than the version

of the standard endorsed by NRC in Regulatory Guide 3.71 (NRC, 1998). NRC staff reviewed

the differences between the two versions of ANSI/ANS-8.7 and determined that it was

acceptable for the applicant to use the newer version without exception.

Based on the review of the information provided, the staff finds the applicant has identified

appropriate ANSI/ANS Series-8 standards and NRC Regulatory Guides relating to NCS.

5.3.3 Management of the NCS Program
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In Chapters 2, 3, 5, 8, and 11 of the SAR (LES, 2005a) as well as in the facility Emergency Plan

(LES, 2005b), the applicant described how it meets the acceptance criteria for management of

the NCS program.

Specific information about the management of the NCS program is in Sections 5.3.4 (NCS

Organization and Administration) and 5.3.6 (NCS Program) of this SER.

Regarding the specific acceptance criteria in Section 5.4.3.1 of NUREG-1520 (NRC, 2002) for

management of the NCS program, in Section 5.1 of the SAR (LES, 2005a), the applicant:

• Committed to develop, implement, and maintain an NCS program to meet the regulatory

requirements of 10 CFR Part 70;

• Stated that the NCS program objectives are to:

< Prevent an inadvertent criticality;

< Protect against the occurrence of an identified accident sequence; in the

ISA Summary that could lead to an inadvertent criticality;

< Comply with the NCS performance requirements of 10 CFR 70.61;

< Establish and maintain NCS safety parameters and procedures;

< Establish and maintain NCS limits;

< Conduct NCS Analyses and Evaluations to assure that under normal and

credible abnormal conditions, all nuclear processes are subcritical and

maintain an approved margin of subcriticality for safety;

< Establish and maintain NCS IROFS, based on current NCS Analyses and

NCS Evaluations;

< Provide training in emergency procedures in response to an inadvertent

criticality;

< Comply with NCS baseline design criteria in 10 CFR 70.64(a)(9);

< Comply with the NCS ISA Summary requirements in 10 CFR 70.65(b);

• Established NCS safety parameters and will establish NCS procedures;

• Outlined an NCS program structure and defined the responsibilities and authorities of

key program personnel;

• Committed to keep NCS methodologies and NCS technical practices applicable to

current configuration by means of the configuration management function;

• Committed to the preparation of NCS postings, NCS training, and NCS emergency

procedure training;

• Committed to adhere to the NCS baseline design criteria requirements in 

10 CFR 70.64(a)(9); and

• Committed to use the NCS program to evaluate modifications to operations, to

recommend process parameter changes necessary to maintain the safe operation of the

facility, and to select appropriate NCS IROFS and management measures to ensure the

availability and reliability of the IROFS.
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Based on its review of the information provided, the staff finds the applicant has committed to

adequate management measures for the NCS program.

5.3.4 NCS Organization and Administration

In Chapters 2, 5, and 8 of the SAR (LES, 2005a) as well as in the facility Emergency Plan (LES,

2005b), the applicant described how it meets the NCS organization and administration

acceptance criteria, including the technical qualifications, training, and experience of the

applicant and members of its staff to engage in the proposed NCS activities.

5.3.4.1 NCS Organization

Specific information about the commitments to ANSI/ANS Series-8 NCS standards is in Section

5.3.2 of this SER. In Chapter 2 of the SAR (LES, 2005a), the applicant identified that during the

design phase of the facility, the NCS function is performed within the design engineering

organization, which is responsible for implementing the NCS program. During the operations

phase of the facility, the NCS function will report to the Health, Safety, and Environment

Manager, who is responsible for implementing the NCS program. Also, the applicant identified

that the Chief Operating Officer; Plant Manager; Safety Review Committee; Health, Safety, and

Environment Manager; Criticality Safety Engineer; Shift Crew; and NCS staff have and will have

responsibilities related to NCS.

Regarding the specific acceptance criteria in Section 5.4.3.2 of NUREG-1520 (NRC, 2002) for

NCS organization, the applicant:

• Identified and functionally described the organizational groups responsible for managing

the design, construction, and operation of the facility for NCS;

• Committed to describe organizational positions and outline organizational relations

among the individual positions for NCS; and

• Committed to ANSI/ANS-8.1, "Nuclear Criticality Safety in Operations with Fissionable

Materials Outside Reactors" (ANSI/ANS, 1988a), ANSI/ANS-8.1, "Nuclear Criticality

Safety in Operations with Fissionable Materials Outside Reactors" (ANSI/ANS, 1998a),

and ANSI/ANS-8.19, “Administrative Practices for Nuclear Criticality” (ANSI/ANS,

1996b), as they relate to the NCS organization.

5.3.4.2 NCS Administration

In Chapters 2, 5, and 8 of the SAR (LES, 2005a) and in the facility Emergency Plan (LES,

2005b), the applicant identified that responsible managers have (for design) and will have (for

operations) the authority to delegate tasks to other individuals; however, the responsible

manager retains (for design) and will retain (for operations) the ultimate responsibility and

accountability for implementing the applicable requirements. Additional information about NCS

administration is in Sections 5.3.2 and 5.3.6 of this SER.

In Chapters 2, 5, and 8 of the SAR (LES, 2005a) and the facility Emergency Plan (LES, 2005b),

the applicant identified that the Chief Operating Officer; Plant Manager; Safety Review
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Committee; Health, Safety, and Environment Manager; Criticality Safety Engineer; Shift Crew;

and NCS staff have and will have the following responsibilities related to NCS:

The Chief Operating Officer will:

1. Provide Criticality Safety Engineers in sufficient numbers to implement and support the

operation of the NCS program; and

2. Have the appropriate management and supervisory responsibilities identified in

ANSI/ANS-8.19 (ANSI/ANS, 1996b).

The Plant Manager will:

1. Provide Criticality Safety Engineers in sufficient numbers to implement and support the

operation of the NCS program; and

2. Have the appropriate management and supervisory responsibilities identified in 

ANSI/ANS-8.19 (ANSI/ANS, 1996b).

The Safety Review Committee will:

1. Provide technical and administrative review and audit of operations that could impact

facility worker, public safety, and environmental impacts, including NCS;

2. Include experts on operations and all safety disciplines, including NCS; and

3. Conduct at least one facility review and audit per year in areas, including NCS control.

The Health, Safety, and Environment Manager will:

1. Assure safety at the facility through activities, including NCS;

2. Work with the other facility managers to ensure consistent interpretations of Health,

Safety, and Environment requirements, including NCS;

3. Perform independent reviews, including NCS;

4. Support facility and operations change control reviews, including NCS;

5. Ensure objective Health, Safety, and Environment audit, review, and control activities by

being independent from other management positions;

6. Have the authority to shut down operations, if they appear to be unsafe, including NCS;

7. Need to consult with the Plant Manager with respect to restart of shutdown operations,

after the deficiency or unsatisfactory condition has been resolved, including NCS;

8. Have the authority and responsibility to assign and direct activities for the NCS Staff;
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9. Provide Criticality Safety Engineers in sufficient numbers to implement and support the

operation of the NCS program; and

10. Have the appropriate management and supervisory responsibilities identified in 

ANSI/ANS-8.19 (ANSI/ANS, 1996b).

The Criticality Safety Engineer will:

1. Report to the Health, Safety, & Environment Manager (via a designated supervisory

position, if applicable);

2. Prepare NCS Evaluations and NCS Analyses;

3. Review NCS Evaluations and NCS Analyses;

4. Conduct and report on periodic NCS Assessments;

5. Implement the NCS program; and

6. Have the appropriate NCS Staff responsibilities identified in ANSI/ANS-8.19

(ANSI/ANS,1996b).

The Shift Crew will have at least one Criticality Safety Engineer available to be contacted by the

Shift Manager to respond to any routine request or emergency condition. This availability may

be offsite, if adequate communication ability is provided to allow response as needed.

During the operations phase of the facility, the NCS Staff responsibilities includes:

1. Establishing the NCS program, including design criteria, procedures, and training;

2. Providing NCS support for the integrated safety analysis and configuration control;

3. Assessing normal and credible abnormal conditions for NCS;

4. Determining NCS limits for controlled parameters;

5. Developing and validating methods to support NCS Evaluations (i.e., non-calculational

engineering judgments regarding whether existing NCS Analyses bound the issue being

evaluated or whether new or revised NCS Analyses are required);

6. Performing NCS Analyses (i.e., calculations), writing NCS Evaluations, reviewing and

approving proposed changes in process conditions on equipment involving fissionable

material;

7. Specifying NCS control requirements and functionality;

8. Providing advice and counsel on NCS control measures, including reviewing and

approving operating procedures;
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9. Supporting planning for emergency response and responding to events;

10. Evaluating the effectiveness of the NCS program using audits and assessments; and

11. Providing NCS postings that identify administrative controls for operators in applicable

work areas.

In Chapters 2, 5, and 8 of the SAR (LES, 2005a) and the facility Emergency Plan (LES, 2005b),

the applicant identified qualifications related to NCS for the Chief Operating Officer; Plant

Manager; Safety Review Committee; Health, Safety, and Environment Manager; Criticality

Safety Engineer; Shift Crew; and NCS staff.

There are no minimum requirements for the Chief Operating Officer, related to NCS.

The minimum requirements for the Plant Manager, related to NCS are:

1. Knowledgeable of the enrichment process, enrichment process controls and ancillary

processes, and other safety aspects that apply to the overall safety of the facility,

including NCS control;

2. Bachelor’s degree (or equivalent) in an engineering or scientific field; and

3. Ten years of responsible nuclear experience.

The minimum requirements for an NCS member of the Safety Review Committee related to

NCS are:

1. Academic degree in an engineering or physical science field;

2. Five years technical experience; and

3. Three years related directly to NCS, which is included in the 5 years of technical

experience.

The minimum requirements for a Health, Safety, and Environment Manager related to NCS are:

1. Bachelor's degree (or equivalent) in an engineering or scientific field;

2. Five years of responsible nuclear experience in Health, Safety, and Environment or

related disciplines; and

3. One year of direct experience in the administration of NCS Evaluations and NCS

Analyses.

The minimum requirements for a Criticality Safety Engineer related to NCS are:

1. Bachelor’s degree (or equivalent) in an engineering or scientific field;
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2. Two years experience in the implementation of an NCS program;

3. Two years nuclear industry experience in NCS;

4. Understanding and experience in the application and direction of NCS programs; and

5. Successful completion of a training program, applicable to the scope of operations, in

the physics of criticality and in associated safety practices.

In addition, should a change to the facility require an NCS Evaluation or an NCS Analysis, then

the person who would do that NCS Evaluation or NCS Analysis must meet the following

minimum requirements:

1. Possess the equivalent qualifications of a Criticality Safety Engineer; and

2. Have 2 years of experience performing NCS Analyses and implementing NCS

programs.

The minimum requirement for the Shift Crew is for a Criticality Safety Engineer to be able to be

contacted by the Shift Manager to respond to any routine request or emergency condition.

The minimum requirement for the NCS staff is the same as for a Criticality Safety Engineer.

The applicant will implement the intent of the administrative practices for NCS, as contained in

Section 4.1.1 of ANSI/ANS-8.1 (ANSI/ANS, 1998a). Also, a policy will be established whereby,

“Personnel must report defective NCS conditions and perform actions only in accordance with

written, approved procedures. Unless a specific procedure deals with the situation, personnel

must report defective NCS conditions and take no action until the situation has been evaluated

and recovery procedures provided.”

Regarding the specific acceptance criteria in Section 5.4.3.2 of NUREG-1520 (NRC, 2002) for

NCS administration, the applicant:

• Identified functional responsibilities, experience, and qualification of personnel

responsible for NCS;

• Committed to ANSI/ANS-8.1, "Nuclear Criticality Safety in Operations with Fissionable

Materials Outside Reactors" (ANSI/ANS, 1988a), ANSI/ANS-8.1, "Nuclear Criticality

Safety in Operations with Fissionable Materials Outside Reactors" (ANSI/ANS, 1998a),

and ANSI/ANS-8.19, “Administrative Practices for Nuclear Criticality” (ANSI/ANS,

1996b), as they relate to NCS administration;

• Committed to implement the intent of the administrative practices for NCS, as contained

in Section 4.1.1 of ANSI/ANS-8.1 (ANSI/ANS, 1998a), which is to use personnel, skilled

in the interpretation of data pertinent to NCS and familiar with the operation of the

facility, as a resource in NCS management decisions and the specialists should be

independent of operations supervision;
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• Committed to provide NCS postings in areas, operations, work stations, and storage

locations;

• Committed to the establish the following policy: “Personnel must report defective NCS

conditions and perform actions only in accordance with written, approved procedures.

Unless a specific procedure deals with the situation, personnel must report defective

NCS conditions and take no action until the situation has been evaluated and recovery

procedures provided;” and

• Committed to staff the NCS program with suitably trained personnel and to provide

sufficient resources for its operation.

5.3.4.3 NCS Organization and Administration Conclusion

Based on its review of the information provided, the staff concludes that the applicant has

described an adequate NCS organization and associated administration to meet the

requirements of 10 CFR 70.22(a)(6).

5.3.5 NCS Management Measures

In Chapters 3 and 11 of the SAR (LES, 2005a), the applicant described how it meets the NCS

management measures acceptance criteria.

Additional information about training and procedures is in Sections 5.3.2 and 5.3.4 of this SER. 

Specific information about training for the NCS criticality accident alarm system is in Section

5.3.6 of this SER.

In Section 3.0.3 of the SAR (LES, 2005a), the applicant identified that management measures

are functions that will be applied to IROFS and will be applied to any items that may affect the

function of IROFS. Management measures will be applied to particular structures, systems,

equipment, components, and activities of personnel and may be graded commensurate with the

reduction of the risk. Management measures will ensure that those structures, systems,

equipment, components, and activities of personnel within the identified IROFS boundary are

designed, implemented, and maintained, as necessary, to ensure that the IROFS are available

and reliable to perform their intended function, when needed to comply with the performance

requirements of 10 CFR 70.61.

In Section 11.1 of the SAR (LES, 2005a), the applicant identified that, for any change [i.e., new

design or operation, or modification to the facility, or modification to activities of personnel 

(e.g., site structures, systems components, computer programs, processes, operating

procedures, management measures)], that involve or could affect uranium on site, an NCS

Evaluation and, if required, an NCS Analysis, will be prepared and approved. Before

implementing the change, it will be determined that the entire process will be subcritical with an

applicable margin for safety under both normal and credible abnormal conditions. Each

modification will also be evaluated and documented for radiation exposure to minimize worker

exposures. IROFS, any items that affect the function of an IROFS, and, in general, items

required to satisfy regulatory requirements, will be designated as Quality Level 1. NCS

Analyses and NCS Evaluations will be captured in the document control and records

management procedures. During design, configuration control for NCS is accomplished,
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through the use of procedures for controlling design, including preparation, review (including

interdisciplinary review and preparation of NCS Analyses and NCS Evaluations, as applicable),

and design verification (where appropriate), approval, release, and distribution for use.

In Section 11.2 of the SAR (LES, 2005a), the applicant identified that maintenance procedures

involving IROFS will have: (a) corrective and preventive maintenance; (b) functional testing

after maintenance; and (c) surveillance/monitoring maintenance activities. For new procedures,

or work activities that involve or could affect uranium on site, an NCS Evaluation (and, if

required, an NCS Analysis) will be prepared and approved.

In Section 11.3 of the SAR (LES, 2005a), the applicant identified that all persons under the

supervision of facility management, including contractors, must participate in the General

Employee Training and that training encompasses nuclear safety training. Nuclear safety

training programs will be established for the various types of job functions commensurate with

the NCS responsibilities associated with each position. That training will be highlighted to

stress the high level of importance placed on the NCS of protecting the worker. Those training

sessions covering NCS will be conducted on a regular basis and will include principles of NCS. 

NCS training for personnel associated with fissile material operations outside reactors, where

potential exists for an inadvertent criticality, will be in accordance with ANSI/ANS-8.20

(ANSI/ANS, 1991).

In Section 11.4 of the SAR (LES, 2005a), the applicant identified that all activities involving

licensed materials or IROFS will be conducted in accordance with approved procedures. 

Procedures will be used to control activities to ensure that the activities are carried out in a safe

manner and in accordance with regulatory requirements. The facility training program will train

the required personnel in the use of the latest operating procedures. Administrative procedures

will be used to perform activities that support processing operations, including NCS. 

Procedures will be established and implemented for NCS in accordance with ANSI/ANS-8.19

(ANSI/ANS, 1996b). The NCS procedures will be written such that no single, inadvertent

departure from a procedure could cause an inadvertent criticality. NCS postings at the facility

will be established that identify administrative controls applicable and appropriate to the activity

or area in question. NCS procedures and postings will be controlled by procedure to ensure

that they are maintained current. For proposed procedure changes having a potential impact on

NCS, an NCS Evaluation, and, if required, an NCS Analysis, will be performed. Any necessary

controlled safety parameters, limits, IROFS, management measures, or NCS Analyses that

must be imposed or revised will be adequately reflected in appropriate procedures and/or

design basis documents. A second Criticality Safety Engineer will independently review

changes and then the Health, Safety, and Environment Manager will approve them (if

appropriate) before they go into effect.

In Section 11.5 of the SAR (LES, 2005a), the applicant identified that there will be a two-tiered

approach to verifying compliance to procedures and performance to regulatory requirements. 

Audits will be focused on verifying compliance with regulatory requirements, procedural

requirements, and licensing commitments. Assessments will be focused on 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. Audits and assessments will be

conducted in the NCS functional area. NCS audits will be conducted and documented

quarterly, such that all aspects of the NCS program will be audited at least every 2 years. The

Operations Group will be assessed periodically to ensure that NCS procedures are being
5-12

followed and that the process conditions have not been altered to adversely affect NCS. NCS

assessments will be performed in accordance with ANSI/ANS-8.19 (ANSI/ANS, 1996b) and will

ensure that operations conform to NCS requirements. The frequency of NCS assessments will

be based on the controls identified in the NCS Analyses and NCS Evaluations and will be

conducted at least semiannually. Weekly NCS walk-throughs of UF6 process areas will be

conducted and documented.

In Section 11.6 of the SAR (LES, 2005a), the applicant identified that the incident investigation

process will be a mechanism available for use by anyone at the facility for reporting

deficiencies, abnormal events, and potentially unsafe conditions or activities. NCS

requirements will be addressed when performing an incident investigation.

In Section 11.7 of the SAR (LES, 2005a), the applicant identified that records management will

be performed in a controlled and systematic manner to provide identifiable and retrievable

documentation, including for NCS.

In Section 11.8 of the SAR (LES, 2005a), the applicant identified that the Quality Assurance

Program and its supporting manuals, procedures, and instructions will be applicable to items

and activities designated as Quality Level 1 and Quality Level 2. The Quality Level 1 program

will be applied to those structures, systems, components, and administrative controls that have

been determined to be IROFS, items that affect the function of the IROFS, and, in general,

items required to satisfy regulatory requirements.

Regarding the specific acceptance criteria in Section 5.4.3.3 of NUREG-1520 (NRC, 2002) for

NCS management measures, the applicant:

• Committed to ANSI/ANS-8.19 (ANSI/ANS, 1996b) and ANSI/ANS-8.20, “Nuclear

Criticality Safety Training” (ANSI/ANS, 1991), as they relate to training;

• Committed to provide training to all personnel to recognize the criticality accident alarm

system signal and to evacuate promptly to a safe area;

• Committed to provide training regarding the policy that: “Personnel must report

defective NCS conditions and perform actions only in accordance with written, approved

procedures. Unless a specific procedure deals with the situation, personnel must report

defective NCS conditions and take no action until the situation has been evaluated and

recovery procedures provided.”;

• Committed to ANSI/ANS-8.19 (ANSI/ANS, 1996b), as it relates to procedures and to the

policy that no inadvertent departure from a procedure could cause an inadvertent

criticality;

• Committed to ANSI/ANS-8.19 (ANSI/ANS, 1996b), as it relates to audits and

assessments;

• Committed to conduct and document weekly NCS walk-throughs of UF6 process areas;

and
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• Committed to NCS audits to be conducted and documented quarterly, such that all

aspects of the NCS program, including management measures, will be audited at least

every 2 years.

Based on its review of the information provided, the staff concludes that the applicant has

described adequate NCS management measures to meet the requirements of 10 CFR 70.62(d).

5.3.6 NCS Program

The NCS program includes the commitments and descriptions of how to meet those

commitments to prevent an inadvertent criticality and to respond to an inadvertent criticality.

In Chapters 3 and 5 of the SAR (LES, 2005a) and the ISA Summary (LES, 2005c), the applicant

described the NCS program, which included descriptions in the following areas: 

(1) NCS methodologies and technical practices, (2) NCS criticality accident alarm system, 

(3) NCS subcriticality of operations and margin of subcriticality for safety, (4) NCS baseline

design criteria, (5) NCS in the integrated safety analysis (ISA) program and the ISA Summary,

and (6) additional NCS program commitments.

5.3.6.1 NCS Methodologies and Technical Practices

In Section 5.1 of the SAR (LES, 2005a), the applicant identified that the facility will be designed,

constructed, and operated, such that an inadvertent criticality will be prevented. NCS at the

facility will be assured by designing the facility, systems, and components with safety margins,

such that safe conditions are maintained under normal and credible abnormal conditions.

In Section 5.1.1 of the SAR (LES, 2005a), the applicant committed to the double contingency

principle, as stated in ANSI/ANS-8.1 (ANSI/ANS, 1998a), which is that “Process designs should

incorporate sufficient factors of safety to require at least two unlikely, independent, and

concurrent changes in process conditions before a criticality accident is possible.” Each

process that has accident sequences that could result in an inadvertent criticality at the facility

will meet the double contingency principle.

In Section 5.1.3 of the SAR (LES, 2005a), the applicant identified that process operations will

require the establishment of NCS limits. The facility UF6 systems will involve mostly gaseous

operations. These operations will be carried out under reduced atmospheric conditions 

(i.e., vacuum) or at slightly elevated pressures, not exceeding three atmospheres. Within the

Separations Building, significant accumulations of enriched UF6 could reside only in the Product

Low-Temperature Take-off Stations, Product Liquid Sampling Autoclaves, Product Blending

System, and the UF6 Cold Traps. Except for the UF6 Cold Traps, all of these will contain the

UF6 in Type 30B or Type 48Y cylinders. All these significant accumulations will be within

enclosures protecting them from water ingress. The facility design will minimize the possibility

of accidental moderation (e.g., water or hydrogen) by eliminating direct water contact with these

cylinders of accumulated UF6. In addition, the facility’s procedural controls for enriching the UF6

will assure that it does not become unacceptably hydrogen-moderated while in process. The

facility’s UF6 systems’ operating procedures will contain measures against the loss of

moderation control, according to ANSI/ANS-8.22 (ANSI/ANS, 1997c).
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In Section 5.1.4 of the SAR (LES, 2005a), the applicant identified that each portion of the

facility, system, or component that may possibly contain enriched uranium will be designed with

NCS as an objective. Where there will be significant in-process accumulations of enriched

uranium as UF6, the facility design will include multiple features to minimize the possibilities for

breakdown of the moderation control limits. These features will eliminate direct ingress of water

to product cylinders while in-process.

Table 5.3-1 provides the safety criteria (i.e., parameter, critical value, safe value, and safety

factor) for uniform aqueous solutions of enriched UO2F2 for the facility, assuming a single

component and 6.0 wt percent U-235 enrichment. In the table, the term safety factor means the

ratio between the safe value and the critical value and the term keff means the effective neutron

multiplication factor. keff represents the neutron production rate relative to the neutron loss rate. 

Therefore, the critical value - when keff is one - represents the point at which an inadvertent

criticality (i.e., nuclear chain reaction) occurs. When keff is less than one, the system is

subcritical. NRC reviewed the ‘Critical Value’ in the Table 5.3-1 for 6.0 wt percent against the

values in ANSI/ANS-8.1 (ANSI/ANS, 1998a) at 5.0 and 10.0 wt percent. By using knowledge,

experience, and best estimate for interpolation, NRC determined that the applicant’s values in

Table 5.3-1 are consistent with the values in ANSI/ANS-8.1 (ANSI/ANS, 1998a).

Table 5.3-1

Safety Criteria

Parameter Critical Value

(keff = 1.0)

Safe Value

(keff = 0.95)

Safety

Factor

Volume 24 L (6.3 gal.) 18 L (4.8 gal.) 0.75

Cylinder Diameter 24.4 cm (9.6 in.) 21.9 cm (8.6 in.) 0.90

Slab Thickness 11.5 cm (4.5 in.) 9.9 cm (3.9 in.) 0.86

Water Mass 15.4 kg (34.0 lb) 11.5 kg (25.4 lb) 0.75

Areal Density 9.5 g/cm2

 (19.5

lb/ft2

)

7.5 g/cm2

 (15.4

lb/ft2

)

0.79

Uranium Mass with no double batching 27 kg (59.5 lb) 19.5 kg (43.0 lb) 0.72

Uranium Mass with double batching 27 kg (59.5 lb) 12.2 kg (26.9 lb) 0.45

Table 5.3-2 shows how the above safety criteria for single components will be specifically

applied at the facility (i.e., building/system/component, control mechanism, safety criteria) to

prevent an inadvertent criticality, assuming 6.0 wt percent U-235 enrichment, which is

conservative, because the facility will be limited to 5.0 wt percent U-235 enrichment, except for

the Contingency Dump Trap System that will be limited (based on physics) to 1.5 wt percent U235.

Table 5.3-2

Application of Safety Criteria
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Building, System, or Component Control

Mechanism

Safety Criteria

Enrichment Enrichment 6.0 wt percent U-235

Centrifuges Diameter < 21.9 cm (8.6 in.)

Product Cylinders (30B) Moderation Hydrogen < 0.95 kg (2.09 lb)

Product Cylinders (48Y) Moderation Hydrogen < 1.05 kg (2.31 lb)

UF6 Piping Diameter < 21.9 cm (8.6 in.)

Chemical Traps Diameter < 21.9 cm (8.6 in.)

Product Cold Trap Diameter < 21.9 cm (8.6 in.)

Contingency Dump System Traps Enrichment < 1.5 wt percent U-235

Tanks Mass < 12.2 kg U (26.9 lb U)

Feed Cylinders Enrichment < 0.72 wt percent U-235

Uranium Byproduct Cylinders Enrichment < 0.72 wt percent U-235

UF6 Pumps (first stage) Not Applicable Safe, by explicit calculation

UF6 Pumps (second stage) Volume < 18.0 L (4.8 gal.)

Individual Uranic Liquid Containers (e.g.,

Fomblin Oil Bottle, Laboratory Flask, Mop

Bucket)

Volume < 18.0 L (4.8 gal.)

Vacuum Cleaners Oil Containers Volume < 18.0 L (4.8 gal.)

In Section 5.2.1.2 of the SAR (LES, 2005a), the applicant described the process that was used

to identify the margin of subcriticality for safety when doing calculations to demonstrate safety

(i.e., not scoping calculations) for normal and credible abnormal conditions. The keff equation

that the applicant used for those calculations is: keff = kcalc + 3Fcalc < 0.95. The applicant used

the MONK8 Monte Carlo computer program (AEAT, 1998) to calculate neutron multiplication to

ascertain the keff values.

In Section 5.1.2 of the SAR (LES, 2005a), the applicant identified that the major controlling

safety parameters to be used in the facility will be enrichment control, geometry control,

moderation control, or limitations on the mass as a function of enrichment. In addition,

reflection, interaction, and heterogeneous effects are important parameters that will be

considered and applied. NCS Evaluations and NCS Analyses will be used to identify the

significant safety parameters affected within a particular system. All assumptions relating to

process, equipment, material function, and operation, including credible abnormal conditions,

will be justified, documented, and independently reviewed. Where possible, passive engineered

controls will be used to ensure NCS.
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In Section 5.1.1 of the SAR (LES, 2005a), the applicant committed to the following elements

related to NCS methodologies and technical practices in the NCS program:

• Adherence to the double contingency principle, as stated in ANSI/ANS-8.1 (ANSI/ANS,

1998a);

• Safety parameters will be established and procedures will be established;

• The NCS program structure will be provided, including definition of the responsibilities

and authorities of key program personnel;

• The NCS methods and technical practices will be kept applicable to current configuration

by means of the configuration management function and the NCS program will be

upgraded, as necessary, to reflect changes in the ISA or NCS methodologies and to

modify operating and maintenance procedures in ways that could reduce the likelihood

of occurrence of an inadvertent criticality;

• The NCS program will be used to establish and maintain NCS safety limits and NCS

operating limits for IROFS in nuclear processes and there is a commitment to maintain

adequate management measures, to ensure the availability and reliability of the IROFS;

• NCS postings will be provided and maintained current;

• NCS emergency procedure training will be provided;

• The NCS baseline design criteria requirements in 10 CFR 70.64(a)(9) will be adhered to

(i.e., see Table 5.3-2, “Application of Safety Criteria”);

• The NCS program will be used to evaluate modifications to operations, to recommend

process parameter changes necessary to maintain the safe operation of the facility, and

to select appropriate IROFS and management measures;

• The NCS program will be used to promptly detect NCS deficiencies by means of

operational inspections, audits, and investigations and deficiencies will be entered into

the corrective action program so as to prevent recurrence of unacceptable performance

deficiencies in IROFS, NCS functions, or management measures;

• NCS program records will be retained as described in Section 11.7 of the SAR (LES,

2005a); and

• NCS training will be provided to individuals who handle nuclear material at the facility

based on the training program described in ANSI/ANS-8.20 (ANSI/ANS, 1991) and the

training program will be developed and implemented with input from the NCS staff,

training staff, and management. The training will focus on the following:

< Appreciation of the physics of NCS;

< Analysis of jobs and tasks to determine what a worker must know to perform

NCS tasks efficiently; and
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< Design and development of learning objectives, based on the analysis of jobs

and tasks, that reflect the knowledge, skills, and abilities needed by the worker.

In Section 5.1.1 of the SAR (LES, 2005a), the applicant committed to the following general NCS

philosophy: (1) to prevent accidental uranium enrichment excesses; (2) to provide geometrical

safety, when practical; (3) to provide for moderation control within the UF6 processes; and (4) to

impose mass limits on containers of aqueous, solvent-based, or acid solutions containing

uranium. Interaction control will provide for safe movement and storage of components. In

addition, facility and equipment features will assure prevention of excessive enrichment.

In Section 5.1.1 of the SAR (LES, 2005a), the applicant described that the facility will be divided

into six distinctly separate Assay Units (i.e., Cascade Halls) with no common UF6 piping. UF6

blending will be done in a physically separate portion of the facility. Process piping, individual

centrifuges, and chemical traps, other than the Contingency Dump Chemical Traps, will be safe

by limits placed on their diameters. Product cylinders will rely on uranium enrichment,

moderation control, and mass limits, to protect against the possibility of an inadvertent criticality. 

Liquid effluent collection tanks that hold uranium in solution will be mass-controlled because

none will be geometrically safe.

In Section 5.1.2 of the SAR (LES, 2005a), the applicant identified that the major controlling

safety parameters to be used in the facility will be enrichment control, geometry control,

moderation control, or limitations on the mass as a function of enrichment. In addition,

reflection, interaction, and heterogeneous effects will be important parameters that will be

considered and applied where appropriate in NCS Analyses. NCS Evaluations and NCS

Analyses will be used to identify the significant safety parameters affected within a particular

system. All assumptions relating to process, equipment, material function, and operation,

including credible abnormal conditions, will be justified, documented, and independently

reviewed. Where possible, passive engineered controls will be used to ensure NCS.

Other than the Type 30B Cylinders, Type 48Y Cylinders, first-stage UF6 Pumps, and

Contingency Dump Chemical Traps, all Separation Plant components that handle enriched UF6

will be safe by favorable geometry, to preclude an inadvertent criticality. A centrifuge array will

use a probability argument with multiple operational procedure barriers to demonstrate NCS. A

product cylinder will be safe by total moderator or hydrogen-to-uranium (H/U) ratio control, as

appropriate, to preclude an inadvertent criticality. In the Technical Services Building, NCS for

uranium loaded liquids will be ensured by limiting the mass of uranium in any single tank to less

than or equal to 12.2 kg U (26.9 lb U) to preclude an inadvertent criticality. Individual liquid

storage bottles will be safe by volume to preclude an inadvertent criticality. Interaction in

storage arrays will be accounted for to preclude an inadvertent criticality.

The control safety parameters to be applied to the facility are as follows:

Enrichment
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Enrichment will be controlled to limit the wt percent U-235 within any process, vessel, or

container. The Contingency Dump System will have a maximum enrichment of 1.5 wt percent

(based on physics), but all other locations in the facility will have a maximum enrichment of 

5.0 wt percent. As an added non-quantified conservatism, even though those parts of the

facility will be limited to a maximum enrichment of 5.0 wt percent, NCS will be analyzed at 

6.0 wt percent.

Geometry/Volume

Geometry/volume control may be used to ensure NCS within specific process operations or

vessels and within storage containers. When performing calculations to determine safety, the

geometry/volume limits will be chosen to ensure that: keff = kcalc + 3Fcalc < 0.95. The favorable

geometry/favorable volume safe values for control define the characteristic dimension of

importance for a single unit such that NCS will not dependent on any other parameter,

assuming an enrichment of 6.0 wt percent.

Moderation

Water and oil will be the moderators considered at the facility. The only system where

moderation will be used as a control parameter is in the Product Cylinders. Moderation control

will be established consistent with ANSI/ANS-8.22 (ANSI/ANS, 1997c) and will incorporate the

criteria below:

• Controls will be established to limit the amount of moderation entering the cylinders;

• When moderation is not considered a control parameter, either optimum moderation or

worst-case H/U ratio will be assumed when performing the NCS Analysis; and

• When moderation is the only parameter used for NCS control, the following additional

criteria will be applied that will assure that at least two independent controls would have

to fail before an inadvertent criticality is possible:

< Two independent controls will be established and used to monitor and limit

uncontrolled moderator before returning a cylinder to production, thereby limiting

the amount of uncontrolled moderator from entering a system to an acceptable

limit; and

< The NCS Evaluation of the cylinders under moderation control will include the

establishment of limits for the ratio of maximum moderator-to-fissile material for

both normal and credible abnormal conditions and the NCS Analysis will be

supported by parametric studies.

Mass

Mass control may be used to limit the quantity of uranium within specific process operations,

vessels, or storage containers. Mass control may be used on its own or in combination with

other control methods. Analysis or sampling (using instrumentation) will be employed to verify

the mass of the material. Conservative administrative limits for each operation will be specified

in operating procedures.
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Whenever mass control is established for a container, records will be maintained for mass

transfers into and out of the container. Establishment of mass limits for a container will involve

consideration of potential moderation, reflection, geometry, spacing, and enrichment. The NCS 

Evaluation will consider normal and credible abnormal conditions for determination of the

operating mass limit for the container and the definition of subsequent controls necessary to

prevent reaching the safety limits. When only administrative controls are used for masscontrolled systems, double batching will be assumed in the NCS Analysis, which sets a more

conservative limit.

Reflection

Reflection will be considered when performing NCS Evaluations and NCS Analyses. The

possibility of full-water reflection was considered, but the layout of the facility will be a very open

design and so it is not credible that those vessels and facility components requiring NCS control

could become flooded from a source of water within the facility. In addition, neither automatic

sprinkler nor standpipe and hose systems will be provided in the Technical Support Building,

Separation Buildings, Blending and Liquid Sampling Areas, Cylinder Receipt and Dispatch

Building, Centrifuge Assembly Building, and Centrifuge Post Mortem Areas. Therefore, fullwater reflection of vessels will be discounted. For conservatism, some NCS Analyses will be

performed using full-water reflection. Partial reflection of 2.5 cm (0.984 in.) of water will be

assumed where limited moderating materials may be present (e.g., human beings). Since

concrete can be a more efficient reflector than water; it will be modeled in NCS Analyses where

it is present. When moderation control is identified in the ISA Summary, it will be established

within the guidelines of ANSI/ANS-8.22 (ANSI/ANS, 1997c).

Interaction

NCS Evaluations and NCS Analyses will consider the potential effects of interaction. A noninteracting unit is defined as a unit that is spaced an approved distance from other units such

that the multiplication of the subject unit is not increased. Units may be considered noninteracting when they are separated by more than 60 cm (23.6 in.). If units are considered

interacting, then an NCS Analysis will be performed to determine individual unit multiplication

and array interaction using the Monte Carlo computer code MONK8A to ensure that: 

keff = kcalc + 3Fcalc < 0.95.

In Section 5.2.1.4 of the SAR (LES, 2005a), the applicant indicated that NCS will be evaluated

for the design features of the facility system or component and for the operating practices that

relate to maintaining NCS. The NCS Analysis of individual systems or components and their

interaction with other systems or components containing enriched uranium will be performed to

assure that the NCS criteria are met. The NCS Analyses and Table 5.3-1, “Safety Criteria,”

provide a basis for the facility design and for NCS hazards identification performed as part of

the ISA process. Each portion of the facility, system, or component that may possibly contain

enriched uranium will be designed with NCS as an objective.

The facility will be designed and operated in accordance with the parameters in Table 5.3-2,

“Application of Safety Criteria.” The applicant provided a general description of the elements of

an NCS Analysis. During the design phase of the facility, a qualified Criticality Safety Engineer

will perform the NCS Analysis and a second qualified Criticality Safety Engineer will

independently review it. During the operation of the facility, a qualified Criticality Safety
5-20

Engineer will perform the NCS Analysis, a second qualified Criticality Safety Engineer will

independently review it, and the Health, Safety, and Environment Manager will approve it (if

appropriate) before it goes into effect.

In Section 5.2.1.5 of the SAR (LES, 2005a), the applicant indicated that the facility NCS

Analyses will be performed using the methods and assumptions in Section 5.2.1.4 of the SAR

(LES, 2005a) and that NCS Analyses will be performed in order to demonstrate compliance with

the following criteria:

• Methods will be validated and used only within demonstrated acceptable ranges;

• NCS Analyses will adhere to ANSI/ANS-8.1 (ANSI/ANS, 1998a), as it relates to analysis

methods;

• Compliance with the intent of the validation report statement in NRC Regulatory Guide

3.71 (NRC, 1998), including that it applies to ANSI/ANS-8.1 (ANSI/ANS, 1998a), which

states that the applicant should demonstrate: (1) the adequacy of the margin of safety

for subcriticality, by assuring that the margin is large compared to the uncertainty in the

calculated value of keff; (2) the calculation of keff is based on a set of variables whose

values lie in a range for which the method used to determine keff has been validated; and

(3) trends in the bias support the extension of the method to areas outside the area or

areas of applicability;

• A specific reference to (including date and revision number), and summary description

of, either a manual or a documented, reviewed, and approved validation report for each

methodology, will be included. Any change in the reference manual or validation report

will be reported to NRC by letter;

• The reference manual and documented reviewed validation report will be kept at the

facility;

• The reference manual and validation report will be incorporated into the configuration

management program;

• NCS Analyses will be performed in accordance with the methods specified and

incorporated into the configuration management program;

• The NCS methods and technical practices in Section 5.4.3.4 of NUREG-1520 (NRC,

2002) will be used to evaluate NCS accident sequences in operations and processes;

• The acceptance criteria in Section 3.4 of NUREG-1520 (NRC, 2002), as they relate to

the identification of NCS accident sequences; consequences of NCS accident

sequences; likelihoods of NCS accident sequences; and descriptions of IROFS for NCS

accident sequences, will be utilized;

• The applicant will use NCS controls and controlled safety parameters to assure that,

under normal and credible abnormal conditions, all nuclear processes are subcritical –

including use of an approved margin of subcriticality for safety;
5-21

• As stated in ANSI/ANS-8.1 (ANSI/ANS, 1998a), process specifications will incorporate

margins to protect against uncertainties in process variables and against a limit being

accidentally exceeded;

• ANSI/ANS-8.7 (ANSI/ANS, 1998b), as it relates to the requirements for subcriticality of

operations, margin of subcriticality for safety, and selection of controls requirements in

10 CFR 70.61(d), will be used;

• ANSI/ANS-8.10 (ANSI/ANS, 1988b), as it relates to the determination of consequences

of NCS accidents, as modified by NRC Regulatory Guide 3.71 (NRC, 1998), will be

used;

• If administrative keff margins for normal and credible abnormal conditions are used, then

NRC pre-approval of the administrative margins will be sought;

• Subcritical limits for keff calculations such that: keff subcritical = 1.0 - bias - margin, where

the margin includes adequate allowance for uncertainty in the method, data, and bias, to

assure subcriticality, will be used;

• Studies to correlate the change in a value of a controlled parameter and its keff value will

be performed. The studies will include changing the value of one controlled parameter

and determining its effect on another controlled parameter and keff;

• The double contingency principle will be used in determining NCS controls and NCS

IROFS; and

• Acceptance criteria in Section 3.4 of NUREG-1520 (NRC, 2002), as they relate to

subcriticality of operations and margin of subcriticality for safety, will be utilized. That is:

“For nuclear criticality accident sequences, the method evaluates compliance

with 10 CFR 70.61(d). That is, even in a facility with engineered features to limit

the consequences of nuclear criticalities, preventive control(s) must be in place

that are sufficient to ensure that the likelihood of criticality is controlled to be

‘highly unlikely.’ A moderately higher standard of likelihood may be permitted in

preventing such events, consistent with ANSI/ANS Standard 8.10 [(ANSI/ANS,

1988b)]. In particular, criticality cannot result from the failure of any single

IROFS. In addition, potential criticality accidents must meet an [NRC] approved

margin of subcriticality for safety. Acceptance criteria for such margins are

reviewed as programmatic commitments, but the ISA methods must consider

and the ISA Summary must document the actual magnitude of those margins

when they are part of the reason why the postulated accident sequence resulting

in criticality is highly unlikely.”

In Section 5.2.1.6 of the SAR (LES, 2005a), the applicant indicated that the NCS Evaluation

process will be in accordance with ANSI/ANS-8.19 (ANSI/ANS, 1996b). For any change 

[i.e., new design or operation, or modification to the facility, or modification to activities of

personnel (e.g., site structures, systems, components, computer programs, processes,

operating procedures, management measures)] that involves or could affect uranium, an NCS

Evaluation will be prepared and approved. Before implementing the change, it will be
5-22

determined that the entire process will be subcritical, with an approved margin of safety under

both normal and credible abnormal conditions. If this condition cannot be demonstrated with

the NCS Evaluation, then either a new or revised NCS Analysis will be generated that meets the

criteria, or the changes will not be made. The NCS Evaluation will determine and explicitly

identify the controlled safety parameters and associated limits on which NCS depends, assuring

that no single inadvertent departure from a procedure could cause an inadvertent criticality and

that the safety basis of the facility will be maintained during the lifetime of the facility. The NCS

Evaluation will ensure that all potentially affected uranic processes will be evaluated to

determine the effect of the change on the safety basis of the process, including the effect on

bounding process assumptions, the effect on the reliability and availability of NCS controls, and

the effect on the NCS of connected processes. The NCS Evaluation process will involve a

review of the proposed change, discussions with the subject matter experts to determine the

processes that need to be considered, development of the controls necessary to meet the

double contingency principle, and identification of the assumptions and equipment (e.g.,

physical controls and/or management measures) needed to ensure NCS.

The applicant provided a general description of the elements of an NCS Evaluation. The NCS

Evaluation will be performed and documented by a qualified Criticality Safety Engineer. 

Engineering judgment of the qualified Criticality Safety Engineer will be used to ascertain the

NCS impact of the proposed change. The basis for this judgment will be documented with

sufficient detail in the NCS Evaluation to allow the independent review by a second qualified

Criticality Safety Engineer to confirm the conclusions of the judgment of results. After the NCS

Evaluation is completed and documented and the independent review by the second Criticality

Safety Engineer is performed and documented, the Health, Safety, and Environment Manager

will approve it (as appropriate) before it goes into effect.

In Section 5.2.1.7 of the SAR (LES, 2005a), the applicant identified that the facility NCS

Evaluations will be performed according to the following criteria:

• NCS Evaluations will be performed in accordance with the procedures specified and

incorporated in the configuration management program;

• NCS methods and technical practices in Sections 5.4.3.4.1(10)(a), (b), (d), and (e) of

NUREG-1520 (NRC, 2002), will be used to evaluate NCS accident sequences in

operations and processes;

• Acceptance criteria in Section 3.4 of NUREG-1520 (NRC, 2002), as they relate to: 

identification of NCS accident sequences, consequence of NCS accident sequences,

likelihood of NCS accident sequences, and descriptions of IROFS for NCS accident

sequences, will be utilized;

• NCS controls and controlled parameters to assure that, under normal and credible

abnormal conditions, all nuclear processes are subcritical, including use of an approved

margin of subcriticality for safety, will be used;

• The double contingency principle will be used in determining NCS controls and NCS

IROFS; and 
5-23

• Acceptance criteria in Section 3.4 of NUREG-1520 (NRC, 2002), as they relate to

subcriticality of operations and margin of subcriticality for safety, will be utilized. That is:

“For nuclear criticality accident sequences, the method evaluates compliance

with 10 CFR 70.61(d). That is, even in a facility with engineered features to limit

the consequences of nuclear criticalities, preventive control(s) must be in place

that are sufficient to ensure that the likelihood of criticality is controlled to be

‘highly unlikely.’ A moderately higher standard of likelihood may be permitted in

preventing such events, consistent with ANSI/ANS Standard 8.10 [(ANSI/ANS,

1988b)]. In particular, criticality cannot result from the failure of any single

IROFS. In addition, potential criticality accidents must meet an [NRC] approved

margin of subcriticality for safety. Acceptance criteria for such margins are

reviewed as programmatic commitments, but the ISA methods must consider

and the ISA Summary must document the actual magnitude of those margins

when they are part of the reason why the postulated accident sequence resulting

in criticality is highly unlikely.”

Regarding the specific acceptance criteria in Sections 5.4.3.4.1 and 5.4.3.4.2 of NUREG-1520

(NRC, 2002) for NCS methodologies and technical practices, the applicant:

For NCS Methodologies:

• Committed to appropriately apply NCS controlled parameters;

• Committed to appropriately determine NCS limits on IROFS;

• Committed to use acceptable methodologies to perform NCS Analyses and NCS

Evaluations;

• Committed to establish NCS limits on controls and controlled parameters to ensure an

adequate margin of subcriticality for safety;

• Committed to use validated methods to develop NCS limits to ensure that they are within

acceptable ranges, by utilizing both appropriate assumptions and acceptable computer

codes;

• Committed to promptly detect an inadvertent criticality to ensure that radiation exposures

to workers are minimized;

• Committed to ANSI/ANS-8.1 (ANSI/ANS, 1988a) and ANSI/ANS-8.1 (ANSI/ANS, 1998a)

as they relate to NCS methodologies;

• Committed to the intent of the validation report statement in NRC NCS Regulatory Guide

3.71 (NRC, 1998);

• Committed to include a reference to and summary description of the reference manual

or validation report for each methodology that will be used in each NCS Analysis and

NCS Evaluation;
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• Committed to have, at the facility, the reference manual or documented, reviewed, and

approved validation report for each methodology that will be used in each NCS Analysis

and NCS Evaluation;

• Committed to incorporate each reference manual or documented, reviewed, and

approved validation report for each methodology, including the assumptions used, into

the configuration management program; and

• Committed to perform NCS Analyses and NCS Evaluations in accordance with the

following principles:

< NCS safety limits, NCS operating limits, and limits on NCS controlled parameters

will be established assuming credible optimum conditions, unless specified

controls are implemented to control the limit to a certain range of values;

< NCS safety limits, NCS operating limits, and limits on NCS controlled parameters

will be derived from the NCS Analyses and NCS Evaluations;

< NCS safety limits, NCS operating limits, and limits on NCS controlled parameters

will be based on the proper application of the NCS methodology to the process;

< NCS operating limits will be derived from NCS safety limits by taking into

consideration changes in operating parameters to ensure processes will remain

subcritical under both normal and credible abnormal conditions;

< NCS operating limits will establish sufficient margins of safety for processes and

take into consideration the variability and uncertainty in processes and the NCS

subcritical limits;

< NCS safety limits will establish sufficient margins of safety for processes and

take into consideration the variability and uncertainty in processes and the NCS

operating limits;

< The margin of subcriticality for safety for a process should be relative compared

to the calculated value of keff; and

< The keff will be calculated from a set of variables whose values lie in a range for

which the validity of the NCS methodology will have been demonstrated.

For NCS Technical Practices:

• Committed to the policy that no single credible event or failure can result in an

inadvertent criticality;

• Committed to the following general statements about controls: (1) the preferred use of

passive engineered controls to ensure NCS, and (2) the order of preference for NCS

controls will be passive engineered, active engineered, enhanced administrative, and

administrative;

• Committed to consider heterogeneous effects when evaluating a controlled parameter;

• Committed to perform either an NCS Analysis or NCS Evaluation for all controlled

parameters that demonstrates that during both normal and credible abnormal conditions,

the controlled parameter will be maintained;

• Committed to mass as a controlled parameter:
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< When the mass is measured, instrumentation will be used. 

< If double batching is physically possible, then the limit will be no more than 

45 percent of the minimum critical mass; and

< If double batching is not physically possible, then the limit will be no more than 72

percent of the minimum critical mass;

• Committed to geometry as a controlled parameter:

< As part of the Quality Assurance Program and that geometry controls will be

Quality Level 1, before beginning operations, all dimensions and nuclear

properties that use geometry control will be verified and the facility configuration

management program will be used to maintain the dimensions and nuclear

properties;

< The margin of safety for a single unit will be 90 percent of the minimum critical

cylinder diameter;

< The margin of safety for a single unit will be 86 percent of the minimum critical

slab thickness. The NUREG-1520 (NRC, 2002) acceptance criteria is 85

percent, however, the applicant’s commitment to 86 percent is acceptable

because the overall conservatism of limiting the enrichment to 5.0 wt percent, but

calculating limits based on 6.0 wt percent; and

< The margin of safety for a single unit will be 75 percent of the minimum critical

volume;

• Committed to enrichment as a controlled parameter. A method of segregating

enrichments will be used to ensure differing enrichments will not be interchanged (i.e,

Contingency Dump System will be physically limited to 1.5 wt percent and all other parts

of the facility will be limited to 5.0 wt percent).

• Committed to reflection as a controlled parameter;

• Committed to moderation as a controlled parameter to use ANSI/ANS-8.22 (ANSI/ANS,

1997c);

• Committed to interaction as a controlled parameter when maintaining a physical

separation between units, engineered controls to ensure a minimum spacing or

enhanced administrative controls will be used. In addition, the structural integrity of the

spacers or racks should be sufficient for normal and credible abnormal conditions.

Based on its review of the information provided, the staff finds the applicant’s NCS

methodologies and technical practices to be acceptable.

5.3.6.2 NCS Criticality Accident Alarm System

In Section 5.3 of the SAR (LES, 2005a), the applicant indicated that the facility will be provided

with a CAAS, as required by 10 CFR 70.24. Areas where special nuclear material will be

handled, used, or stored in amounts at or above the 10 CFR 70.24 mass limits will be provided
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with CAAS coverage. Emergency management measures are covered in the facility Emergency

Plan (LES, 2005b).

In Chapter 8 of the SAR (LES, 2005a), the applicant committed to the facility Emergency Plan

(LES, 2005b). The facility Emergency Plan (LES, 2005b) included a description of the intended

activities related to NCS, the CAAS, and what the response will be if an inadvertent criticality

occurs. No inadvertent criticality has ever occurred at a facility processing low enriched

uranium (LEU). An inadvertent criticality with low enriched uranium requires a precise

combination of conditions. The facility will include multiple design, administrative, and

engineered controls to prevent an inadvertent criticality. The Technical Services Building, three

Cascade Halls, Cylinder Receipt and Dispatch Building, Blending and Liquid Sampling Area,

and UF6 Handling Area are the only areas in the facility where an inadvertent criticality could

occur. A CAAS will be provided to detect and alarm if an inadvertent criticality occurs in an area

where uranium at or above the 10 CFR 70.24 mass limits will be handled, used, or stored. If an

inadvertent criticality were detected, then: (1) the CAAS would interface with the Plant Control

System to provide a visual and audible alarm of the CAAS and CAAS status in the Control

Room; (2) facility personnel would be alerted to evacuate the specific area, by a specific CAAS

visual and audible alarm; (3) such evacuated facility personnel would assemble immediately in

the Assembly Areas as required by the facility Emergency Plan Implementing Procedures; and

(4) emergency response and protective measures for a Site Area Emergency would be

performed because the applicant classified an inadvertent criticality as a Site Area Emergency. 

Also, criticality dosimeters will be located within the facility buildings and, when recovered and

evaluated, the dosimeters will provide spectrum information to assist in reconstruction of the

inadvertent criticality event.

In Section 3.1.5 of the ISA Summary (LES, 2005c), the applicant indicated that the facility will be

provided with a CAAS, as required by 10 CFR 70.24. Areas where special nuclear material will

be handled, used, or stored in amounts at or above the 10 CFR 70.24 mass limits will be

provided with CAAS coverage. The CAAS is designed, will be installed, and will be maintained

in accordance with ANSI/ANS-8.3 (ANSI/ANS, 1997a), as modified by Regulatory Guide 3.71

(NRC, 1998). CAAS coverage will consist of an overlapping detection layout, where all required

covered areas will be monitored by a minimum of a pair of gamma detectors. Detectors will trip

based on both steady radiation rate and time integrated total radiation dose levels. The

detectors will have a stated trigger response of 1 milligray per hour (mGy/hr) (0.1 rad/hr) as a

gamma-radiation rate meter detector. Based on this design and the guidance provided in

Appendix B of ANSI/ANS-8.3 (ANSI/ANS, 1997a), the radius of detection must be less than 

106 m (348 ft). Because of building-steel spacing and equipment arrangement, as well as a

desire to maintain a factor of two safety margin, a radius of detection of 40 m (131 ft) is used in

the design of the facility. This ensures that the CAAS is capable of detecting a criticality that

produces an absorbed dose, in soft tissue, of 0.2 Gy (20 rads) of combined neutron and gamma

radiation at an unshielded distance of 2 m (6.6 ft) from the reacting material, within 1 minute. 

The CAAS will be uniform throughout the facility for the type of radiation detected, the mode of

detection, the alarm signal, and the system dependability. When tripped, the CAAS will

automatically initiate a clearly audible signal in areas that must be evacuated. The CAAS will be

provided with emergency power and is designed to remain operational during credible events or

conditions, including fire, explosion, corrosive atmosphere, or seismic shock 

(i.e., equivalent to the site-specific design-basis earthquake or the equivalent value specified by

the uniform building code). After the CAAS is installed, whenever the CAAS is not functional,

compensatory measures will be implemented (e.g., limiting access, restricting special nuclear
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material movement). Should the CAAS coverage be lost and not restored within a specified

number of hours, the operations will be rendered safe by shutdown and quarantine, if

necessary. Onsite guidance in this respect will be provided based on process specific

considerations that consider applicable risk trade-off of the duration of reliance on

compensatory measures versus the risk associated with process upset in shutdown. Also, the

applicant provided a diagram of criticality alarm system detector locations.

Regarding the specific acceptance criteria in Sections 3.4.3.2(4)(c) and 5.4.3.4.3 of 

NUREG-1520 (NRC, 2002) for NCS criticality accident alarm system, the applicant:

• Described the method for evaluating an acceptable response of at least two detectors to

an inadvertent criticality at any location where special nuclear material may be handled,

used, or stored;

• Provided a diagram giving the locations of all detectors relative to the potential locations

of special nuclear material;

• Provided information supporting the determination of the gamma and neutron emission

characteristics of the minimum credible accident of concern capable of producing the

effects specified in 10 CFR 70.24;

• Provided information to demonstrate that the applicant’s equipment and procedures will

be adequate to ensure that specific emergency preparation requirements in 

10 CFR 70.24 will be met;

• Documented that the facility criticality accident alarm system will meet the requirements

of 10 CFR 70.24;

• Committed to the ANSI/ANS-8.3 (ANSI/ANS, 1997a) standard, as modified by NRC NCS

Regulatory Guide 3.71 (NRC, 1998);

• Committed to have a criticality accident alarm system that is uniform throughout the

facility for the type of radiation detected, the mode of detection, the alarm signal, and the

system dependability;

• Committed to have a criticality accident alarm system that is designed to remain

operational during credible events;

• Committed to have a criticality accident alarm system that is clearly audible in areas that

must be evacuated or else provide alternate notification methods that are documented to

be effective in notifying personnel that an evacuation is necessary;

• Committed to render operations safe, by shutdown and quarantine, if necessary, in any

area where criticality accident alarm system coverage has been lost and not restored

within a specified number of hours, the number of hours will be determined on a

process-by-process basis, because shutting down certain processes, even to make

them safe, may carry a larger risk than being without a criticality accident alarm system

for a short time, and there will be compensatory measures when the criticality accident

alarm system is not functional;
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• Committed to ANSI/ANS-8.23 (ANSI/ANS, 1997d), as they relate to NCS;

• Provided a facility Emergency Plan (LES, 2005b);

• Committed to provide fixed and personnel accident dosimeters in areas that require a

criticality accident alarm system. These dosimeters will be readily available to personnel

responding to an emergency; and

• Committed to provide emergency power for the criticality accident alarm system.

Based on its review of the information provided, the staff concluded that the applicant has

described an adequate NCS criticality accident alarm system to meet the requirements in 10

CFR 70.24 and 70.65(b)(4).

5.3.6.3 NCS Subcriticality of Operations and Margin of Subcriticality for Safety

Specific information about the commitments to ANSI/ANS Series-8 NCS standards is in Section

5.3.2 of this SER. Information that demonstrates that the applicant met the acceptance criteria

in this part of the SER are also in other parts of this section in this SER.

In Section 5.2.1 of the SAR (LES, 2005a), the applicant indicated that the MONK8A Monte

Carlo code (AEAT, 1998) was used to perform the NCS Analyses. MONK8A has accuracy over

a wide range of applications and is distributed with a generic validation database comprising of

critical experiments covering uranium, plutonium, and mixed systems over a wide range of

moderation and reflection. However, NRC does not allow a generic vendor validation to be

used as a demonstration of meeting the regulatory requirements for NCS validation. Since NRC

staff did not accept the applicant’s generic vendor validation report, the applicant provided a

specific validation report, as discussed below and in Section 5.2.1.1 of the SAR (LES, 2005a). 

In addition, by December 30, 2005, the applicant will provide NRC with a revised validation

report that meets LES' commitment to ANSI/ANS-8.1-1998 (ANSI/ANS, 1998a) and includes

details of validation that state computer codes used, operations, recipes for choosing code

options (where applicable), cross-section sets, and any numerical parameters necessary to

describe the input.

In Section 5.2.1.1 of the SAR (LES, 2005a), the applicant described the validation process. The

applicant validated the MONK8A code with the JEF2.2 cross-section library against experiments

in the 2002 version of the International Handbook of Evaluated Criticality Safety Benchmark

Experiments (ICSBEP) (NEA, 2002). This specific application validation was performed using

the uranium solution experiments from ICSBEP. Using 36 LEU solution experiments, the

calculated keff was 1.0007 +/- 0.0005. In the applicant’s “MONK8A Validation and Verification

Report” (LES, 2004), the applicant identified modifications to the descriptions of the ICSBEP

experiments that were described in the SAR (LES, 2005a). The results of the validation were

documented in the ISA. The MONK8A computer code and JEF2.2 cross-section library are

within the scope of the facility Quality Assurance Program in Appendix A of the SAR (LES,

2005a).

In Section 5.2.1.2 of the SAR (LES, 2005a), the applicant provided the basis for the keff equation

(i.e., keff = kcalc + 3Fcalc < 0.95) used at the facility. The validation process established a bias by
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comparing calculations to measured critical experiments. With the bias determined, an upper

safety limit (USL) was determined by using the following equation from NUREG/CR-6698 (NRC,

2001):

USL = 1.0 + bias - Fbias - FSM - FAoA.

The critical experiments were assumed to have a keff = 1.0. The calculated keff (from above) was

1.0007, which was > 1.0, and so the bias was positive. Since a positive bias may be nonconservative, the bias was set to zero. The Fbias (from above) was 0.0005. The arbitrary

subcritical margin, FSM, was assigned a value of 0.05. The FAoA term is an additional margin to

account for being beyond the area of applicability. Since the experiments were representative

of the specific application, the FAoA was set to zero.

Thus, the USL = 1 - 0.0005 - 0.05 = 0.9495. NUREG/CR-6698 (NRC, 2001) states that for

normal and credible conditions, kcalc + 2FCalc < USL. Since FCalc is greater than Fbias, the facility

was designed using the more conservative equation: keff = kcalc + 3Fcalc < 0.95. Thus, margin of

subcriticality for safety will be 0.05.

Regarding the specific acceptance criteria in Section 5.4.3.4.4 of NUREG-1520 (NRC, 2002) for

NCS subcriticality of operations and margin of subcriticality for safety, the applicant:

• Committed to the use of NCS controls and controlled parameters to assure that under

normal and credible abnormal conditions, all nuclear processes are subcritical, including

use of an approved margin of subcriticality for safety;

• Committed to the following policy: “Process specifications shall incorporate margins to

protect against uncertainties in process variables and against a limit being accidentally

exceeded;”

• Committed to the following standards, as they relate to these requirements, ANSI/ANS8.7 (ANSI/ANS, 1998b), ANSI/ANS-8.10 (ANSI/ANS, 1988b), ANSI/ANS-8.12

(ANSI/ANS, 1993), ANSI/ANS-8.15 (ANSI/ANS, 1995c), and ANSI/ANS-8.17 (1997b);

• Requested NRC pre-approval of administrative keff margins for normal and credible

abnormal conditions;

• Committed to determine subcritical limits for keff calculations such that 

ksubcritial = 1.0 - bias - margin, where the margin includes adequate allowance for

uncertainty in the methodology, data, and bias to assure subcriticality; and

• Committed to perform studies to correlate the change in a value of a controlled

parameter and its keff value and the studies will include changing the value of one

controlled parameter and determining its effect on another controlled parameter.

Based on its review of the information provided, the staff concludes that the applicant has

adequately described how it assures subcriticality of operations under normal and credible

abnormal conditions and has defined an adequate margin of subcriticality for safety to meet the

requirements of 10 CFR 70.61(d).
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5.3.6.4 NCS Baseline Design Criteria

In Section 3.1.7(I) of the ISA Summary (LES, 2005c), the applicant indicated that safety margins

and control methods are used at the facility for criticality control to meet the NCS baseline

design criteria in 10 CFR 70.64(a)(9). The design of process and storage systems will include

demonstrable margins of safety for the NCS parameters that are commensurate with the

uncertainties in the process and storage conditions, in the data and methods used in

calculations, and in the nature of the immediate environment under accident conditions. All

process and storage systems will be designed and maintained with sufficient factors of safety to

require at least two unlikely, independent, and concurrent changes in process conditions before

an inadvertent criticality is possible. The major NCS controlling parameters used in the facility

are enrichment control, geometry control, moderation control, or limitations on the mass as a

function of enrichment.

Regarding the specific acceptance criteria in Section 5.4.3.4.5 of NUREG-1520 (NRC, 2002) for

NCS baseline design criteria, the applicant committed to the double contingency principle in

determining NCS controls and IROFS in the design of new facilities or new processes that

require a license amendment under 10 CFR 70.72.

Based on its review of the information provided, the staff concludes that the applicant has

described criticality control, including adherence to the double contingency principle, to meet the

NCS baseline design criteria of 10 CFR 70.64(a)(9).

5.3.6.5 NCS in the Integrated Safety Analysis (ISA) Program and the ISA Summary

In Chapter 3 of this SER, NRC reviewed the ISA program that was used to meet 10 CFR 70.62. 

NRC reviewed the ISA method that the applicant used for NCS when performing the ISA of the

process accident sequences. In this ISA method, the applicant identified the hazard, developed

a risk matrix with consequence and likelihoods, defined “highly unlikely” and developed IROFS

and general management measures to make NCS accident sequences meet “highly unlikely.” 

The consequences of an inadvertent criticality were conservatively assumed to be high for the

workers. For accident sequences postulated to result in an inadvertent criticality, IROFS were

specified to ensure subcriticality under all normal and credible abnormal conditions and general

management measures were specified. NRC reviewed the ISA method that the applicant used

for NCS when performing the ISA for safe-by-design components. In this ISA method, “highly

unlikely” is achieved with a significant margin and other conditions (i.e., there is no credible way

to change the applicable geometric parameters without effecting a design change), rather than

with accident sequences, IROFS, and management measures.

In Chapter 3 of this SER, NRC reviewed the ISA Summary (LES, 2005c) related to whether the

applicant met certain general contents of the ISA Summary in 10 CFR 70.65. However, the

review to determine whether the applicant met other contents of the ISA Summary for NCS are

reviewed in this section of the SER.

5.3.6.5.1 Demonstration of Meeting the Contents of an ISA Summary for NCS

NRC staff reviewed the ISA Summary (LES, 2005c) as well as classified documents pertaining

to meeting the performance requirements when using both the ISA method for NCS and the

safe-by-design ISA method for NCS. The focus of the NRC review was on the risk-significant
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NCS aspects in the implementation of both the ISA method for NCS and the safe-by-design ISA

method for NCS. Specific information about the commitments to NRC NCS Regulatory Guide

3.71 and the ANSI/ANS Series-8 NCS standards is in Section 5.3.2 of this SER.

5.3.6.5.1.1 NCS ISA Method

In Sections 3.6 and 3.7 of the ISA Summary (LES, 2005c), the applicant identified potential

hazards and accident sequences which could result in an inadvertent criticality. Thus, the

sequences were assumed to be high consequence and no specific consequence analysis was

performed. The hazards and accident sequences were identified using the HAZOPS method.

The applicant identified IROFS for each accident sequence and described its expected function. 

There are no IROFS that are frequently or continuously challenged. Management measures will

ensure that IROFS are designed, implemented, and maintained, as necessary, to be available

and reliable to perform their safety function when needed. IROFS will be designed, constructed,

tested, and maintained to Quality Level 1. The four types of IROFS are: Passive Engineered

Control (PEC), Active Engineered Control (AEC), Enhanced Administrative Control (EAC), and

Administrative Control (AC). For the PEC, AEC, and AC IROFS, the applicant used a generally

conservative index score compared to the guidance in NUREG-1520 (NRC, 2002). For the

EAC IROFS, the applicant included a justification for the less conservative index score within

Section 3.8 of the ISA Summary (LES, 2005c).

In Table 3.8-2 of the ISA Summary (LES, 2005c), the applicant provided a descriptive list of sole

IROFS (summarized in the table below) to meet 10 CFR 70.65(b)(8) for NCS when using the

ISA method. All six are EAC IROFS, and Table 5.3-3 below identifies the IROFS number, the

NCS process accident sequences, and the description of the IROFS.

In Sections 3.7 and 3.8 of the ISA Summary (LES, 2005c), the applicant provided a

demonstration of meeting “highly unlikely” for each accident sequence. The applicant provided

43 specific process NCS accident sequences. NRC sorted them into 12 general process NCS

accident sequence groups that have similar characteristics (i.e., description, initiating event,

IROFS, and index scores). The applicant used either AC IROFS or EAC IROFS along with

independent verification for 11 of the 12 groups. The applicant used sampling as part of 4 of

the 12 groups.

NRC reviewed the 12 groups of sequences and determined the following:

• Group 1 (sequence PT2-2, uses enrichment control):

For this sequence, a criticality could occur if a product cylinder were placed in a feed

station. The sequence is well described with appropriate specific IROFS and

appropriate index numbers for the initiating event and IROFS failures. The PEC IROFS

is extremely important and plays a significant role in minimizing the overall risk. Also,

the AC IROFS are appropriately incorporated such that they contribute to minimizing the

risks.

Table 5.3-3

Identification of IROFS
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IROFS APPLICANT IDENTIFIED

SEQUENCE(S)

DESCRIPTION OF IROFS

14a FR1-1, FR2-1, DS1-1, DS2-1,

DS3-1,SW1-1, LW1-2

Verify use of safe-by-design transfer frame prior to movement.

14b FR1-2, FR2-2, DS1-2, DS2-2,

DS3-2, SW1-2, LW1-3

Verify acceptable use of storage array prior to movement.

15 PT3-5 Prevent additional transfer of material if the container has material in

it and it is a non-safe-by-design container.

16a PB2-2, CP1-2 Allow no visible oil and limit cylinder vapor pressure.

45 PB1-3, RD1-1 Verify non-stacked condition and no other cylinder in movement prior

to moving cylinder.

C6 EC3-1 Calculate and set cascade enrichment control device.

• Group 2 (sequences PT2-3 and PB2-5, uses moderator control):

In this group, a criticality could occur if excessive moderator is introduced into a cylinder. 

The group is clearly described with general IROFS and appropriate index numbers for

the initiating event and IROFS failures. The two AC IROFS are appropriately

incorporated such that they contribute to minimizing the risks.

• Group 3 (sequences PT2-5 and PB2-6, uses moderator control):

In this group, a criticality could occur if moderator is introduced into a cylinder. The

group is clearly described with general IROFS and appropriate index numbers for the

initiating event and IROFS failures. The two AC IROFS are appropriately incorporated

such that they contribute to minimizing the risks.

• Group 4 (sequences PT3-1, PT3-3, PB2-3, PB3-1, PB3-2, PB4-5, VR1-1, VR1-2, CL3-1,

CP1-1, EC4-2, uses mass control):

In this group, a criticality could occur if product forms on a filter over time. The group is

clearly described with general IROFS and appropriate index numbers for the initiating

event and IROFS failures. The two AEC IROFS are important and play a significant role

in minimizing the overall risk.

• Group 5 (sequence PT3-5, uses item control):

For this sequence, a criticality could occur if enriched material is transferred many times

into a non-safe-by-design container. The sequence is clearly described with a general

IROFS and appropriate index numbers for the initiating event and IROFS failure. This

sequence has a sole NCS EAC IROFS. The NRC finds this acceptable because the

failure of the IROFS would have to occur many times before a criticality could occur. 

The multiple conditions that would have to exist for enriched material to be in the

container to begin with also contribute to minimizing the risks.

• Group 6 (sequences PB1-3, RD1-1, uses interaction control):
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In this group, a criticality could occur if more than one product cylinder is moved and

stacked. The group is clearly described with a general IROFS and appropriate index

numbers for the initiating event and IROFS failure. This group has a sole NCS EAC

IROFS. The NRC finds this acceptable because of the limited number of cranes that

can be used to move product cylinders. The conditions that would have to exist for the

moving and stacking also contribute to minimizing the risk.

• Group 7 (sequences PB2-2, CP1-2, uses moderator control):

In this group, a criticality could occur if excessive moderator is introduced into a receiver

cylinder. The group is clearly described with a general IROFS and appropriate index

numbers for the initiating event and IROFS failure. This group has a sole NCS EAC

IROFS. The NRC finds this acceptable because the IROFS does not allow the operator

to make qualitative judgements about the amount of moderation present. The conditions

that would have to exist for moderation to exist contribute to minimizing the risks.

• Group 8 (sequence VR2-7, uses item control):

For this sequence, a criticality could occur if sufficient mass accumulates in a container. 

The sequence is clearly described with general IROFS and appropriate index numbers

for the initiating event and IROFS failures. The two AC IROFS are appropriately

incorporated such that they contribute to minimizing the risks.

• Group 9 (sequences FR1-1, FR2-1, DS1-1, DS2-1, DS3-1, SW1-1, LW1-2, uses

interaction control):

In this group, a criticality could occur if appropriate distance is not maintained when

moving a container. The group is well described with general IROFS and appropriate

index numbers for the initiating event and IROFS failure. This group has a sole NCS

EAC IROFS. The NRC finds this acceptable because this group uses a piece of

equipment that meets the alternate safe-by-design ISA method for NCS for moving the

container. The equipment is extremely important and plays a significant role in

minimizing the overall risk. Also, the conditions that would have to exist for enriched

material to exist in the container contribute to minimizing the risks.

• Group 10 (sequences FR1-2, FR2-2, DS1-2, DS2-2, DS3-2, SW1-2, LW1-3, uses

interaction control):

In this group, a criticality could occur if the appropriate distance is not maintained when

storing a container. The group is well described with general IROFS and appropriate

index numbers for the initiating event and IROFS failure. This group has a sole NCS

EAC IROFS. The NRC finds this acceptable because this group uses a piece of

equipment that meets the alternate safe-by-design ISA method for NCS for storing the

container. The equipment is extremely important and plays a significant role in

minimizing the overall risk. Also, the conditions that would have to exist for enriched

material to exist in the container contribute to minimizing the risk.

• Group 11 (sequences DS1-3, DS2-3, LW1-1, LW2-1, LW3-1, LW5-1, 
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uses mass control):

In this group, a criticality could occur if sufficient mass accumulates in a tank. The group

is clearly described with general IROFS and appropriate index numbers for the initiating

event and IROFS failures. The two AC IROFS are appropriately incorporated such that

they contribute to minimizing the risks.

• Group 12 (sequence EC3-1, uses enrichment control):

For this sequence, a criticality could occur if the enrichment setting for the centrifuge

process is not correct. The sequence is well described with a general IROFS and

appropriate index numbers for the initiating event and IROFS failure. This sequence has

a sole NCS EAC IROFS. The NRC finds this acceptable because more than one

operator would have to independently incorrectly set the controls. Also, the conditions

that would have to exist in the process contribute to minimizing the risks.

Regarding the specific acceptance criteria in Section 5.4.3.4.6 of NUREG-1520 (NRC, 2002) for

both the ISA method for NCS and the safe-by-design ISA method for NCS, the applicant:

• Identified NCS accident sequences, consequences of NCS accident sequences,

likelihoods of NCS accident sequences, and descriptions of IROFS for NCS accident

sequences or else used the safe-by-design ISA method for NCS appropriately;

• Committed to use Appendix A of ANSI/ANS-8.1 (ANSI/ANS, 1988a) and ANSI/ANS-8.1

(ANSI/ANS, 1998a) in determining NCS accident sequences; and

• Committed to ANSI/ANS-8.10 (ANSI/ANS, 1988b), as modified by NRC NCS Regulatory

Guide 3.71 (NRC, 1998), in determining the consequences of NCS accident sequences.

Based on the above review, the staff has reasonable assurance that: (1) the applicant identified

all hazards and accident sequences for NCS; and (2) the applicant’s IROFS and management

measures will ensure that it is highly unlikely for each accident sequence 

(i.e., inadvertent criticality) to occur.

The staff notes that the above analysis does not apply when using the safe-by-design ISA

method because when using the safe-by-design ISA method, “highly unlikely” is achieved with a

significant margin and other conditions (i.e., there is no credible way to change the applicable

geometric parameters without effecting a design change), rather than with accident sequences,

IROFS, and management measures (see below, in Section 5.3.6.5.1.2 of this SER).

5.3.6.5.1.2 NCS Safe-By-Design ISA Method

In Section 3.1.2 of the ISA Summary (LES, 2005c), the applicant provided a demonstration of

meeting “highly unlikely” for NCS when using the safe-by-design ISA method to meet 

10 CFR 70.65(b)(4). The demonstration of significant margin to meet “highly unlikely” was

provided for each of the components listed in Tables 3.7-6 through 3.7-21 of the ISA Summary

(LES, 2005c) in the following classified documents: ETC4009554 through ETC4009559,
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ETC40009561, ETC4009565 through ETC4009567, ETC4009609, ETC4009614, ETC4009677,

ETC4009679, ETC4009723, and ETC4009730. These classified documents are incorporated

by reference into the ISA Summary. Also, the configuration management system required by 10

CFR 70.72, which is implemented by the facility Configuration Management Program, will

ensure the maintenance of the safety function of these components and will assure compliance

with both the double contingency principle and the defense-in-depth criterion of 

10 CFR 70.64(b).

NRC reviewed the classified information above for all the applicant-identified safe-by-design

components. For each piece of favorable geometry equipment, NRC reviewed the dimensions

provided, calculated the geometry percentages for the equipment, and compared that

percentage versus the geometry percentage criteria. For each non-favorable geometry

equipment, NRC reviewed the appropriateness of the conservative assumption(s) and

compared the calculated keff value versus the keff limit criteria. Therefore, NRC determined that

the classified information met the safe-by-design criteria. The applicant slightly revised the

classified information and then confirmed that all the information in the new classified

documents met the criteria for using the safe-by-design ISA method for those components.

Based on the above review, the staff has reasonable assurance that: (1) the applicant used the

safe-by-design ISA method appropriately; and (2) it is highly unlikely for an inadvertent criticality

to occur with those safe-by-design components.

5.3.6.6 Additional NCS Program Commitments

Specific information about the commitments to ANSI/ANS Series-8 NCS standards is in Section

5.3.2 of this SER. Information that demonstrates that the applicant met the acceptance criteria

in this part of the SER are also in Section 5.3.5 of this SER as well as other parts of this section

in this SER.

In Section 5.4 of the SAR (LES, 2005a), the applicant committed to provide a program for

evaluating the NCS significance of events and for making the required notification to the NRC

Operations Center. Qualified individuals will make the determination of significance of NCS

events. The determination of loss or degradation of IROFS or double contingency principle

compliance will be made against the Materials License and Appendix A to Part 70. The

reporting criteria of Appendix A to Part 70 and the report content requirements of 10 CFR 70.50

will be incorporated into the facility emergency procedures. The necessary report, based on

whether the IROFS credited were lost, irrespective of whether the safety limits of the associated

parameter(s) were actually exceeded, will be issued. If it cannot be ascertained, within 1 hour,

whether the criteria in Appendix A to Part 70, paragraph (a) or (b) apply, then the event will be

treated as a 1-hour reportable event.

Regarding the specific acceptance criteria in Section 5.4.3.4.7 of NUREG-1520 (NRC, 2002) for

additional NCS program commitments, the applicant:

• Committed to use the NCS program to promptly detect any NCS deficiencies by means

of operational inspections, audits, or investigations, and to refer to the facility’s corrective

action function any unacceptable performance deficiencies in IROFS, NCS function, or

management measures, so as to prevent recurrence;
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• Committed to support the facility change mechanism process by performing NCS

Analyses and NCS Evaluations to evaluate changes to processes, operating

procedures, IROFS, and management measures;

• Committed to upgrade the NCS program to reflect changes in the ISA or new NCS

methodologies and to modify operating and maintenance procedures in ways that could

reduce the likelihood of occurrence of an inadvertent criticality;

• Committed to retain records of NCS programs and to document any corrective actions

taken;

• Committed to use the NCS methodologies and NCS technical practices to evaluate NCS

accident sequences in operations and processes;

• Committed to have a program for evaluating the NCS significance of NCS events, to

have an apparatus in place for making the required notification to the NRC Operations

Center, to have the determination of NCS significance made by qualified individuals, and

to have the determination made against the Materials License and the Regulations;

• Committed to incorporate the reporting criteria of Appendix A and the report contents of

10 CFR 70.50 into the facility emergency procedures;

• Committed to issue the necessary report based on whether the IROFS credited were

lost, irrespective of whether the safety limits of the associated parameters were actually

exceeded; and

• Committed to the following: “If the applicant cannot ascertain within one hour of whether

the criteria of 10 CFR 70 Appendix A Paragraph (a) or (b) apply, the event will be treated

as a one-hour reportable event.”

Based on its review of the information provided, the staff finds the applicant’s information

regarding additional NCS program commitments to be acceptable and the staff concludes that

the applicant has described reporting criteria to meet the requirements of Appendix A to 10 CFR

Part 70.

5.4 EVALUATION FINDINGS

NRC staff reviewed the NCS and other information for the facility according to Chapters 3 and 5

of NUREG-1520 (NRC, 2002). The staff determined that:

1. The applicant will have in place a staff of managers, supervisors, engineers, process

operators, and other support personnel who are qualified to develop, implement, and

maintain the NCS program in accordance with the facility organization and

administration, as well as management measures, according to the requirements in 

Part 70, including those in 10 CFR 70.62 and 70.65;

2. The applicant's conduct of operations will be based on NCS methods and NCS technical

practices, which will ensure that the fissile material will be possessed, handled, stored,
5-37

and used safely, according to the requirements in Part 70, including those in 10 CFR

70.22, 70.62, 70.65, and 70.72;

3. The applicant will develop, implement, and maintain a criticality accident alarm system in

accordance with the facility emergency management program and the requirements in

Part 70, including those in 10 CFR 70.24 and 70.65(b)(4);

4. The applicant will have in place an NCS program, in accordance with the requirements

in Part 70, including those subcriticality of operations, margin of subcriticality for safety,

and selection-of-controls requirements in 10 CFR 70.61(d);

5. The applicant will have in place an NCS program, in accordance with the requirements

in Part 70, including those baseline design criteria for NCS in 10 CFR 70.64(a)(9); and

6. The applicant will have in place an NCS program, in accordance with the reporting

requirements in Part 70, including those in 10 CFR 70.52 and Appendix A to Part 70.

Based on this NCS review, the staff concludes that the applicant's NCS program meets the

requirements of Part 70 and provides reasonable assurance for the protection of public health

and safety, including workers and the environment.

NRC staff reviewed the NCS and other information for the facility according to Chapter 3 of

NUREG-1520 (NRC, 2002) for NCS. The staff determined that:

1. The applicant has in place an ISA program, that is adequately established and will be

maintained pursuant to 10 CFR 70.72, such that the performance requirements for NCS

will be met in the future, if changes are made;

2. The applicant has adequately performed an ISA for NCS;

3. The applicant has adequately documented the ISA results in an ISA Summary for NCS;

4. The applicant has adequately identified and evaluated the hazards for NCS;

5. The applicant has adequately identified and evaluated the potential accident sequences

for NCS;

6. The applicant has adequately demonstrated that credible NCS accidents sequences are

highly unlikely;

7. The applicant has adequately demonstrated that the failure of safe-by-design

components for NCS is highly unlikely, as required by the regulations.

Based on this NCS review, the staff concludes that the applicant's ISA program commitments

for NCS and the ISA Summary for NCS meet the requirements of 10 CFR Part 70 and provide

reasonable assurance for the protection of public health and safety, including workers and the

environment.
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5.5 REFERENCES

(AEAT, 1998) AEA Technology (AEAT). “MONK: A Monte Carlo Program for Nuclear Criticality

Safety and Reactor Physics Analyses,” Version 8, 1998.

(ANSI/ANS, 1988a) American National Standards Institute/American Nuclear Society

(ANSI/ANS). ANSI/ANS-8.1, "Nuclear Criticality Safety in Operations with Fissionable Materials

Outside Reactors," 1988.

(ANSI/ANS, 1988b) American National Standards Institute/American Nuclear Society

(ANSI/ANS). ANSI/ANS-8.10, “Criteria for Nuclear Criticality Safety Controls in Operations with

Shielding and Confinement,” 1988.

(ANSI/ANS, 1991) American National Standards Institute/American Nuclear Society

(ANSI/ANS). ANSI/ANS-8.20, “Nuclear Criticality Safety Training,” 1991.

(ANSI/ANS, 1993) American National Standards Institute/American Nuclear Society

(ANSI/ANS). ANSI/ANS-8.12, “Nuclear Criticality Control and Safety of Plutonium-Uranium Fuel

Mixtures Outside Reactors,” 1993.

(ANSI/ANS, 1995a) American National Standards Institute/American Nuclear Society

(ANSI/ANS). ANSI/ANS-8.6, “Safety in Conducting Subcritical Neutron-Multiplication

Measurements In Situ,” 1995.

(ANSI/ANS, 1995b) American National Standards Institute/American Nuclear Society

(ANSI/ANS). ANSI/ANS-8.9, “Nuclear Criticality Safety Criteria for Steel-Pipe Intersections

Containing Aqueous Solutions of Fissile Materials,” 1995.

(ANSI/ANS, 1995c) American National Standards Institute/American Nuclear Society

(ANSI/ANS). ANSI/ANS-8.15, “Nuclear Criticality Control of Special Actinide Elements,” 1995.

(ANSI/ANS, 1995d) American National Standards Institute/American Nuclear Society

(ANSI/ANS). ANSI/ANS-8.21, “Use of Fixed Neutron Absorbers in Nuclear Facilities Outside

Reactors,” 1995.

(ANSI/ANS, 1996a) American National Standards Institute/American Nuclear Society

(ANSI/ANS). ANSI/ANS-8.5, “Use of Borosilicate-Glass Raschig Rings as a Neutron Absorber

in Solutions of Fissile Material,” 1996.

(ANSI/ANS, 1996b) American National Standards Institute/American Nuclear Society

(ANSI/ANS). ANSI/ANS-8.19, “Administrative Practices for Nuclear Criticality Safety,” 1996.

(ANSI/ANS, 1997a) American National Standards Institute/American Nuclear Society

(ANSI/ANS). ANSI/ANS-8.3, “Criticality Accident Alarm System,” 1997.

(ANSI/ANS, 1997b) American National Standards Institute/American Nuclear Society

(ANSI/ANS). ANSI/ANS-8.17, “Criticality Safety Criteria for the Handling, Storage, and

Transportation of LWR Fuel Outside Reactors,” 1997.
5-39

(ANSI/ANS, 1997c) American National Standards Institute/American Nuclear Society

(ANSI/ANS). ANSI/ANS-8.22, “Nuclear Criticality Safety Based on Limiting and Controlling

Moderators,” 1997.

(ANSI/ANS, 1997d) American National Standards Institute/American Nuclear Society

(ANSI/ANS). ANSI/ANS-8.23, “Nuclear Criticality Accident Emergency Planning and

Response,” 1997.

(ANSI/ANS, 1998a) American National Standards Institute/American Nuclear Society

(ANSI/ANS). ANSI/ANS-8.1, "Nuclear Criticality Safety in Operations with Fissionable Materials

Outside Reactors," 1998.

(ANSI/ANS, 1998b) American National Standards Institute/American Nuclear Society

(ANSI/ANS). ANSI/ANS-8.7, “Guide for Nuclear Criticality Safety Criteria in the Storage of

Fissile Materials,” 1998.

(ASME, 1994) American Society of Mechanical Engineers (ASME). Quality Assurance (QA)

Standard NQA-1-1994, “Quality Assurance Program Requirements for Nuclear Plants,” 1994.

(LES, 2004) Louisiana Energy Services (LES). letter to U.S. Nuclear Regulatory Commission,

“MONK 8A Validation and Verification,” May 7, 2004.

(LES, 2005a) Louisiana Energy Services (LES). “National Enrichment Facility Safety Analysis

Report,” Revision 5, 2005.

(LES, 2005b) Louisiana Energy Services (LES). “National Enrichment Facility Emergency

Plan,” Revision 3, 2005.

(LES, 2005c) Louisiana Energy Services (LES). “National Enrichment Facility Integrated Safety

Analysis Summary,” Revision 4, 2005.

(NEA, 2002) Nuclear Energy Agency (NEA). NEA/NSC/DOC(95)03, “International Handbook

of Evaluated Criticality Safety Benchmark Experiments,” 2002.

(NRC, 1998) U.S. Nuclear Regulatory Commission (NRC). NRC Regulatory Guide 3.71,

“Nuclear Criticality Safety Standards for Fuels and Material Facilities," 1998.

(NRC, 2001) U.S. Nuclear Regulatory Commission (NRC). NUREG/CR-6698, “Guide for

Validation of Nuclear Criticality Safety Calculational Methodology,” 2001.

(NRC, 2002) U.S. Nuclear Regulatory Commission (NRC). NUREG-1520, “Standard Review

Plan for the Review of a License Application for a Fuel Cycle Facility,” 2002.
6-1

6.0 CHEMICAL PROCESS SAFETY

The primary purpose of this review is to determine that the applicant has designed a facility that

will adequately protect workers, public, and the environment during normal operations, and also

against chemical hazards of licensed material and its byproducts. It must also protect against

facility conditions or operator actions that can affect the safety of licensed materials and thus

present an increased chemical risk.

6.1 REGULATORY REQUIREMENTS

The regulatory bases for the review are the general and additional contents of an application

that addresses chemical-process safety, as required by 10 CFR 30.33, 10 CFR 40.32, 10 CFR

70.22, and 70.65. In addition, the chemical-process safety review should provide a

determination of compliance with 10 CFR 70.61, 70.62, and 70.64.

6.2 REGULATORY ACCEPTANCE CRITERIA

The acceptance criteria for the U.S. Nuclear Regulatory Commission’s (NRC's) review of

chemical-process safety for the proposed facility are outlined in Section 6.4.3 of NUREG-1520

(NRC, 2002). 

6.3 STAFF REVIEW AND ANALYSIS

The NRC staff reviewed the Safety Analysis Report (SAR) (LES, 2005a) and the Integrated

Safety Analysis (ISA) (LES, 2005b) submitted by LES (the applicant) and considered the

following areas:

1. Chemical Process Description;

2. Chemical Accident Sequences;

3. Chemical Accident Consequences;

4. Chemical-Process Items Relied on for Safety (IROFS);

5. Management Measures; and

6. Baseline Design Criteria (BDC).

The staff reviewed the applicant’s responses to requests for additional information and ISA

documents during an on-site visit, as necessary, to have a better understanding of the process

and safety requirements. The staff evaluated the information to determine if the facility’s design

complied with the BDC and defense-in-depth requirements specified in 10 CFR 70.64(a) and

70.64(b), respectively. The evaluation is summarized in the following sections.

6.3.1 Process Description
6-2

The plant process is designed to enrich natural uranium hexafluoride (UF6) by separating a feed

stream containing the naturally occurring proportions of uranium isotopes into a product stream

enriched in the uranium-235 (U 235) isotope and a tails stream depleted in the U 235 isotope. The

process, entirely physical in nature, is a mechanical separation of isotopes using a fast rotating

cylinder (centrifuge) based on a difference in centrifugal forces from differences in molecular

weight of the uranic isotopes. No chemical changes nor nuclear reactions take place. The

feed, product, and tails streams are all in the form of UF6. 

The nominal capacity of the facility is 3 million separative work units (SWUs) per year. The

maximum gross output of the facility is sightly greater than 3 million SWUs thus allowing for a

production margin for centrifuge failures and occasional production losses during the

operational lifetime of the facility.

6.3.1.1 Gas Centrifuge Process

The enrichment process proposed by the applicant, housed in the Separation Building, is

comprised of four major systems: a UF6 Feed System, a Cascade System, a Product Take-off

System, and a Tails Take-off System. Other product-related functions include the Product

Liquid Sampling System and Product Blending System. Supporting functions include sample

analysis, equipment decontamination and equipment rebuild, liquid effluent treatment, and solid

waste management.

The major equipment used in the UF6 feed process are Solid Feed Stations. UF6 is delivered to

the plant in American Nuclear Standards Institute (ANSI) N14.1 (ANSI, 1995) standard Type

48X or 48Y international transit cylinders. Feed cylinders are loaded into Solid Feed Stations,

vented for removal of light gases, primarily air and hydrogen fluoride (HF), and heated to

sublime the UF6. 

The light gases and UF6 gas generated during feed purification are routed to the Feed

Purification Subsystem. The major pieces of equipment in the Feed Purification Subsystem are

UF6 Cold Traps, a Vacuum Pump/Chemical Trap Set, and a Low-Temperature Take-off Station

(LTTS). The UF6 is captured in UF6 cold traps and ultimately recycled as feed, whereas HF is

captured on chemical traps.

After purification, gaseous UF6 is flow-controlled through a pressure control system for

distribution to the cascade system at subatmospheric pressure. Individual centrifuges are not

able to produce the desired product and tails concentration in a single step. They are grouped

together in series and in parallel to form arrays known as cascades. A typical cascade is

comprised of many centrifuges. Each centrifuge has a thin-walled, vertical, cylindrically shaped

rotor that spins around a central post within an outer casing. Feed enters, and product and tails

streams leave the centrifuge through the central post. Control valves, restriction orifices, and

controllers provide uniform flow of product and tails.

Depleted UF6 from the cascades is desublimed in a Tails Take-off System comprised of vacuum

pumps, Tails LTTS, and Vacuum Pumps/Chemical Trap Sets. Depleted UF6 exiting the

cascade is transported into Uranium Byproduct Cylinders (UBCs) for desublimation at

subatmospheric pressure. The primary components of the Tails Take-off System are the

vacuum pumps and the Tails LTTS. Chilled air flows over the cylinder in the Tails LTTS to
6-3

effect the desublimation. Filling of the cylinders is monitored with a load cell system, and filled

cylinders (solid UF6) are transferred to an outdoor storage area (UBC Storage Pad).

Enriched UF6 from the cascades is desublimed in a Product Take-off System comprised of

vacuum pumps, Product LTTS, UF6 cold traps, and Vacuum Pumps/Chemical Trap Sets. The

vacuum pumps transport the UF6 from the cascades to the Product LTTS at subatmospheric

pressure. The heat of desublimation of the UF6 is removed by cooling air routed through the

LTTS. The product stream normally contains small amounts of light gases that have passed

through the centrifuges. Therefore, a UF6 Cold Trap and Vacuum Pump/Trap Set are provided

to vent these gases from the product cylinder. Any UF6 captured in the cold trap is periodically

transferred to another product cylinder for use as product or blending stock. Filling of the

product cylinders is monitored with a load cell system, and filled cylinders (solid UF6) are

transferred to the Product Liquid Sampling System for sampling.

Sampling is performed to verify product assay level (weight percent U 235). The Product Liquid

Sampling Autoclave is an electrically heated, closed-pressure vessel used to liquify the UF6 and

allow collection of a sample. The autoclave is fitted with a hydraulic tilting mechanism that

elevates one end of the autoclave so that liquid UF6 pours into a sampling manifold connected

to the cylinder valve. After sampling, the autoclave is brought back to the horizontal position

and the cylinder is indirectly cooled by water flowing through coils located on the outer shell of

the autoclave.

With the exception of liquid-sampling operations, the entire enrichment process operates at

subatmospheric pressure. This safety feature helps ensure that releases of UF6 or HF are

minimized, because leakage would typically be inward to the system. During sampling

operations, UF6 is liquified within an autoclave that provides the heating required to homogenize

the material for sampling. The autoclave is an American Society of Mechanical Engineers

(ASME), Section VIII, Division 1, “Boiler and Pressure Vessel Code”-rated pressure vessel that

serves as a secondary containment for the UF6 product cylinder while the UF6 is in liquid state.

6.3.1.2 Chemical Screening and Classification

The applicant classifies all site chemicals based on their potential for harm by placing each

chemical or related waste into one of three categories: Chemicals of Concern (Class 1);

Interaction Chemicals (Class 2); or Incidental Chemicals (Class 3).

Chemicals of Concern (Class 1)

The applicant determined the Chemicals of Concern (Class 1) based on one or more

characteristics of the chemical or the quantity in storage/use at the facility. For licenced

materials or hazardous chemicals produced from licensed materials, chemicals of concern are

those that, in case of release, have the potential to exceed any of the concentrations defined in

10 CFR 70.61(b) and 70.61(c).

Chemicals of concern that are not related to licensed materials are those that are listed and

above the specified threshold quantities of either 29 CFR 1910.119, “OSHA Process Safety

Management,” or 40 CFR Part 68, “EPA Risk Management Program.”
6-4

These chemicals represent, based on their inherent toxic, reactive, or flammable properties, a

potential for a severe chemical release or acute chemical exposure to an individual that:

1. Could endanger the life of a worker; or

2. Could lead to irreversible or other serious, long-lasting health effects to any individual

located outside the controlled area.

UF6 is the only licensed material-related chemical of concern (Class 1) that the applicant will be

using at the facility. There are no non-licensed chemicals of concern at the facility.

Interaction Chemicals (Class 2)

Interaction chemicals (Class 2) are those chemicals/chemical systems that require evaluation

for their potential to precipitate or propagate accidents in chemicals of concern (Class 1)

systems, but by themselves are not chemicals of concerns. Class 2 chemicals are listed in

Table 6.1-1 of the SAR (LES, 2005a).

Incidental Chemicals (Class 3)

The facility will use other chemicals that are neither chemicals of concern nor interaction

chemicals. Some of these incidental chemicals (Class 3) include those that have the potential

to result in injurious occupational or environmental exposure, but represent no potential for

acute exposure to the public and that, via their nature, quantity, or use, have no potential for

affecting chemicals of concern (Class 1). Class 3 chemicals are listed in Table 6.1-1 of the SAR

(LES, 2005a), but do not impact the performance requirements of 10 CFR 70.61.

6.3.1.3 Hazardous Chemicals and Chemical Interactions

The only chemical present in significant quantities in the plant is UF6, and it constitutes the main

hazard in this facility. Any UF6 that is released to the environment will react exothermically with

water vapor, present in air, producing uranyl fluoride (UO2F2) and HF. The chemical form and

rate of reaction of the products of this reaction will depend on the temperature and the relative

humidity of the air at the time of the release. UO2F2 is a solid whereas HF is a gas (see Eq. 1). 

UF6(gas) + 2H2O(vapor) v UO2F2(solid) + 4HF(gas) + Heat (1)

At room temperature, hydrated forms of UO2F2 and HF can be produced depending on the

relative humidity in the air (see Eq. 2). 

 

UF6(gas) + (2+4x)H2O(vapor) v UO2F2CH2O(solid) + 4HF(gas) + Heat (2)

In both cases, UO2F2 compounds are deposited or precipitated close to the point of the release.

6.3.1.4 UF6 and Interaction Chemicals

Interaction chemicals represent potential hazards, depending on the quantities present in the

facility. 
6-5

Perfluorinated Polyether (Fomblin) Oil

Beside the interaction with moisture, UF6 can react exothermically with hydrocarbons. Gaseous

UF6 reacts with hydrocarbons to form a black residue of uranium-carbon compounds. 

Hydrocarbons can be explosively oxidized if they are mixed with UF6 in the liquid phase or at

elevated temperatures. The applicant will not be using non-fluorinated hydrocarbon lubricants

in any UF6 system at the proposed facility.

The applicant’s UF6 vacuum pumps are lubricated using perfluorinated polyether (PFPE) oil that

is commonly referred to by a manufacturer’s trade name - Fomblin oil. Fomblin oil is inert, fully

fluorinated, and does not react with UF6 under any operating conditions.

Small quantities of uranium compounds and traces of hydrocarbons may be contained in the

Fomblin oil. The UF6 degrades in the oil or reacts with trace hydrocarbons to form crystalline

compounds - primarily UO2F2 and uranium tetrafluoride (UF4) particles - that gradually thicken

the oil and reduce pump capacity.

Recovery of Flombin oil for reuse in the system is conducted remotely from the UF6 process

systems. The dissolved uranium compounds are removed through a process of precipitation,

centrifugation, and filtration. Anhydrous sodium carbonate (Na2CO3) is added to contaminated

Fomblin oil. Uranium compounds react to form sodium uranyl carbonate, which precipitates out. 

A filter removes the precipitate during subsequent centrifugation of the oil.

Trace amounts of hydrocarbons are then removed by adding activated carbon to the Fomblin oil

and heating, causing absorption of the hydrocarbons. The carbon is in turn removed through a

bed of celite.

Chemical Traps - Activated Carbon, Aluminum Oxide, and Sodium Fluoride

The applicant uses chemisorption for the removal of UF6 and HF from gaseous effluent streams.

The applicant also uses it to remove oil mist from vacuum pumps operating upstream of

gaseous effluents ventilation systems. The applicant places absorbent materials on stationary

beds in chemical traps downstream of the various cold traps. These materials capture HF and

the trace amounts of UF6 that escape desublimation during feed purification or during venting of

residual UF6 contained in hoses or piping that is bled down before disconnection.

The applicant will be using two different types of traps. The first type of trap contains a charge

of activated carbon to capture the small amounts of UF6 that escape desublimation. A second

type of trap is used to absorb HF, since HF is not fully absorbed on carbon at low pressure. 

The second type of trap contains a charge of aluminum oxide (Al2O3).

Activated carbon cannot be used in the Contingency Dump System because the relatively high

UF6 flow rates during this non-routine operation could lead to severe overheating. A chemical

trap containing sodium fluoride (NaF) is installed in the contingency dump flow path to trap UF6. 

NaF is used because the heat of UF6 chemisorption on NaF is significantly lower than the heat

of UF6 chemisorption on activated carbon. 
6-6

Decontamination - Citric Acid

The applicant uses citric acid to decontaminate components (e.g., pumps, valves, piping) from

the residuals of UO2F2 compound layers that are present on the surfaces, once the components

are removed from the process area.

The reaction of the uranium compounds with the citric acid solution produces various uranyl

citrate complexes. The applicant is using personnel protective features for the safe handling of

decontamination chemicals and byproducts.

Nitrogen

The applicant uses gaseous nitrogen in the UF6 systems for purging and filling lines that have

been exposed to the atmosphere for any of several reasons including: connection and

disconnection of cylinders; preparing lines/components for maintenance; providing an airexcluding gaseous inventory for system vacuum pumps; and filling the interstitial space of the

liquid sampling autoclave (secondary containment) before cylinder liquefaction. Nitrogen is not

reactive with UF6 in any plant operational condition.

Silicone Oil

The applicant uses silicone oil as a heat-exchange medium for heating/chilling of various cold

traps. This oil is external to the UF6 process streams in all cases and is not expected to interact

with UF6.

Halocarbon Refrigerants

The applicant uses halocarbon refrigerants (including R23 trifluoromethane, R404A

fluoromethane blend, and R507 penta/trifluoromethane) in individual package chillers that will

provide cooling of UF6 cylinders or silicon oil heat-exchanger media for take-off stations and

cold traps. The applicant is using these halocarbon refrigerants because of their good heattransfer properties; they satisfy environmental restrictions regarding ozone depletion; and are

non-flammable. All halocarbon refrigerants are external to the UF6 process streams in all cases

and are not expected to interact with UF6.

Plant Chilled Water

The applicant uses chilled water in coils as a heat-exchanger medium for cooling of the liquidautoclave, after liquid samples have been drawn. Chilled water is external to the autoclave,

which is a secondary containment for the product cylinder and sampling piping, representing

three physical barriers between the water and the UF6, so no interaction is anticipated.

Centrifuge Cooling Water

The applicant provides centrifuge cooling water from the Centrifuge Cooling Water Distribution

System. The applicant uses this system to provide a supply of deionized cooling water to the

cooling coils of the centrifuge. This system provides stringent control over the operating

temperature of the centrifuge to enable its efficient operation. Centrifuge cooling water is

external to the UF6 process streams in all cases and is not expected to interact with UF6.
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6.3.2 Chemical Accident Sequences

ISA Summary Section 3.7, Table 3.7-1 (LES, 2005b), identifies the chemical accident

sequences, and Table 3.7.2 provides a narrative description. The chemical accident sequences

covered the Tails Take-Off, UF6 Feed, Product Take-Off, Product Blending and Liquid Sampling,

Ventilated Room, Chemical Laboratory, Cylinder Preparation Room, Contingency Dump,

Cascade, and the Centrifuge Test/Centrifuge Post-Mortem areas. A total of 36 different

chemical accident sequences were identified by the applicant. The accident sequences

covered the range of events that could result in a loss of containment of UF6 and the hazardous

chemicals produced from UF6 (i.e., hydrogen fluoride, uranyl fluoride, and interactions with

organic materials such as hydrocarbons). The accident sequences addressed both

intermediate- and high-consequence events. Staff review of the process and hazards involved

did not identify any chemical accident sequences overlooked by the applicant. The staff

concludes that the applicant has identified appropriate chemical accident sequences based on

the applicant’s use of an approved process hazards analysis method (HAZOP) to identify those

sequences and the results of the above staff review.

6.3.3 Chemical Accident Consequences

The applicant proposed chemical exposure limits (LES, 2004a), based on the soluble uranium

values referenced in NUREG-1391 (NRC, 1991), and proposed HF chemical exposure limits

(LES, 2004a) based on Acute Exposure Guideline Level (AEGL) values, in a manner consistent

with NUREG-1520, Table A-5, “Consequence Severity Categories Based on 10 CFR 70.61”

(NRC, 2002). The applicant also proposed a worker-exposure strategy that incorporates 10-

minute AEGL values for HF, as used in NUREG-1391 (NRC, 1991). The staff finds this

approach to be acceptable for the determination of compliance with the performance criteria of

10 CFR 70.61. 

The applicant used the methods prescribed in NUREG/CR-6410 (NRC, 1998) to determine the

source terms. Many source term values are in the classified portion of the ISA. Staff review of

the ISA and supporting documentation found the source term values to be reasonable. Site

boundary atmospheric dispersion factors were generated based on Regulatory Guide 1.145

(NRC, 1982). Meteorological data collected at the Midland/Odessa, Texas, National Weather

Service Station, during 1987-1991, are used. The applicant also used modeling methods for

source-term determination, release fractions, dispersion factors, and meteorological conditions,

as prescribed in Regulatory Guide 1.145 (NRC, 1982). The staff finds the applicant’s proposed

methodology for source-term determination to be acceptable.

6.3.4 Items Relied on for Safety (IROFS) and Management Measures

6.3.4.1 Chemical-Process IROFS

ISA Summary Section 3.7, Table 3.7-2 (LES, 2005b), describes each of the applicant identified

accident sequences and the specific IROFS that are applied to prevent or mitigate the

consequences of those accident sequences. ISA Summary Table 3.8-1 (LES, 2005b) describes

the safety functions of all identified IROFS and the specific accident sequences that take credit

for each IROFS. The staff reviewed the listed IROFS and the process descriptions and process

flow diagrams provided in ISA Summary Sections 3.4 and 3.5 to identify where each IROFS

would be used and how the IROFS would function to prevent or mitigate the consequences of
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the identified accident sequences. The identified IROFS provide protection to prevent a loss of

confinement of licensed material during operation of the facility. Based on this system level

review and the staff’s on-site visit to a similar gas centrifuge uranium enrichment facility in

Almelo, The Netherlands, which uses the same technology and process systems, the staff

concludes that the applicant has identified chemical-process IROFS to prevent the

consequences of accident sequences involving the chemical hazards of licensed material and

hazardous chemicals produced from licensed material.

6.3.4.2 Management Measures

The applicant has identified management measures to ensure the availability and reliability of

chemical-safety IROFS in SAR Section 6.4 (LES, 2005b). The applicant’s Quality Assurance

(QA) Program is described in Appendix A of the SAR. The applicant states that it will meet the

requirements of 10 CFR Part 50, Appendix B, “Quality Assurance Criteria for Nuclear Power

Plants and Fuel Reprocessing Plants,” and the guidelines of ASME QA Standard NQA-1-1994,

“Quality Assurance Program Requirements for Nuclear Facilities” (ASME, 1994). The applicant

further states that it will apply all sections of the QA Program to the chemical-safety IROFS. 

The following provides a brief description of the management measures applied to chemical

safety. Chapter 11 of this SER offers more detail of the management measures in the facility. 

Configuration Management

The Configuration Management proposed for the facility includes those controls that ensure that

the facility design basis is throughly documented and maintained, and that changes to the

design basis are controlled. This includes the following:

S That management commitment and staffing is appropriate to ensure configuration

management is maintained;

S That the proper QA is in place for design control, document control, and records

management; and

S That all structures, systems, and components (SSCs), including IROFS, are under

appropriate configuration management.

Maintenance

The applicant proposes to help maintain chemical process safety through the implementation of

administrative controls that ensure that process system integrity is maintained and that IROFS

and other engineered controls are available and operate reliably. These controls include

planned and scheduled maintenance of equipment and controls so that design features will

function when required. Appropriate plant management is responsible for ensuring the

operational readiness of IROFS under this control. For this reason, the maintenance function is

closely coupled to operation. The maintenance function plans, schedules, tracks, and maintains

records for maintenance activities.
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Maintenance activities usually fall into the following categories:

1. Surveillance/monitoring;

2. Corrective maintenance;

3. Preventive maintenance; and

4. Functional testing.

Training

The applicant proposes to provide training to individuals who handle licensed materials and

other chemicals at the facility. The training program is developed and implemented with input

from the chemical safety staff, training staff, and management. The program includes the

following:

1. Analysis of jobs and tasks to determine what a worker must know to perform tasks

efficiently;

2. Design and development of learning objectives, based on the analysis of jobs and tasks

that reflect the knowledge, skills, and abilities needed by the worker;

3. Design and development of qualification requirements for positions where a level of

technical capability must be achieved and demonstrated for safe and reliable

performance of the job function;

4. Development and implementation of standard and temporary operating procedures;

5. Development and implementation of proper inspection, test, and maintenance programs

and procedures;

6. Development of chemical safety awareness throughout the facility so that all individuals

know what their roles and responsibilities are in coordinating chemical-release mitigation

activities, in support of the Emergency Plan, in case of a severe chemical release; and

7. Coordination of the chemical-process safety training curriculum with that of other areas,

including radiological safety, criticality safety, facility operations, emergency response,

and related areas.

Procedures

The applicant proposes to use four types of plant procedures to control activities: (1) operating

procedures; (2) administrative procedures; (3) maintenance procedures; and (4) emergency

procedures.

Operating procedures, developed for workstation and control room operators, are used to

directly control process operations. Operating procedures include:
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1. Direction for normal operations, including start-up and some testing, operation, and

shutdown, as well as off-normal conditions of operation, including alarm response;

2. Required actions to ensure radiological and nuclear criticality safety, chemical safety,

fire protection, emergency planning, and environmental protection;

3. Operating limits, controls, and specific direction regarding administrative controls to

ensure operational safety; and

4. Safety checkpoints such as hold points for radiological or criticality safety checks, QA

verifications, or operator independent verification.

The applicant uses administrative procedures to perform activities that support the process

operations, including, but not limited to, management measures such as:

1. Configuration management;

2. Nuclear criticality, radiation, chemical, and fire safety;

3. QA;

4. Design control;

5. Plant-personnel training and qualification;

6. Audits and assessments;

7. Incident investigations;

8. Record-keeping and document control; and 

9. Reporting.

Administrative procedures are also used for:

1. Implementing the Fundamental Nuclear Material Control Plan (FNMCP);

2. Implementing the Emergency Plan;

3. Implementing the Physical Security Plan; and 

4. Implementing the Standard Practice Procedures Plan for the Protection of Classified

Matter.

Maintenance procedures address:

1. Preventive and corrective maintenance of IROFS;

2. Surveillance (including calibration, inspection, and other surveillance testing);
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3. Functional testing of IROFS; and 

4. Requirements for pre-maintenance activity involving review of the work to be preformed

and review of procedures.

Emergency procedures address the preplanned actions of operators and other plant personnel

in case of an emergency.

Audits and Assessments

The applicant proposes to conduct audits to determine that plant operations are performed in

compliance with regulatory requirements, license conditions, and written procedures. The

applicant assesses activities related to radiation protection, criticality safety control, hazardous

chemical safety, fire protection, and environmental protection.

The applicant performs the audits in accordance with a written plan, which identifies and

schedules audits to be performed. Audit team members have no direct responsibility for the

function and area being audited. Team members have technical expertise or experience in the

area being audited and are indoctrinated in audit techniques. The applicant conducts audits

annually, on selected functions and areas, as defined above. The chemical process safety

functions and areas will be audited at least once every 3 years.

Personnel, qualified by the applicant, who are not directly responsible for production activities,

are used to perform routine surveillance/assessments. Deficiencies noted during the

inspection, requiring corrective action, are forwarded to the manager of the applicable area or

function, for action. Future surveillance/assessments include a review to evaluate if corrective

actions have been effective.

Emergency Planning

The facility has an emergency plan and program that includes responses to mitigate the

potential impact of any process chemical release, including requirements for notification and

reporting of accidental chemical releases. The LES fire brigade/emergency response team is

outfitted, equipped, and trained for hazardous material response and local agencies can

supplement LES with additional response teams.

Incident Investigation and Corrective Actions

The applicant has a facility-wide incident investigation process that includes chemical processrelated incidents. This process is available for use by any person at the facility for reporting

abnormal events and potentially unsafe conditions or activities. Events that potentially threaten

or lessen the effectiveness of health, safety, or environmental protection will be identified and

reported to and investigated by the Health Safety and Environment Manager. Each event is

considered in terms of its requirements for reporting in accordance with regulations and is

evaluated to determine the level of investigation required. These evaluations and investigations

are conducted in accordance with approved procedures. The depth of the investigation

depends on the severity of the incident in terms of the levels of uranium/chemical released or

the degree of potential for exposure of workers, the public, or environment.
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6.3.5 Baseline Design Criteria

In the SAR (LES, 2005a) and ISA Summary (LES, 2005b), the applicant provides design basis

information for chemical-process-safety IROFS identified for the proposed facility. For chemical

protection, 10 CFR 70.64(a)(5) states:

“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 materials.”

The only chemical of concern (Class 1) is UF6. Details of design and safety features of all

chemical process systems are found in Chapter 3, ISA Summary (LES, 2005b). 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 managing the arrangement and size of material containers and

processes; selection and use of material compatible with process chemicals; providing

inherently safer operating conditions (e.g., vacuum handling); and providing process interlocks,

controls, and alarms within the process.

The staff reviewed the applicant’s proposed design of the gas centrifuge uranium enrichment

facility contained in ISA Summary Sections 3.3-3.5 and the process hazards description in

Section 3.6 (LES, 2005b). The staff notes that the uranium enrichment process is basically a

physical process that separates the U235 isotope from the U238 isotope based on their mass

difference. The uranium is in the chemical form of UF6. The entire process, with the exception

of the product sampling step, is conducted under a significant vacuum. Furthermore, the

process design involves very limited inventories of UF6 throughout the entire process. As a

result, any process leak would result in air in-leakage into the system. Any uranium that could

escape through a system breach would be limited by the available inventory and molecular

diffusion. The chemical behavior of UF6 and its reaction products (upon contact with the

moisture in air, discussed in Section 6.3.1.3 of this SER) are such that most of the uranium

bearing material is likely to accumulate near any process breach. Maintenance of the vacuum

is necessary to operate the centrifuge machines. Any significant leakage would be quickly

detected, as operation with even relatively small amounts of air would result in damage to the

machines. Based on the need to operate at, and maintain, a significant vacuum throughout the

gaseous portion of the process, and the limited inventories of licensed material contained in any

portion of the gaseous process, the staff concludes that the design basis provides for adequate

protection against chemical risks.

The staff notes that the applicant’s design of product liquid sampling system utilizes an ANSI

N14.1 qualified cylinder as the primary confinement vessel and an ASME Code pressure vessel

as a secondary confinement system. The staff concludes that this design approach for the

liquid portion of the process is acceptable because it uses internationally recognized nuclear

fuel cycle industry codes and standards. 

The staff reviewed the results of the applicant’s HAZOP analysis as discussed in SER Chapter

3. 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 NUREG-1513 (NRC,

2001). As applied to the gas centrifuge uranium enrichment process, the HAZOP considered a

variety of internal process, facility and external hazards that could breach the process and
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release licensed material and hazards chemicals produced from licensed materials. The results

of the applicant’s ISA are presented in ISA Summary Table 3.7-1 (LES, 2005b). The table

contains information concerning the accident sequences identified as a result of the HAZOP,

the unmitigated risk of each applicant identified accident sequence, and the IROFS applied to

prevent or mitigate the accident sequence. The staff also reviewed selected high-consequence

and intermediate consequence accident scenarios to confirm that chemical events that could

exceed the performance requirements of 10 CFR 70.61 were addressed. 

Based on the above, the staff concludes that the applicant’s design provides 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, and meets the requirements of 10 CFR 70.64(a)(5).

6.4 EVALUATION FINDINGS

The staff evaluated the application using the criteria previously listed. Based on the review of

the license application, the NRC staff has concluded that the applicant has described and

assessed accident consequences 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 identifies and evaluates 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 has stated that controls are maintained, available, and reliable to

perform their safety-related functions when needed. The staff has reviewed these safety

controls and the applicant's plan for managing chemical-process safety and finds them

acceptable. 

The staff concludes that the applicant's plan for managing chemical-process safety and

chemical-process-safety controls meets the requirements of Parts 30, 40, and 70, and provides

reasonable assurance that the public health and safety, and the environment, will be protected.

6.5 REFERENCES

(ANSI, 1995) American National Standards Institute (ANSI). ANSI N14.1, “Standard for

Nuclear Materials - Uranium Hexafluoride - Packaging for Transportation,” 1995.

(ASME, 1994) American Society of Mechanical Engineers (ASME). NQA-1, “Quality Assurance

Program Requirements for Nuclear Facilities,” 1994.

(LES, 2004a) Louisiana Energy Services (LES), letter to U.S. Nuclear Regulatory Commission,

“License Application and Integrated Safety Analysis Summary Update Related too Chemical

Exposure Limits,” December 22, 2004.

(LES, 2005a) Louisiana Energy Services (LES). “National Enrichment Facility Safety Analysis

Report,” Revision 6, 2005.
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(LES, 2005b) Louisiana Energy Services (LES). “National Enrichment Facility Integrated

Safety Analysis,” Revision 4, 2005.

(NRC, 1982) U.S. Nuclear Regulatory Commission (NRC). Regulatory Guide 1.145,

“Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear

Power Plants,” 1982.

(NRC, 1991) U.S. Nuclear Regulatory Commission (NRC). NUREG-1391, “Chemical Toxicity

of Uranium Hexafluoride Compare to Acute Affects of Radiation,” 1991.

(NRC, 1998) U.S. Nuclear Regulatory Commission (NRC). NUREG/CR-6410, “Nuclear Fuel

Cycle Facility Accident Analysis Handbook,” 1998.

(NRC, 2002) U.S. Nuclear Regulatory Commission (NRC). NUREG-1520, “Standard Review

Plan for the Review of a License Application for a Fuel Cycle Facility,” 2002.
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7.0 FIRE SAFETY

The purpose of this review is to determine, with reasonable assurance, whether the applicant

has designed a facility that provides adequate protection against fires and explosions that could

affect the safety of licensed materials and thus present an increased radiological risk. The

review should also establish that the applicant has considered the radiological consequences of

fires and will institute suitable safety controls to protect workers, the public, and the

environment.

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 30.33, 10 CFR 40.32, 10 CFR 70.22, and 10 CFR 70.65. In

addition, the fire safety review should focus on providing reasonable assurance of compliance

with 10 CFR 70.61, 70.62, and 70.64.

7.2 REGULATORY ACCEPTANCE CRITERIA

The acceptance criteria the U.S. Nuclear Regulatory Commission (NRC) uses for reviews of fire

safety are outlined in Sections 7.4.3.1 through 7.4.3.5 of NUREG-1520 (NRC, 2002). 

7.3 STAFF REVIEW AND ANALYSIS

7.3.1 Process Fire Hazards and Special Hazards

The applicant plans to use gas centrifuge machines to enrich uranium up to 5 weight (wt.)

percent uranium-235 (U 235). The feed material will be uranium hexafluoride (UF6), which will be

enriched by using 48 enrichment cascades with hundreds of gas centrifuge machines per

cascade. The UF6 normal feed is 0.711 wt. percent U 235; the expected product is a maximum of

5.0 wt. percent U 235; and the depleted tails are typically 0.2 wt. percent U 235. Enriched and

depleted UF6 streams are withdrawn from the cascade by pumps and returned to a solid phase

in product and tails low-temperature take-off stations, respectively. The remainder of this

section describes the NRC staff review of key fire hazards and risks associated with the

proposed facility and gas centrifuge enrichment process. 

UF6: UF6 is not flammable and does not disassociate to flammable constituents under

conditions at which it will be handled at the facility. UF6 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 UF6 in the liquid state or at elevated temperatures. 

For this reason, non-fluorinated hydrocarbon lubricants are not utilized in the UF6 processes at

the facility. UF6 pumps are lubricated using a perfluorinated polyether oil that is referred to by

the manufacturer’s trade name, Fomblin oil. 
7-2

Hydrogen Fluoride (HF): HF is a byproduct of the chemical reaction of UF6 with water vapor. 

HF is extremely reactive in both gaseous and aqueous form. HF by itself is not flammable nor

combustible. It can, however, react exothermically with water to generate sufficient heat to

ignite nearby combustibles.

Uranyl Fluoride (UO2F2): UO2F2 is also a byproduct of the chemical reaction of UF6 with water

vapor. UO2F2 is stable in air to 300o

 C (572o F). It is not flammable nor combustible and will not

decompose to combustible constituents under conditions that will exist at the facility.

Centrifuge Machines and Components: The model TC-12 centrifuge contains a rotor assembly,

under a vacuum, inside an aluminum outer casing. The casing also provides a vacuum

enclosure outside the rotor to reduce drag. The rotor is driven by a electromagnetic motor. The

only combustibles of any significance are the electrical cabling going to the drive motors. 

Therefore, any fire originating in one of the cascades will most likely result in limited damage to

the centrifuge and its components, resulting in a small release.

Control Room: The control room will be provided with automatic smoke detection throughout. 

Additionally, the control room will house the fire alarm control panel and will be continuously

staffed. Hand-portable fire extinguishers will be provided in accordance with National Fire

Protection Association (NFPA) Standard 10 (NFPA, 1994). Items Relied on for Safety (IROFS)

boundaries will include appropriate electrical separation from normal instrument and control

functions to ensure that fire-induced spurious actuations do not occur. Based on the current

design, all active engineered components that are IROFS will fail in the safe configuration.

Storage and Handling of UF6: UF6 cylinders are stored or handled in the Uranium Byproduct

Cylinder (UBC) storage pad; the Cylinder Receipt and Dispatch Building (CRDB); the UF6

handling areas; and the blending and liquid-sampling areas. On the UBC storage pad, fire

concerns include the cylinder transport vehicle, a fire exposure from nearby vegetation, and fire

exposure from a nearby vehicle accident. The licensee performed evaluations of 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 adequately mitigated. In the CRDB, the primary fire

concern was from a truck fire at the loading dock. The licensee also analyzed this and

determined that the cylinders could be adequately protected by storing them at least 1 meter

(m) [3.3 feet (ft)] from the edge of the loading dock. Combustible loadings in the UF6 handling

areas and the blending and liquid-sampling areas are limited and transient combustibles will be

controlled. Therefore, any fire originating in these areas will be limited. In addition, smoke

detection and an interlock with the ventilation system will limit releases to a low-consequence

event in the Ventilated Room of the Technical Services Building (TSB).

Hydrogen Control: Hydrogen is used within the TSB Chemical Laboratory and may be

generated at battery-charging stations in the facility. The laboratory will be protected by one or

more of the following features:

• Hydrogen piping will be provided with excess flow control;

• Hydrogen supply will be isolated by emergency shutoff valves interlocked with hydrogen

detection in the areas served by the hydrogen piping; and
7-3

• Natural or mechanical ventilation will be provided to ensure that hydrogen

concentrations do not exceed 25 percent of the lower explosive limit (LEL). If

mechanical ventilation is provided, it will be continuous or will be interlocked to start on

detection of hydrogen in the area. Mechanical ventilation will also be provided with

airflow sensors, to sound an alarm if the fan becomes inoperative.

Hydrogen control in battery-charging stations will be provided by measures identified in NFPA

70E (NFPA, 2004) and American National Standards Institute (ANSI) Standard C2 (ANSI,

1981). It is expected that natural or mechanical ventilation will be provided to ensure that

hydrogen concentrations do not exceed 25 percent of the LEL. If mechanical ventilation is

provided, it will be continuous or will be interlocked to start on detection of hydrogen in the area. 

Mechanical ventilation will also be provided with airflow sensors, to sound an alarm if the fan

becomes inoperative.

Combustible Material Hazards: Materials of construction for the centrifuge process building, the

supporting buildings, and centrifuge machines and components are predominantly

noncombustible (e.g., steel, aluminum, 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. The largest quantity of combustible material is

two 19,000 L (5000 gal.) tanks of diesel fuel located outside. Other quantities of combustible

materials are as follows:

• Silicone oil in the UF6 handling area and the blending and liquid sampling area is

contained within the heater and chiller units associated with the cold traps, with each

unit containing approximately 72 L (19 gal.) of oil. Some units are paired, but each pair

is located at least 9 m (30 ft) from any adjacent unit. The staff considers this distance to

be sufficient to limit the potential involvement in a fire to one pair of cold trap heater or

chiller units.

• Oxygen gas (oxidizer), acetylene gas, and propane gas in the Mechanical, Electrical,

and Instrumentation Workshop;

• Acetone, toluene, petroleum ether, and peroxide in the Chemical Laboratory; and

• Primus gas, degreaser solvent, penetrating oil, and cutting oil in the vacuum pump

rebuild workshop.

The NRC staff has reasonable assurance that the applicant has adequately identified all fire and

explosion hazards. In accordance with guidance provided by NUREG-1520 (NRC, 2002), the

applicant has reasonably identified and evaluated the facility and process hazards and risks

associated with the proposed operations. In its review, NRC has taken into account the

potential presence of these combustibles in the various accident scenarios. The identification of

fire hazards and related analyses are documented in the applicant’s “Integrated Safety Analysis

(ISA) Summary” which provides the supporting safety basis for the license application. 
7-4

7.3.2 Accident Scenarios

The applicant’s ISA Summary 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 in the Centrifuge Test Facility;

• Fire in the CRDB;

• Fire involving Cylinder Transporters/Movers

• Fire inside the Cascade Halls;

• Fire in the Process-Services Area;

• Fire inside the UF6 handling area/Blending and Liquid Sampling Area;

• Fire inside the TSB; and

• Fire affecting the UBC Storage Pad.

Exterior and interior building explosions as initiating events to accident sequences were

evaluated and found to be highly unlikely, without the need for IROFS.

7.3.2.1 Fire in the Centrifuge Test Facility

The Centrifuge Test Facility is located in the Centrifuge Assembly Building with the Centrifuge

Post-Mortem Facility. The centrifuge test facility has UF6 contained in stainless steel

components and piping. The fire loading in the centrifuge test area consists of cable insulation,

activated carbon, and a low amount of transient combustibles. Transient combustibles will be

controlled by administrative combustible loading controls (IROFS36a). With these controls and

the estimated in-situ fire loading, the applicant determined that the UF6 inventory would not be

released from the stainless steel confinement. The applicant also evaluated the likelihood of a

fire propagating into the area from an adjacent area. Fire propagation will be prevented by

rated barriers (IROFS35) between adjacent fire areas and the Centrifuge Test Facility. Based

on the above, the applicant has 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 (c)

occurs, is highly unlikely. 

The staff concludes with reasonable assurance that the applicant has demonstrated that the

facility will be in compliance with the performance requirements of 10 CFR 70.61 in the event of

a fire in the Centrifuge Test Facility.

7.3.2.2 Fire in the CRDB

All UF6 feed cylinders and empty product and byproduct cylinders enter the facility through the

CRDB. The applicant considered three separate fire scenarios in the CRDB. These fire

scenarios were a fire at the CRDB loading dock, a fire in the CRDB general areas, and a fire

involving a cylinder delivery vehicle.

At the CRDB loading dock, UF6 is contained in 48Y, 48X, and 30B cylinders on the loading dock

and scales adjacent to the dock. The most severe fire was postulated to be a vehicle fire at the

loading dock. The applicant evaluated the effect of this fire by calculation and showed that

there could be a potential fire threat to UF6 cylinders, but that this threat could be eliminated by

assuring that cylinders were stored with a 1 m (3.3 ft) setback from the edge of the loading
7-5

dock. The 1 m (3.3 ft) setback is an administrative control (IROFS36b). Based on the above,

the applicant has 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 (c) occurs, is

highly unlikely. The staff reviewed the applicant’s calculation (NRC, 2004) and agrees with this

assessment.

The CRDB General Areas contain UF6 in 48Y, 48X, and 30B cylinders. Combustible loading is

expected to be very low and transient combustibles will be controlled with IROFS36a. No liquid

combustibles are listed for the CRDB general areas. Hence, a fire with the intensity and

duration to heat a large UF6 cylinder to its critical temperature is not considered credible. Fire

propagation will be prevented by rated barriers (IROFS35) between adjacent fire areas and the

CRDB general areas. Based on the above, the applicant has 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 (c) occurs, is highly unlikely. The staff agree with this

assessment.

At the staff’s request, LES also evaluated a fire involving UF6 cylinders present on a delivery

vehicle. U.S. Department of Transportation regulations require thermal protection (e.g.,

overpack or other protective assembly), which will withstand the thermal test criteria, specified in

10 CFR 71.73(c)(4), without rupture of the containment system, for all off-site UF6 shipments. 

Hence, both incoming cylinders and outgoing cylinders will be protected by approved thermal

protection. The handling practice for incoming cylinders containing UF6 will be to offload the

integral cylinder in its protective assembly to the loading dock, before opening or removal of the

protective assembly. Outgoing cylinders will be individually loaded into a protective assembly

before placement on truck trailers. The applicant determined that the worst postulated truck fire

involving diesel fuel and other combustibles associated with a truck fire would burn for no more

than 22 minutes. Approved protective assemblies (IROFS36h) are designed to protect a

cylinder for 30 minutes in an 800°C (1472°F) hydrocarbon fire. Because of the location of the

cylinder in the assembly on the truck trailer and the duration of the fire, the UF6 cylinders will be

adequately protected from rupture from a truck fire. The staff also evaluated the applicant’s

calculation of fire duration and agrees with the result (NRC, 2005). Based on the above, the

applicant has 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 (c) occurs, is highly

unlikely. The staff agrees with this assessment.

The staff concludes with reasonable assurance that the applicant has demonstrated that the

facility will be in compliance with the performance requirements of 10 CFR 70.61 in the event of

a fire in the CRDB.

7.3.2.3 Fire Involving Cylinder Transporters/Movers

The assumed inventory for a fire involving a cylinder transporter/mover is the amount of UF6 in a

UF6 cylinder (48X, 48Y, or 30B) in transit. Only electric drive cylinder transporters will be used

for cylinder transport inside of the proposed facility’s buildings. When filled 30B cylinders are

transported outside of the buildings, they are protected by DOT approved overpacks (IROFS

36h). For other cylinders which may be transported outside of the buildings by diesel drive

transporters, the fuel capacity of the transporters is limited to less than 280 L (74 gal.). This

limit on fuel is IROFS 36c and is described in more detail in Section 7.3.2.8 of this SER. 

Based on the above, the applicant has determined that the likelihood of a fire being initiated,
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and of the IROFS failing, so that a release exceeding the consequence threshold of 10 CFR

70.61(b) or (c) occurs, is highly unlikely. 

The staff concludes with reasonable assurance that the applicant has demonstrated that the

facility will be in compliance with the performance requirements of 10 CFR 70.61 in the event of

a fire on a cylinder transporter either inside or outside of the proposed facility’s buildings.

7.3.2.4 Fire in the Cascade Halls

The six cascade halls contain the centrifuges and are located in the Separations Building. The

fire scenario inside the cascade halls is assumed to take place inside a module that holds eight

cascades, each cascade containing hundreds of centrifuges. If the entire module were engulfed

in a fire and the total inventory released, a high-consequence event would result. However, a

fire is prevented from propagating into the module by fire barriers (IROFS35). A fire originating

inside the module is presumed to involve the cables feeding the centrifuge drive motors. If

transient combustibles are controlled (IROFS36a), this fire, at worst, is expected to involve one

and possibly an adjacent centrifuge and can be assumed to result in the release of no more

than 1 percent of the inventory in the module, resulting in consequences below the threshold of

the 10 CFR 70.61(c) intermediate-consequence limit. Based on the above, the applicant has

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 (c) occurs, is highly unlikely. 

The staff concludes with reasonable assurance that the applicant has demonstrated that the

facility will be in compliance with the performance requirements of 10 CFR 70.61 in the event of

a fire in the Cascade Halls.

7.3.2.5 Fire in the Process Services Area 

The Process Services Area contains the gas transport equipment (i.e, the piping to the Product

Take-Off System and the piping to the Tails Take-Off System) and the Contingency Dump

System. The Process Services Area is located in the Separations Building. The UF6 inventory

in the Process Services Area is UF6 in feed, product, and tails piping manifolds. In addition,

there could be significantly more in the 48 sodium fluoride traps that are part of the contingency

dump system. If the entire Process Services Area were engulfed in a fire and the total inventory

released, a high-consequence event would result. However, a fire is prevented from

propagating into the Process Services Area by fire barriers (IROFS35). A fire originating in the

Process Services Area involving expected in-situ combustibles is considered capable of causing

failures in the aluminum piping and manifolds, but not in the aluminum traps. Combustible

loading controls (IROFS36a) are intended to limit additional transient combustibles. With failure

of the piping only, 50 percent of the UF6 is assumed released. This would result in

consequences that are below the 10 CFR 70.61(c) intermediate-consequence limit. Based on

the above, the applicant has 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 (c)

occurs, is highly unlikely.

The staff concludes with reasonable assurance that the applicant has demonstrated that the

facility will be in compliance with the performance requirements of 10 CFR 70.61 in the event of

a fire in the Process Services Area.

7.3.2.6 Fire in the UF6 Handling Area/Blending and Liquid Sampling Area
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The UF6 Handling Area contains the UF6 Feed System, Product Take-Off System, and Tails

Take-Off System. The Blending and Liquid Sampling Area contains the Product Liquid

Sampling System and the Product Blending System. The UF6 inventory in the blending and

liquid sampling area of the UF6 Handling Area is contained in cylinders, piping, manifolds, and

hoses. The applicant states that additional uranic material may be present on the

carbon/alumina traps that capture residual traces of UF6 from the various feed, product, and

tails system cold traps. A fire is prevented from propagating into the UF6 Handling Area, and

Blending and Liquid Sampling Area, by fire barriers (IROFS35). A fire originating in these areas

with improperly placed combustibles is considered capable of failing only a single-cylinder hose. 

A fire, involving expected in-situ and transient combustibles, could cause failure in the aluminum

piping manifold and release 50 percent of the inventory feeding one module. This would result

in consequences that are below the 70.61 (c) intermediate consequence limit. Severe fires will

be prevented by IROFS36a, which will control the location and the amount of transient

combustibles in the area. Based on the above, the applicant has 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 (c) occurs, is highly unlikely. 

The staff concludes with reasonable assurance that the applicant has demonstrated that the

facility will be in compliance with the performance requirements of 10 CFR 70.61 in the event of

a fire in the UF6 Handling Area, and the Blending and Liquid Sampling Area.

7.3.2.7 Fire in the TSB

The TSB contains support areas for the facility, such as the Solid Waste Collection Room;

Vacuum Pump Rebuild Workshop; Decontamination Workshop; Ventilated Room; Cylinder

Preparation Room; Mechanical, Electrical, and Instrumentation Workshop; TSB Gaseous

Effluent Ventilation System Room; various laboratories; the Control Room; and monitoring

centers. In the TSB, fires were postulated in all uranic material areas and IROFS were found to

be needed in the Solid Waste Collection Room, the Decontamination Workshop, the Ventilated

Room, and the Chemical Laboratory Sample Storage Room. 

The uranium inventory in the Solid Waste Collection Room is contained in 12-L (3.2-gal.) metal

containers and 210-L (55-gal.) metal drums. A fire is prevented from propagating into the Solid

Waste Collection Room by rated fire barriers (IROFS35). For a fire originating in the area, and

involving expected in-situ and transient combustibles, the applicant postulates that only a few kg

of uranic materials would be present in open containers or drums, during transfer/packing

operations, and driven off, in case of a fire. Preventive measures are to administratively limit

transient combustible loading in areas containing uranic material, to ensure integrity of uranic

components/containers and to limit the quantity of uranic material at risk, to ensure that

consequences to the public are low (IROFS36d). Based on the above, the applicant has

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 (c) occurs, is highly unlikely.

The uranium inventory in the Decontamination Workshop is contained in up to three 12-L (3.2-

gal.) metal containers and three steel tanks. A fire is prevented from propagating into the room

by fire-rated barriers (IROFS35). For a fire originating in the area, and involving expected in-situ

and transient combustibles, the applicant postulates that only a few kilograms of uranic

materials would be present in open containers or drums during transfer/packing operations, and

driven off, in case of a fire. Preventive measures are to administratively limit transient
7-8

combustible loading in areas containing uranic material, to ensure integrity of uranic

components/containers and to limit the quantity of uranic material at risk, to ensure that

consequences to the public are low (IROFS36d). Based on the above, the applicant has

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 (c) occurs, is highly unlikely. 

The uranium inventory in the ventilated room is contained in 12-L (3.2-gal.) metal and 210-L

(55-gal.) drums. Additional inventory is present in single 48Y, 48X, or 30B cylinders, present in

the room, for valve maintenance/change-out. Various fire scenarios were postulated for the

ventilated room, each with different combinations of IROFS, consequences, or likelihood. A fire

in improperly placed transient combustibles (but with IROFS36d in effect) could cause failure of

the nitrogen hose or vent-line piping used to bleed gas from cylinders during valve servicing or

subsequent nitrogen-pressure tests. This was determined to be a low-consequence event. 

Another fire, however, with expected in-situ and transient combustibles within the area, but with

the uranic inventory in open containers, which could be driven off in a fire, is considered to be

an intermediate-consequence event. This postulated accident sequence is reduced to a lowconsequence sequence by adding a second IROFS (IROFS37), which is smoke detection

interlocked to shut down the building ventilation system. The other sequences the applicant

evaluated assumed various combinations of failure of the combustible loading controls and the

smoke detector and evaluated the resulting consequence level and likelihood. Based on the

above, the applicant has 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 (c)

occurs, is highly unlikely or unlikely, respectively, in accordance with the performance

requirements

The TSB Chemical Laboratory Sample Storage Room also contains a uranium inventory that

could potentially result in a high consequence. A fire is prevented from propagating into the

Chemical Laboratory Sample Storage room by rated fire barriers (IROFS35). For a fire

originating in the Sample Storage Room with controls in place (IROFS36a), with expected insitu and transient combustibles, the consequences would be low. The fire would not have

sufficient combustibles to fail a sample cylinder. Based on the above, the applicant has

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 (c) occurs, is highly unlikely. 

The staff concludes with reasonable assurance that the applicant has demonstrated that the

facility will be in compliance with the performance requirements of 10 CFR 70.61 in the event of

a fire in the TSB.

7.3.2.8 UBC Storage Pad

The UBC Storage Pad provides storage for UBCs and 6 months of empty feed cylinders. The

UBC Storage Pad occupies approximately 9 hectares (23 acres) and is sized to accommodate

enough cylinders for 30 years of operation.

For the UBC storage pad, fires were postulated on the transporter, on the UBC perimeter road,

and in vegetation off the property. The fire on the transporter was assumed to be the same fire

that NRC staff analyzed in Section 11.3.2.6, “UF6 Storage Area Fire,” of NUREG-1491 (NRC,

1994). In this event, the NRC staff assumed a hydrocarbon pool fire with a temperature of

800°C (1500°F). NRC estimated heat transfer to the cylinder from this fire and calculated that it
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would take 20 minutes for the fire to cause rupture of a UF6 cylinder. The NRC staff further

calculated that it would require 280 L (74 gal.) of diesel fuel for the fire to burn for 20 minutes. 

The applicant used the results of the NRC staff calculation as a basis for a control under

IROFS36c, which limits fuel in a diesel-fueled UBC transporter to 280 L (74 gal). The applicant

will also employ IROFS36e for other vehicles, which administratively limits transient combustible

loading on the UBC storage pad, to ensure cylinder integrity. For the review, the staff reevaluated the estimation of burning time, assuming 280 L (74 gal) of fuel and a 3.1-m (10-ft)

long pool fire approximating a 48X cylinder. The staff calculated about 20 minutes or less of

burn time for conservative assumptions of pool configuration (NRC, 2005). Based on the

above, the applicant has 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 (c)

occurs, is highly unlikely.

The applicant performed its own calculations to determine the possible effects of a servicevehicle fire on a perimeter road at the closest point to the UBC. This fire was based on a pool,

confined by the access road, formed by 500 L (132 gal.) of diesel fuel contained in the truck

along with other truck combustibles. The heat transfer from the equivalent pool fire was

determined to be insufficient to heat the cylinder to the temperature where it might rupture. A

control for fires of this type will be IROFS36f, which will administratively limit designated routes

for bulk-fueling vehicles onsite, to ensure UBC cylinder integrity. Based on the above, the

applicant has 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 (c) occurs, is highly

unlikely.

The applicant also evaluated the fire in the off-site vegetation, by means of a heat-transfer

calculation from an assumed large fire approximately 89 m (292 ft) from the UBC pad. The heat

transfer from this fire was also found to be insufficient to heat the cylinder to a point where it

could rupture. The applicant will control this scenario by IROFS36g, which will administratively

limit on-site vegetation fire sources to ensure the integrity of important targets. Based on the

above, the applicant has 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 (c)

occurs, is highly unlikely.

The staff review of these analyses is documented in an in-office review report (NRC, 2004).

The staff concludes with reasonable assurance that the applicant has demonstrated that the

facility will be in compliance with the performance requirements of 10 CFR 70.61 in the event of

a fire in the UBC Storage Pad.

7.3.2.9 Conclusions

The NRC staff has reasonable assurance that the applicant has identified and evaluated all fire

related accident scenarios credible for the proposed centrifuge process. The applicant has

reasonably identified and evaluated possible fire initiators and consequences and has identified

IROFS for accident scenarios that could result in intermediate or large consequences leading to

unacceptable performance, in accordance with 10 CFR Part 70, and as described in guidelines

established in NUREG-1520 (NRC, 2002). 

7.3.3 IROFS Related to Fire Safety
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The NRC staff has reasonable assurance that the applicant has identified the required IROFS

for preventing or mitigating fire accident scenarios that could lead to unacceptable performance

in accordance with the requirements in 10 CFR 70.61. The applicant has 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. These

IROFS are listed in Table 7.3-1.

The NRC staff considers the failure probability indices assigned to these IROFS, with a failure

probability index of -3, to be achievable, with their respective bases, as described in Section

3.8.3 of the ISA Summary. Section 3.8.3 of the ISA Summary provides proposed surveillance

frequencies, safety margins, and other measures that will support the low-failure probabilities

assumed for these measures. All of the above-listed IROFS will also be supported by the

general management measures, as described in Section 3.1.8.3 of the ISA Summary. In

conclusion, the staff finds the selection of accident sequences and the determination of IROFS

to be acceptable to satisfy the performance requirements of 10 CFR 70.61.

The remaining features described in the license application are fire-protection measures that

provide overall defense-in-depth protection of fire safety for operations. The applicant’s ISA

Summary has adequately addressed fire risks, in accordance with the regulation and the

guidance established in NUREG-1520 (NRC, 2002).

7.3.4 Facility Fire Protection

7.3.4.1 Fire Safety Management Measures

The Health, Safety and Environmental manager is responsible for fire protection and is assisted

by the Industrial Safety Manager, who is responsible for day-to-day safe operation of the facility,

including fire safety. The Industrial Safety Manager is assisted by fire safety personnel who are

trained in fire protection and have nuclear fire-safety experience. The fire protection staff is

responsible for:

• 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 record-keeping, as related to fire protection;

• Organization and training of the fire brigade; and

• Pre-fire planning.

Fire prevention at the facility consists of administrative controls to: (a) govern the handling of

transient combustibles; (b) control ignition sources; (c) ensure that open flames or combustiongenerated smoke is not used for leak-testing; (d) conduct periodic fire prevention inspections; 

(e) perform periodic house-keeping inspections; and (f) implement a system to control the

disarming of fire-detection or fire-suppression systems. The inspection, testing, and 

maintenance of fire-protection systems will comply with industry standards. The applicant’s 

Table 7.3-1

Identification of IROFS
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IROFS Failure

probability

index

Description of IROFS

IROFS35 -3 Automatic closure of fire-rated-barrier opening protections

(e.g., doors, dampers, penetration seals), to ensure that the

integrity of area fire barriers prevents fires from propagating

into areas containing uranic material.

IROFS36a -3 Administratively limit transient combustible loading in areas

containing uranic material, to ensure integrity of uranic material

components/containers and limit the quantity of uranic material

at risk, to ensure that consequences to the public are low.

IROFS36b -3 Administratively limit storage of UF6 cylinders in the CRDB, to

ensure $ 1-m (3-ft) setback from the edge of the loading dock.

IROFS36c -3 Administratively limit on-site UF6 cylinder transporters/movers,

to ensure only use of electric drive or diesel power with a fuel

capacity of less than 280 L (74 gal.).

IROFS36d -3 Administratively limit transient combustible loading in areas

containing uranic material, to ensure integrity of uranic material

components/containers and limit the quantity of uranic material

at risk, to ensure that consequences to the public are low.

IROFS36e -3 Administratively limit transient combustible loading on the UBC

Storage Pad, to ensure cylinder integrity.

IROFS36f -3 Administratively limit designated routes for bulk-fueling

vehicles on-site, to ensure UBC cylinder integrity.

IROFS36g -3 Administratively limit on-site vegetation fire sources, to ensure

integrity of important targets.

IROFS36h -3 Use of cylinder overpack/thermal protective assemblies

IROFS37 -2 Automatic trip of the Ventilated Room HVAC and isolation from

TSB GEVS on smoke detection, and Ventilated-Room-design

leakage limited to ensure offsite exposure from building outflow maintains consequences to the public low.

ISA Summary has adequately addressed fire safety management measures, in accordance with

the guidance established in NUREG-1520 (NRC, 2002). In addition, the applicant’s fire safety

management measures meet the requirements of 10 CFR 30.33, 10 CFR 40.32, 10 CFR 70.22,

10 CFR 70.64, and 10 CFR 70.65 as they pertain to the fire protection aspects of the facility.

7.3.4.2 Facility Passive-Engineered Fire-Protection Systems
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Buildings containing UF6 are the TSB, CRDB, Cascade Halls, Blending and Liquid Sampling

Areas, and the Centrifuge Assembly Building (including the Centrifuge Test and Centrifuge Post

Mortem Area), which are of pre-cast concrete-frame and concrete-panel construction. This

construction is classified as Type I-443, in accordance with NFPA 220 (NFPA, 1999). This

means that the exterior and bearing walls, interior-bearing walls supporting more than one floor,

columns supporting more than one floor, and structural members supporting more than one

floor, have a fire resistance of at least 4 hours. All interior-bearing walls, columns, and

structural members supporting one floor only, or a roof only, shall have a fire resistance of at

least 3 hours. Floors have a resistance of at least 3 hours and roofs have a fire resistance of at

least 2 hours.

The Administration Building, Fire Water Pump Building, and the Central Utilities Building are

unprotected steel-frame buildings, with insulated metal panel exterior walls, and with built-up

roofing on a metal deck roof. The Visitor Center and the site security buildings are also

unprotected steel frame with insulated metal-panel exterior walls and built-up roofing on a metal

deck roof. The above buildings are classified as Type II-000, in accordance with NFPA 220

(NFPA, 1999). This means that all structural members are of non-combustible or limited

combustible construction, but no fire rating is required. The applicant’s ISA Summary has

adequately addressed passive engineered fire protection systems, in accordance with the

guidance established in NUREG-1520 (NRC, 2002). In addition, the applicant’s passiveengineered fire protection systems meet the requirements of 10 CFR 30.33, 10 CFR 40.32, 10

CFR 70.22, 10 CFR 70.64, and 10 CFR 70.65 as they pertain to the fire protection aspects of

the facility.

7.3.4.3 Facility Active-Engineered Fire-Protection Systems

Fire Alarm System

Each building of the facility is equipped with a listed modular, multi-zone fire alarm control panel

installed in accordance with NFPA 72 (NFPA, 1996d). Each panel has a dual power supply,

consisting of normal building power and backup power, by either 24-hour battery or the facility

uninterruptible power supply (UPS). Sprinkler system and hose-station water-flow detection are

connected to separate control-panel zone modules. Fire-detector and manual-pull station-alarm

circuits are also on separate modules. Each zone module has separate alarm and trouble

contacts for connection to the central alarm panel in the Control Room. Activation of a fire

detector, manual-pull station, or water-flow detector results in an audible and visual alarm at the

building control panel and the central alarm panel.

The central alarm panel, located in the control room, is a listed, microprocessor-based

addressable console. The central alarm panel has dual power supplies, consisting of normal

building power and backup power by either 24-hour battery or the facility UPS. The central

alarm panel monitors all functions associated with the individual alarm panels and fire pump

controllers. All alarm and trouble functions are audibly and visually annunciated by the central

alarm panel and automatically recorded via printout. Failure of the central alarm panel will not

result in failure of any building-alarm control-panel functions.

The following conditions are monitored by the central alarm console through the fire pump

controllers:
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• Pump running;

• Pump failure to start;

• Pump controller in off or manual position;

• Battery failure;

• Diesel overspeed;

• Diesel high engine-jacket coolant temperature;

• Diesel low oil pressure; and

• Battery charger failure.

Both pumps are maintained in the automatic-start position at all times, except during periods of

maintenance and testing. Remote manual-start switches are provided in the Control Room

adjacent to the alarm console. Fire pumps can only be shut off at the controllers.

Portable Fire Extinguishers

Portable fire extinguishers are installed throughout all buildings, in accordance with NFPA 10

(NFPA, 1994). Multipurpose fire extinguishers are provided generally for Class A (ordinary

combustibles); Class B (flammable and combustible liquids); and Class C (electrical equipment)

fires. Specialized extinguishers are located in areas requiring protection of particular hazards. 

Wheeled extinguishers are provided for use in water-exclusion areas. In areas with moderatorcontrol issues, the extinguishers are filled with carbon dioxide or dry chemical.

Fire-Water Supply

The facility fire-water supply consists of two 946,000-L (250,000-gal.) water storage tanks

designed and constructed in accordance with NFPA 22 (NFPA, 1996c). Within each tank,

473,000-L (125,000-gal.) are reserved for fire protection. Fill and makeup to the tanks are from

the city water supply to the site, and the water supply is capable of filling either tank within an 8-

hour period. The fire pumps consist of two diesel-driven 3785-Lpm (1000-gpm), at 10.35-bar

(150-psia), pumps. Both pumps are horizontal centrifugal pumps designed and installed in

accordance with NFPA 20 (NFPA, 1996b). The maximum anticipated fire demand is 3785 Lpm

(1000 gpm), based on 1892 Lpm (500 gpm) from a building-sprinkler system, plus 3785 Lpm

(500 gpm) for hose streams, for a duration of 2 hours. The combination of two water tanks and

two fire pumps provides 100 percent redundancy for fire protection. 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 38-mm (11⁄2-in.)

diameter hoses and one 64-mm (2 1⁄2-in.)-diameter backup hose, consistent with NFPA 1410

(NFPA, 2000). These lines will have a minimum nozzle pressure of 4.5 bar (65 psia) for the 38-

mm (1 1⁄2-in.) hose, and 6.9 bar (100 psia) for the 64-mm (2 1⁄2-in.) hose.

Engineered Automatic Fire-Suppression Systems

Wet pipe sprinkler systems are provided in the following buildings:

• Administration Building;

• Central Utilities Building; and

• Fire Pump House
7-14

These systems are designed and tested in accordance with NFPA 13 (NFPA, 1996a).

Sprinkler-system control valves are monitored under a periodic inspection program, and their

proper positioning is supervised in accordance with NFPA 801 (NFPA, 2003).

Fire-rated enclosures are provided for several chemical traps on the second floor of the Process

Services Area, in each Separations Building Module. These enclosures will be protected with a

gaseous-suppression system. The type of system used will be determined in the final design

and will either be a carbon dioxide system designed in accordance with NFPA 12 (NFPA, 1993)

or a clean-agent system designed in accordance with NFPA 2001 (NFPA, 1996f).

The NRC staff concludes that the applicant has reasonably determined the required fireprotection features for preventing or mitigating fire-accident scenarios that could lead to

unacceptable performance, in accordance with the requirements in 10 CFR 70.61.

The applicant’s ISA Summary has adequately addressed active engineered fire protection

systems, in accordance with the guidance established in NUREG-1520 (NRC, 2002). In

addition, the applicant’s active-engineered fire protection systems meet the requirements of 10

CFR 30.33, 10 CFR 40.32, 10 CFR 70.22, 10 CFR 70.64, and 10 CFR 70.65 as they pertain to

the fire protection aspects of the facility.

7.3.4.4 Fire Safety and Emergency Response

The facility will maintain a fire brigade made up of employees trained in fire fighting techniques,

first aid procedures, and emergency response. The fire brigade is organized, operated, trained

and equipped in accordance with NFPA 600 (NFPA, 1996e) for incipient fire fighting capability. 

The intent of the facility fire brigade is to be able to handle all minor fires and to provide a firstresponse effort designed to supplement the local fire department for major fires at the plant. 

The plant fire brigade, working with the plant’s emergency operations center, will coordinate offsite fire department activities to ensure moderator control and criticality safety. The fire brigade

is staffed so that there is a minimum of five brigade members available per shift.

Periodic training is provided to off-site assistance organization personnel in the facility

emergency training procedures. Facility emergency response personnel meet at least annually

with each off-site assistance group, to accomplish training and review items of mutual interest,

including relevant changes to the program. A Memoranda of Understanding between Louisiana

Energy Services and the City of Eunice, New Mexico Fire and Rescue Agency, and the City of

Hobbs, New Mexico, Fire Department defines the fire-protection and emergency-response

commitments between the organizations. The Eunice Fire and Rescue Agency is the initial

response agency and can respond between 11 and 15 minutes. The department has three

pumpers, three grass trucks, one tanker, and a roster of approximately 20 volunteers. The

Hobbs Fire Department can respond in about 25 to 30 minutes. The Hobbs Fire Department

has five pumpers, one ladder truck, three grass trucks, one tanker, and a roster of about 70 paid

personnel.

The NRC staff concludes with reasonable assurance that the on-site fire brigade, on-site water

supply, on-site hose lines, and mutual aid from adequately equipped fire departments can

provide defense-in-depth protection from releases from all identified credible fire scenarios and

satisfy the requirements of 10 CFR 70.64(b), and are in accordance with the guidance in

NUREG-1520 (NRC, 2002). In addition, the applicant’s emergency response capability meet
7-15

the requirements of 10 CFR 30.33, 10 CFR 40.32, 10 CFR 70.22, 10 CFR 70.64, and 10 CFR

70.65 as they pertain to the fire protection aspects of the facility.

 

7.4 EVALUATION FINDINGS

The dominant fire risk to safety and health of workers and the public for the proposed process is

a fire that could lead to loss of confinement of UF6. This includes a fire damaging the centrifuge

machines and piping that provide UF6 confinement, or UF6 cylinders inside or on the outdoor

storage pad. The applicant’s submittals provide sufficient information, in accordance with

requirements of 10 CFR 30.33, 10 CFR 40.32, 10 CFR 70.22, and 10 CFR 70.65, regarding

potential fire hazards, consequences, and required controls for the proposed processes. The

NRC staff determined that the applicant demonstrated compliance with the performance

requirements of 10 CFR 70.61 for fire protection related to postulated accident scenarios. The

applicant has identified a reasonable set of IROFS and defense-in-depth protection to ensure

acceptable risks within the performance requirements of 10 CFR 70.61. 

Based on the design of the facility, relative to fire protection and the designation of IROFS and

measures that provide defense-in-depth, the staff concludes with reasonable assurance that the

facility also meets the requirements of 10 CFR 30.33, 10 CFR 40.32, 10 CFR 70.22, and 10

CFR 70.64 (a)(3), regarding baseline design criteria for protection against fires and explosions,

and 10 CFR 70.64 (b) defense-in-depth.

7.5 REFERENCES

(ANSI, 1981) American National Standards Institute (ANSI). ANSI C2, “National Electrical

Safety Code,” 1981.

(NFPA, 1993) National Fire Protection Association (NFPA). NFPA 12, “Standard on Carbon

Dioxide Extinguishing Systems,” 1993.

(NFPA, 1994) National Fire Protection Association (NFPA). NFPA 10, “Standard for Portable

Fire Extinguishers,” 1994.

(NFPA, 1996a) National Fire Protection Association (NFPA). NFPA 13, “Standard for the

Installation of Sprinkler Systems,” 1996.

(NFPA, 1996b) National Fire Protection Association (NFPA). NFPA 20, “Standard for

Installation of Stationary Pumps for Fire Protection,” 1996.

(NFPA, 1996c) National Fire Protection Association (NFPA). NFPA 22, “Standard for Water

Tanks for Private Fire Protection,” 1996.

(NFPA, 1996d) National Fire Protection Association (NFPA). NFPA 72, “National Fire Alarm

Code,” 1996.
7-16

(NFPA, 1996e) National Fire Protection Association (NFPA). NFPA 600, “Standard on

Industrial Fire Brigades,” 1996.

(NFPA, 1996f) National Fire Protection Association (NFPA). NFPA 2001, “Standard on Clean

Agent Fire Extinguishing Systems,” 1996.

(NFPA, 1999) National Fire Protection Association (NFPA). NFPA 220, “Standard on Types of

Building Construction,” 1999.

(NFPA, 2000) National Fire Protection Association (NFPA). NFPA 1410, “Standard on Training

for Initial Emergency Scene Operations,” 2000.

(NFPA, 2003) National Fire Protection Association (NFPA). NFPA 801, “Standard for Fire

Protection for Facilities Handling Radioactive Materials,” 2003.

(NFPA, 2004) National Fire Protection Association (NFPA). NFPA 70E, “Standard for Electrical

Safety in the Workplace,” 2004.

(NRC, 1994) U.S. Nuclear Regulatory Commission (NRC). NUREG-1491, “Safety Evaluation

Report for the Claiborne Enrichment Center, Homer, Louisiana,” January 1994.

(NRC, 2002) U.S. Nuclear Regulatory Commission (NRC). NUREG-1520, “Standard Review

Plan for the Review of a License Application for a Fuel Cycle Facility,” March 2002.

(NRC, 2004) Johnson, T.C., U.S. Nuclear Regulatory Commission (NRC). March 9-10, 2004,

“Meeting Summary: Louisiana Energy Services’ Integrated Safety Analysis In-Office Review.” 

March 25, 2004.

(NRC, 2005) Wescott, R.G., U.S. Nuclear Regulatory Commission (NRC). “Note to File:

Confirmatory Calculations for Fire Protection Review of National Enrichment Facility Integrated

Safety Analysis Summary.” March 22, 2005.
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8.0 EMERGENCY MANAGEMENT

The purpose of this review of the applicant’s emergency management plan is to determine if the

applicant has established, before the start of operations, adequate emergency management

facilities and procedures to protect workers, the public, and the environment.

Emergency capability is incorporated into the baseline design criteria of 10 CFR Part 70, and is

intended to ensure control of licensed material, evacuation of personnel, and availability of

emergency facilities.

8.1 REGULATORY REQUIREMENTS

10 CFR 30.32(i)(1), 10 CFR 40.31(j)(1), and 10 CFR 70.22(i)(1)(i) specifies when an applicant is

not required to submit an Emergency Plan (EP) to the U.S. Nuclear Regulatory Commission

(NRC). If an applicant is required to submit an EP, as described in 10 CFR 30.32(i)(1)(ii), 10

CFR 40.31(j)(1)(ii), and 10 CFR 70.22(i)(1)(ii), then 10 CFR 30.32(i)(3), 10 CFR 40.31(j)(3), and

10 CFR 70.22(i)(3) contains the information that must be included in the EP. In addition, 10

CFR 70.64(a)(6) requires applicants to address the control of licensed material, evacuation of

personnel, and availability of emergency facilities for the design of new facilities.

8.2 REGULATORY ACCEPTANCE CRITERIA

The acceptance criteria for NRC’s review of the applicant’s emergency management program

are outlined in Section 8.4.3.1 of “Standard Review Plan for the Review of a License Application

for a Fuel Cycle Facility,” NUREG-1520 (NRC, 2002).

8.3 STAFF REVIEW AND ANALYSIS

8.3.1 Facility Description

Section 1.0 of the EP (LES, 2005a) contains descriptions of the licensed activity, the facility and

site, and the area near the site. The information provided includes:

• A description of the enrichment process;

• A discussion of chemicals of concern, including form, physical state, location, and

quantity [uranium hexafluoride (UF6) has been identified as the only direct chemical of

concern to be used at the facility];

• 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 Gaseous Effluent Vent Systems, including stack heights, maximum

flow rates, and filter efficiencies;

• A description of the area near the site, including airport, railroad, pipeline, topography,

and area land-use information;

• Demography of the area, water use, and climate; and
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• Detailed maps of the facility and surrounding area.

This fulfills the requirements of 10 CFR 30.32(i)(3)(i), 10 CFR 40.31(j)(3)(i), and 10 CFR

70.22(i)(3)(i).

The applicant will maintain compliance with the Emergency Planning and Community Right-toKnow Act of 1986, in accordance with 10 CFR 30.32(i)(3)(xiii), 10 CFR 40.31(j)(3)(xiii), and 10

CFR 70.22(i)(3)(xiii). This is accomplished by conducting annual inventories; compiling the

inventory information, and providing it to the appropriate agencies; and advising or training

construction and operating personnel regarding the marking, storage, and location of all

hazardous chemicals, including reporting accidental releases of these substances.

8.3.2 Onsite and Offsite Emergency Facilities

Onsite emergency facilities are discussed in Sections 6.1 and 6.3 of the EP (LES, 2005a). The

Control Room will be the primary Emergency Operations Center (EOC) and will have current asbuilt drawings, procedures, and operational engineering information. The designated alternate

EOC is the Security Building, which will be used depending on the nature and location of the

emergency situation, or if the Control Room becomes uninhabitable. The Security Building will

have the same documentation that is available to emergency responders, as is available in the

primary EOC. An offsite location will be available, if necessary, and is described in the

Safeguards Contingency Plan.

Offsite emergency support and equipment is discussed in Section 4.3, including fire, emergency

medical services, and local law enforcement. Support is available from a number of offsite

agencies, if necessary, as discussed in Section 4.3.

Section 6.4 of the EP (LES, 2005a) describes the emergency monitoring equipment that is

available for personnel and area monitoring, which includes criticality accident alarms,

personnel monitoring equipment, liquid effluent monitors, air monitors, hydrogen fluoride

monitors, and a meteorological measurement system for wind speed, direction, and

temperature.

Emergency equipment will be inventoried and tested on a quarterly basis, as discussed in

Section 7.6 of the EP (LES, 2005a). Deficiencies identified will be reported to the Emergency

Preparedness Manager, who will ensure timely corrective action is taken.

Section 6.2 of the EP (LES, 2005a) describes the communication systems that will be used at

the facility, which includes the facility telephone system with facsimile capability, the public

address system, alarms, and two-way radios that are powered by diesel-backed alternating

current sources. The bandwidths of the radio systems will allow communication with the

hospital, ambulance services, fire department, and state, county, and local law enforcement

agencies and the New Mexico Department of Public Safety, Office of Emergency Management. 

The telephone system will be the primary means of offsite emergency communications.
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8.3.3 Types of Accidents
8-4

In Section 2.1 of the EP (LES, 2005a), both postulated high- and intermediate-consequence

events are identified. Accident sequences, as well as mitigating and preventive measures, are

described. Nuclear criticality and loss of containment leading to a very large release of UF6 are

the only postulated events identified for which protective actions may be necessary. A

consequence analysis has been performed for a nuclear criticality scenario and for the Blending

Donor Station Heater Controller Failure/Heater Run-Away scenario. The results of these

analyses are provided in Section 3.7.3 of the “Integrated Safety Analysis Summary” (LES,

2005b). This meets the requirements of 10 CFR 30.32(i)(3)(ii), 10 CFR 40.31(j)(3)(ii), and 10

CFR 70.22(i)(3)(ii).

8.3.4 Classification of Accidents

Section 3.1 of the EP (LES, 2005a) explains the system used to classify an emergency as either

an Alert or a Site-Area Emergency, and defines both types of incidents. The applicant has

established that the threshold for escalating an event from an Alert to a Site-Area Emergency is

based on indications of a release (e.g., indications such as the presence of a white vapor cloud

issuing from buildings that house UF6) that could require a response by an offsite response

organization, to protect persons offsite from reaching the offsite exposure limits set forth in 10

CFR 30.32(i)(1)(i), 10 CFR 40.31(j)(1)(i), and 10 CFR 70.22(i)(1)(i). This system for classifying

events is acceptable to the staff and meets the requirements of 10 CFR 30.32(i)(3)(iii), 10 CFR

40.31(j)(3)(iii), and 10 CFR 70.22(i)(3)(iii). The processes for making the appropriate

classification are provided in the Emergency Plan Implementing Procedures (EPIPs).

Table 3.1-1 in the EP (LES, 2005a) provides examples of site-specific incidents and the

emergency classification that will be declared for each event. The Shift Manager is responsible

for accident classification and assumes the responsibility of the Emergency Director until

relieved by the Plant Manager or his designee.

8.3.5 Detection of Accidents

Section 2.2 of the EP explains the methods and systems available to detect accidents at the

facility, including:

• Automatic seismic trip of ventilation systems and associated alarms;

• Criticality accident alarm system;

• Facility-wide fire detection system;

• Visual detection of uranyl fluoride (UO2F2) and odor of hydrogen fluoride (HF);

• HF monitors;

• Radiation monitors; or

• Other alarms that detect abnormal process conditions, including temperature, pressure,

weight, etc.

Actions in response to accidents will be outlined in EPIPs and directed by the Shift Manager. 

The methods and systems presented in Section 2.2 of the EP (LES, 2005a) fulfill the

requirements of 10 CFR 30.32(i)(3)(iv), 10 CFR 40.31(j)(3)(iv), and 10 CFR 70.22(i)(3)(iv).
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8.3.6 Mitigation of Consequences

Section 5.3 of the EP (LES, 2005a) describes actions and equipment that will be used to

mitigate the consequences of accidents at the facility. The applicant states that since the facility

will operate with only natural and low-enriched uranium in the form of UF6, there will be no

radiological hazards that could likely result in significant offsite radiation doses. The major

hazard would be the chemical hazard caused by a release of UF6.

The main features used at the facility to mitigate the consequences of accidents include

automatic interruption or termination of specific operations, fire detection and suppression

systems, operator response to abnormal conditions/alarms, and shutdown of the ventilation

system, in case of a UF6 release or a criticality. The actions, features, and means for

maintaining them, which are adequately described in Sections 5.3 and 7.6 of the EP (LES,

2005a), fulfill the requirements of 10 CFR 30.32(i)(3)(v), 10 CFR 40.31(j)(3)(v), and 10 CFR

70.22(i)(3)(v).

8.3.7 Assessment of Releases

Section 5.2 of the EP (LES, 2005a) explains the actions that will be taken to assess the extent

of an accident at the facility. In case of an Alert, dose projections of offsite radiation and

hazardous material exposures will be made and provided to offsite response agencies. In

addition, during a Site-Area Emergency, radiation or chemical surveys of the Assembly Area(s),

the EOC, and the facility will be performed. Environmental air sampling will be performed offsite, if necessary. During a UF6 release, water, air, and soil samples will be obtained in the

release path, and urine and fecal samples will be obtained from any workers exposed to the

release. Projections of offsite radiation and chemical exposures will be based on the estimated

amount of UF6 released, the point of release, and the meteorological conditions at the time of

the release, and will be performed using the Radiological Assessment System for Consequence

Analysis (RASCAL) software (NRC, 2001). These actions meet the requirements of 10 CFR

30.32(i)(3)(vi), 10 CFR 40.31(j)(3)(vi), and 10 CFR 70.22(i)(3)(vi).

8.3.8 Responsibilities

The responsibilities of facility personnel during normal operations and during emergency

situations are described in Sections 4.1 and 4.2, respectively, of the EP (LES, 2005a). In case

of an emergency, the Shift Manager assumes the duty of the Emergency Director until the Plant

Manager or designee arrives. Responsibilities of the Emergency Director include:

• Coordinating the response;

• Deciding to provide protective action recommendations to authorities responsible for

offsite emergency measures (shall not be delegated);

• Coordinating the staff and offsite personnel who augment the staff;

• Approving information before release to the press;

• Authority to request support from offsite agencies;

• Authority to delegate responsibilities;

• Directing prompt notification of local and state emergency organizations;

• Directing assessment of onsite and offsite actual or potential consequences;

• Implementing protective actions for onsite personnel; and

• Downgrading/terminating the emergency.
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Section 4.2.2 of the EP (LES, 2005a) summarizes the responsibilities of the remaining onsite

staff. This includes the Community Relations Coordinator, who will have a direct line of

communication to the Emergency Director, to ensure current and factual news releases. Other

information concerning public and media access to information is included in Section 5.8 of the

EP (LES, 2005a). Memoranda of Understanding have been established with local agencies and

will be reviewed annually and renewed at least every 4 years, or more frequently, if necessary.

Responsibilities for developing, maintaining, and updating the EP are discussed in Section

8.3.14 of this SER. The description of the responsibilities in the EP fulfills the requirements of

10 CFR 30.32(i)(3)(vii), 10 CFR 40.31(j)(3)(vii), and 10 CFR 70.22(i)(3)(vii).

8.3.9 Notification and Coordination

As discussed in Section 8.3.4 of this SER, classification of emergencies is outlined in Section

3.1 of the EP (LES, 2005a) and is the responsibility of the Shift Manager.

Section 3.2 of the EP (LES, 2005a) provides a clear commitment to promptly notify offsite

response organizations of an emergency, including notification of the NRC Operations Center

immediately after calling the offsite response organizations, but no later than 1 hour after

declaring an emergency. Sections 4.3 and 4.4 of the EP (LES, 2005a) provide an adequate

description of provisions for assistance from offsite response organizations. Sections 5.6 and

5.7 of the EP (LES, 2005a) provide an adequate description of provisions for medical treatment

of contaminated workers. These sections adequately describe the agreements held between

LES and local off-site organizations and agencies, procedures for these response organizations

access to the site, and equipment/services available from these organizations.

Section 5.2.4 of the EP (LES, 2005a) describes how projections of offsite radiation and

chemical exposures are made, and how this information is disseminated to appropriate

personnel, including State authorities and NRC. Section 5.4.2 of the EP (LES, 2005a)

discusses the use of offsite field monitoring data which will be compared with source-term data

and included in the protective-action recommendation process.

As discussed previously, the Community Relations Manager will have a direct line of

communication to the Emergency Director, to ensure current and factual news releases. Also,

mutual aid agreements are in place to provide additional support if other services or equipment

are not available.

As discussed previously, it is the responsibility of the Emergency Director (initially the Shift

Manager) to: (1) declare an alert or site-area emergency; (2) activate the onsite emergency

response organization; (3) notify offsite response authorities of an emergency and

recommended initial protective actions; (4) notify the NRC Operations Center; (5) decide what

onsite protective actions to initiate; (6) decide what offsite protective actions to recommend; (7)

decide to request support from offsite organizations; and (8) decide to terminate the emergency

or enter recovery mode.

This meets the requirements of 10 CFR 30.32(i)(3)(viii), 10 CFR 40.31(j)(3)(viii), and 10 CFR

70.22(i)(3)(viii).
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8.3.10 Information To Be Communicated

Section 3.3 of the EP (LES, 2005a) provides an adequate description of the type of information

to be given to offsite response organizations during an emergency. The applicant will use Table

3.3-1, “National Enrichment Facility Notification Form,” as a script to be used for initial

notification of an emergency at the facility, to appropriate offsite facilities. Information included

in Table 3.3-1 is contact information for state and local authorities; what type of event has

occurred (Alert or Site-Area Emergency); a description of the event, time and date of the event;

radiological conditions; meteorological data; and protective-action recommendations, if any. 

In the event of a Site Area Emergency, a standard recommendation will be provided to offsite

assistance organizations with more specific data, if available, as discussed in Section 3.3 of the

EP (LES, 2005a).

This meets the requirements of 10 CFR 30.32(i)(3)(ix), 10 CFR 40.31(j)(3)(ix), and 10 CFR

70.22(i)(3)(ix).

8.3.11 Training

Section 7.2 of the EP (LES, 2005a) provides a description of the training the licensee will

provide to workers on how to respond to an emergency. All workers receive general employee

training, which includes Quality Assurance; radiation protection (including the use of dosimetry

and protective clothing); and safety, emergency, and administrative procedures. Training on

criticality safety, radiation protection, and emergency procedures specific to each type of job

function is also provided under the nuclear safety training program. 

Emergency response personnel receive additional training annually, to provide specific

information on how the emergency organization responds during emergency conditions,

including staffing; determining and estimating potential offsite releases of radiation and

chemicals; and interface with offsite assistance organizations. This training is required before

being assigned to the emergency organization, and refresher training is provided at least once

every year.

The nuclear safety training program includes: (a) information on instructions to workers; (b) as

low as is reasonably achievable (ALARA) methods of controlling radiation exposures; (c)

contamination control methods; (d) use of monitoring equipment; (e) emergency procedures and

actions; (f) nature and sources of radiation; (g) safe use of chemicals; (h) biological effects of

radiation; (i) use of personnel monitoring devices; (j) principles of nuclear criticality safety; (k)

risk to pregnant females; (l) radiation-protection practices; (m) protective clothing; (n) respiratory

protection; and (o) personnel surveys. Specific topics, performance objectives, content, training

schedules, and number of training hours required for each position are contained in

administrative procedures.

The Emergency Preparedness Manager is responsible for emergency preparedness training. 

Individuals requiring unescorted access to the Controlled Access Area receive annual

retraining. The Health, Safety, and Engineering Manager, or designee, reviews and updates the

content of the formal nuclear safety training programs, as required, at least every 2 years to

ensure that the programs are current and adequate.
8-8

Facility tours and classroom training are also provided to offsite response organizations. The

training includes: (a) information concerning facility access control; (b) potential accident

scenarios; (c) emergency action levels; (d) notification procedures; (e) exposure guidelines; (f)

personnel-monitoring devices; (g) communications; (h) contamination control; and (i) the role of

the offsite assistance organization in responding to an emergency at the facility. Each group

will meet at least annually, with facility personnel, to accomplish this training and review items of

mutual interest, including relevant changes to the program.

This fulfills the requirements of 10 CFR 30.32(i)(3)(x), 10 CFR 40.31(j)(3)(x), and 10 CFR

70.22(i)(3)(x).

8.3.12 Safe Shutdown (Recovery and Facility Restoration)

Section 9.1 of the EP (LES, 2005a) states that during an emergency, immediate actions will be

directed toward limiting the consequences of the incident to afford maximum protection to facility

personnel and the general public. Once control of the facility has been reestablished, a

systematic and planned approach to full facility recovery will be taken. Criteria to be considered

before terminating an emergency and entering the recovery phase include:

• Conditions requiring emergency classification no longer exist;

• Radioactive and/or hazardous chemical releases are under control;

• Environmental assessment activities in progress are those necessary to determine the

extent of environmental impact from the emergency, not the active tracking of material

still being released;

• Facility radiation, contamination, and hazardous chemical levels are stable or

decreasing and acceptable, given current facility conditions;

• Any fire, flood, earthquake or similar initiating emergency event no longer exists;

• All required notifications have been made;

• At an Alert or Site-Area Emergency, the facility Operations Center is staffed and

activated; and

• Offsite conditions do not unreasonably limit access of outside support to the facility.

Sections 5.2, 5.3, and 5.4 of the EP (LES, 2005a) describe the methods to be used for

assessing the extent of the event and the status of the facility, and the mitigative actions

necessary to reduce or stop any ongoing releases of radioactive material or hazardous

chemicals.

Section 9.2 of the EP (LES, 2005a) contains information regarding the staffing of the facility

during the recovery phase of an event. It addresses minor events requiring only the normal onshift personnel, as well as more significant events, which would require the Emergency

Organization (EO), or portions of the EO, to perform recovery tasks. This fulfills the

requirements of 10 CFR 30.32(i)(3)(xi), 10 CFR 40.31(j)(3)(xi), and 10 CFR 70.22(i)(3)(xi).

8.3.13 Exercises and Drills

Section 7.3 of the EP (LES, 2005a) provides adequate provisions for drills, exercises, and

biennial exercises that are used to test the adequacy of EPIPs, emergency equipment,

instrumentation, and to ensure all emergency response personnel are familiar and proficient
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with their duties. The following guidelines are used to control and evaluate drill and exercise

conduct:

• Drill and exercise performance is assessed against specific scenario objectives, using

postulated accidents that adequately test personnel, equipment, and resources,

including any previously identified weaknesses;

• Participant, controller, evaluator, and observer pre-drill/exercise briefings are conducted;

• Controllers who maintain the timeline but do not interfere with the EO’s response, except

where safety considerations are concerned, provide scenario data and exercise

messages;

• Trained evaluators and independent observers are used to identify and record

participant performance, scenario strengths and deficiencies, equipment problems, and

to provide recommendations for improvement;

• The pre-staging of equipment and personnel is minimized to realistically test the

activation and staffing of emergency facilities; and

• Post-drill and post-exercise critiques will be conducted by those involved, and

appropriate improvements will be implemented.

Each exercise is critiqued according to Section 7.4 of the EP (LES, 2005a). Areas evaluated

include the adequacy of the EP, procedures, equipment, facilities, personnel training, and

overall response effectiveness. The Emergency Preparedness Manager is responsible for

tracking deficiencies and ensuring that corrective actions are implemented and effective.

The drill and exercise program is established to demonstrate:

• Effectively use available resources to control the site, and to obtain additional resources

when necessary;

• Mitigate further damage to the facility;

• Control any radiological or hazardous material releases;

• Perform required onsite activities under simulated elevated radiation/hazardous material

airborne conditions;

• Accurately assess the facility’s status during emergency conditions;

• Initiate recovery;

• Demonstrate personnel protection measures both onsite and offsite;

• Control and minimize hazards to individuals during fires, medical emergencies,

mitigation activities, and search and rescue operations;

• Effectively use communication systems to notify offsite agencies and to support

emergency response activities; and

• Disseminate accurate, reliable, timely, and understandable information.

Offsite organizations are invited to participate in the biennial exercise and NRC is invited to

participate or observe. Exercise objectives and scenarios will be submitted to NRC, for review

and comment, at least 60 days before the exercise. 

Section 7.3.2 of the EP (LES, 2005a) includes an adequate provision for quarterly

communications checks, to verify the operability of initial notification points. The Emergency

Preparedness Manager verifies and updates emergency telephone numbers contained in the

facilities EPIPs. 
8-10

The information provided in Sections 7.3 and 7.4 of the EP (LES, 2005a) meets the

requirements of 10 CFR 30.32(i)(3)(xii), 10 CFR 40.31(j)(3)(xii), and 10 CFR 70.22(i)(3)(xii).

8.3.14 Responsibilities for Developing and Maintaining the Emergency Program and Its

Procedures Current

Section 7.1 of the EP (LES, 2005a) explains there will be site procedures for maintaining the EP

and procedures current. As outlined in Section 7.5 of the EP (LES, 2005a), an annual audit of

the EPIPs will be conducted by an independent group, to ensure the continued effectiveness of

the EP. Administrative procedures are established which assign responsibility for the

development, review, approval, and update of the Emergency Plan and its supporting

procedures.

Any proposed change in the EP that affects an offsite organization will be provided to that

organization for review and comment at least 60 days before the change being implemented,

unless mutually agreed otherwise. The applicant may incorporate changes to the EP without

receiving prior NRC approval, provided those changes do not decrease the effectiveness of the

EP, and NRC and affected offsite response organizations receive copies of the revised EP or

procedures within 6 months, in accordance with 10 CFR 30.34(f), 10 CFR 40.35(f), and 10 CFR

70.32(i). 

8.4 EVALUATION FINDINGS

The staff has evaluated the EP for the facility. In accordance with 10 CFR 30.32(i), 10 CFR

40.31(j), and 10 CFR 70.22(i), the licensee has established an EP for responding to the

radiological hazards resulting from a release of radioactive material or hazardous chemicals

incident to the processing of licensed material. The NRC staff reviewed the EP with respect to

10 CFR 30.32(i), 10 CFR 40.31(j), and 10 CFR 70.22(i) and the acceptance criteria in Section

8.4.3 of NUREG-1520 (NRC, 2002). It determined that the applicant’s EP is adequate to

demonstrate compliance with 10 CFR 30.32(i), 10 CFR 40.31(j), and 10 CFR 70.22(i), in that: 

(1) the facility is properly configured to limit releases of radioactive materials in case of an

accident; (2) a capability exists for measuring and assessing the significance of accidental

releases of radioactive materials; (3) appropriate emergency equipment and procedures are

provided onsite, to protect workers against radiation and other chemical hazards that might be

encountered after an accident; (4) a system has been established to notify Federal, State, and

local Government agencies and to recommend appropriate protective actions to protect

members of the public; and (5) necessary recovery actions are established to return the facility

to a safe condition after an accident.

8.5 REFERENCES

(LES, 2005a) Louisiana Energy Services (LES). “National Enrichment Facility Emergency

Plan,” Revision 3, 2005.

(LES, 2005b) Louisiana Energy Services (LES). “National Enrichment Facility Integrated

Safety Analysis Summary,” Revision 4, 2005.
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(NRC, 1992) U.S. Nuclear Regulatory Commission (NRC). Regulatory Guide 3.67, “Standard

Format and Content for Emergency Plans for Fuel Cycle and Materials Facilities,” 1992.

(NRC, 2001) U.S. Nuclear Regulatory Commission (NRC). NUREG-1741, “RASCAL 3.0:

Description of Models and Methods,” 2001.

(NRC, 2002) U.S. Nuclear Regulatory Commission (NRC). NUREG-1520, “Standard Review

Plan for the Review of a License Application for a Fuel Cycle Facility,” 2002.
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9.0 ENVIRONMENTAL PROTECTION

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

environmental protection plan 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 10 CFR Parts 20, 30, 40, 51, and 70.

9.1 REGULATORY REQUIREMENTS

To be considered acceptable, the applicant must satisfy the following regulatory requirements

regarding environmental protection:

1. Part 20 specifies the effluent control and treatment measures necessary to meet the

dose limits and dose constraints for members of the public specified in Subparts B, D,

and F; the survey requirements of Subpart F; the waste disposal requirements of

Subpart K; the records requirements of Subpart L; and the reporting requirements of

Subpart M.

2. 10 CFR 30.33 specifies in part that an application for the possession and use of

byproduct material will be granted provided that, among other things, the applicant’s

proposed equipment and facilities are adequate to protect health and minimize danger to

life or property, and that the applicant is qualified by training and experience to use the

byproduct material for the purpose requested in such a manner as to protect health and

minimize danger to life and property.

3. 10 CFR 40.31(k) states that, “A license application for a uranium enrichment facility must

be accompanied by an Environmental Report required under subpart A of Part 51 of this

chapter.”

4. 10 CFR 40.32(e) states that, “In the case of an application for a license for a uranium

enrichment facility, or for a license to possess and use source and byproduct material for

uranium milling, production of uranium hexafluoride, or for the conduct of any other

activity which the Commission determines will significantly affect the quality of the

environment, the Director of Nuclear Material Safety and Safeguards or his designee,

before commencement of construction of the plant or facility in which the activity will be

conducted, on the basis of information filed and evaluations made pursuant to subpart A

of part 51 of this chapter, has concluded, after weighing the environmental, economic,

technical and other benefits against environmental costs and considering available

alternatives, that the action called for is the issuance of the proposed license, with any

appropriate conditions to protect environmental values. Commencement of construction

prior to this conclusion is grounds for denial of a license to possess and use source and

byproduct material in the plant or facility. As used in this paragraph, the term

‘commencement of construction’ means any clearing of land, excavation, or other

substantial action that would adversely affect the environment of a site. The term does

not mean site exploration, roads necessary for site exploration, borings to determine

foundation conditions, or other preconstruction monitoring or testing to establish
9-2

background information related to the suitability of the site or the protection of

environmental values.”

5. Part 51 specifies that the applicant must submit an environmental report for construction

and operation of a uranium enrichment facility, as required by 10 CFR 51.60(b)(1)(vii).

6. 10 CFR 70.22(a)(7) specifies that the applicant must provide a description of the

equipment and facilities which will be used by the applicant to protect health and

minimize danger to life or property.

7. 10 CFR 70.59 outlines the radiological effluent monitoring reporting requirements for a

Part 70 licensee.

8. 10 CFR 70.65(b) specifies 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) established by the applicant.

9.2 REGULATORY ACCEPTANCE CRITERIA

The acceptance criteria for NRC’s review of the applicant’s environmental protection program

are outlined in Section 9.4.3.2 of NUREG-1520 (NRC, 2002).

9.3 STAFF REVIEW AND ANALYSIS

9.3.1 Radiation Safety

9.3.1.1 As Low As is Reasonably Achievable (ALARA) Goals for Air and Liquid Effluent

Control

Air Effluent ALARA Goal

The applicant estimated the maximum individual committed effective dose equivalent (CEDE)

for air effluents during normal operations at the proposed facility. The applicant estimated that

the total air effluent releases would be less than 9 megabecquerel (MBq) [240 microcuries (:Ci)]

of uranium-234 (U-234), uranium-235 (U-235), and uranium-238 (U-238) isotopes – the principal

constituents of natural, depleted, and enriched uranium that would be processed at the facility. 

Uranium-236 (U-236) is not considered to be a principal constituent in that it contributes

significantly less than 1 percent to the total releases. Similar Urenco facilities in Europe

typically maintain quantities of radioactivity in air effluent less than 10 grams (g) [0.35 ounces

(oz)] (LES, 2005b). Even if all of the 10 g (0.35 oz) of uranium were conservatively assumed to

be composed of 6 percent enriched uranium, then air effluent from Urenco facilities is generally

limited to no more than 1.1 MBq (30 :Ci) of uranium isotopes. Therefore, the staff finds that a

value of 9 MBq (240 :Ci) uranium represents a reasonably conservative upper bound on the

annual quantity of air effluent from the facility.

As noted in Section 8.7 of the applicant’s ER, the CEDE to the maximally exposed member of

the public located at the south side of the controlled area boundary, resulting from the release to
9-3

the atmosphere of 9 MBq (240 :Ci) of uranium from gaseous release points, would be less than

0.17 microsieverts (:Sv) (0.017 mrem) (LES, 2005b), or 0.017 percent of the 1 mSv (100-mrem)

limit on dose to the public in Part 20. The 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 (e.g.,

between 1 and 2 percent).

NUREG-1520 provides that the applicant’s radiological monitoring will be acceptable if, among

other things, action levels of radionuclide concentrations are selected such that below those

concentrations, doses to the public will be ALARA and will be below the limits specified in 10

CFR Part 20, Subpart B. In Section 6.1.1 of the ER (LES, 2005b), the applicant committed to

develop a program of corrective actions that will be implemented based on established action

levels that are sufficiently low to permit implementation of corrective actions before regulatory

limits are exceeded. As proposed by the applicant, action levels are divided into three priorities:

1. The sample parameter lowest alarm point will be selected based on conditions of

service, for example, 3 times the normal background level;

2. The sample parameter exceeds any of the existing administrative limits; and

3. The sample parameter exceeds any regulatory limit.

The applicant will implement corrective actions to ensure that: (a) the cause for the action-level

exceedance can be identified and immediately corrected; (b) applicable regulatory agencies are

notified, if required; (c) lessons-learned are communicated to appropriate personnel; and (d)

applicable procedures are revised accordingly, if needed.

A reasonable initial ALARA goal for air effluents, described in NRC Regulatory Guides 8.34 and

8.37 (NRC, 1992; NRC, 1993) is 10-20 percent of the regulatory limit. The applicant will

implement corrective actions, under Priority 1, above, that will likely result in doses that are a

small fraction of the regulatory limits in Part 20 and 40 CFR Part 190. In addition, the calculated

dose to the maximally exposed member of the public is a small fraction of the ALARA goals

identified in the above-referenced Regulatory Guides. Therefore, staff finds this initial estimate

of air effluent quantities to be a reasonable ALARA goal for air effluent.

Liquid Effluent ALARA Goal

The applicant estimated the maximum individual CEDE for liquid effluents during normal

operations at the proposed facility. The applicant estimated maximum annual quantity of

radiological material in liquid effluent would be 14.4 MBq (390 :Ci) of uranium as discharge to

the Treated Effluent Evaporative Basin (TEEB) (LES, 2005b). There are no offsite releases to

any surface waters or publicly owned treatment works. Therefore, the release pathway the

applicant assumed is airborne resuspension of particulate matter from the bottom of the basin, if

waste water should evaporate. The applicant evaluated the dose for the 30th year of operations,

which represents a realistically conservative upper bound on the amount of uranium available

for resuspension and atmospheric dispersion.

As noted in Section 8.7 of the applicant’s ER, potential radiological impacts from operation of

the proposed facility would result from controlled releases of small quantities of uranium

hexafluoride (UF6) during normal operations. The CEDE to the maximally exposed member of
9-4

the public located at the south side of the controlled area boundary, resulting from the annual

release to the atmosphere of 2.1 MBq (56 :Ci) of uranium from the TEEB, would be less than

0.017 :Sv (0.0017 mrem) per year if the TEEB is dry only 10 percent of the year (LES, 2005b). 

If the TEEB is dry the entire year, the CEDE to the maximally exposed member of the public

located at the south side of the controlled area boundary would be less than 0.17 :Sv (0.017

mrem) per year. The estimated maximum public dose is also well below the 0.1 mSv (10-mrem)

ALARA constraint on liquid emissions described in 10 CFR 20.1101 (e.g., less than 2 percent).

In Section 6.1.1 of the ER (LES, 2005b), the applicant committed to develop a program of

corrective actions that will be implemented based on established action levels that are

sufficiently low to permit implementation of corrective actions before regulatory limits are

exceeded. As proposed by the applicant, action levels are divided into three priorities:

1. The sample parameter lowest alarm point will be selected based on conditions of

service, for example, 3 times the normal background level;

2. The sample parameter exceeds any of the existing administrative limits; and

3. The sample parameter exceeds any regulatory limit.

The applicant will implement corrective actions to ensure that: (a) the cause for the action-level

exceedance can be identified and immediately corrected; (b) applicable regulatory agencies are

notified, if required; (c) lessons-learned are communicated to appropriate personnel; and (d)

applicable procedures are revised accordingly, if needed.

A reasonable initial ALARA goal for liquid effluent described in NRC Regulatory Guide 8.34

(NRC, 1992) is 10-20 percent of the regulatory limit. The applicant committed to implement

corrective actions under Priority 1, above, that will likely result in doses that are a small fraction

of the regulatory limits in Part 20 and Part 190. In addition, the calculated dose to the maximally

exposed member of the public is a small fraction of the ALARA goals identified in the above

referenced Regulatory Guide. The applicant commits to implement corrective actions well

below this level. Therefore, staff finds this initial estimate of liquid effluent to be a reasonable

ALARA goal for liquid effluent.

9.3.1.2 Air Effluent Controls to Maintain Public Doses ALARA

Air effluent controls at the proposed facility include: (a) the Separations Building Gaseous

Effluent Vent System (GEVS); (b) the Technical Services Building (TSB) GEVS; (c) the TSB

Heating, Ventilating, and Air Conditioning (HVAC) system that services potentially contaminated

areas; and (d) the Centrifuge Test and Post-Mortem Facilities Exhaust Filtration System.

GEVS

The function of each GEVS is to remove particulate matter containing uranium and hydrogen

fluoride (HF) from potentially contaminated gas streams. Each GEVS includes ducts; prefilters;

high-efficiency particulate air (HEPA) filters; potassium carbonate-impregnated activated carbon

filters; fans; monitors and controls; inlet and outlet isolation dampers; and a discharge stack. As

described below, one GEVS unit includes electrostatic filters to remove oil vapor from gaseous

effluent.
9-5

The TSB GEVS contains one filter station and one fan that accommodates 100 percent of the

effluent. The filter train includes a 85-percent efficient prefilter; a 99.97-percent-efficient HEPA

filter; and a 99-percent-efficient activated charcoal filter for removal of HF. Cleaned air effluent

is discharged through a roof-top vent stack.

The Separation Building GEVS contains two 100-percent-capacity filter/fan trains, with one train

in a standby configuration. Like the TSB GEVS, each Separation Building GEVS filter train is

composed of an 85-percent-efficient prefilter and a 99.97-percent-efficient HEPA filter. The

activated charcoal filter for removal of HF is rated to 99-percent efficiency. The exhaust of the

vacuum pump/trap sets is also routed through an electrostatic filter with a removal efficiency of

97 percent. The operational and standby filter/fan trains share the same roof-top vent stack.

TSB HVAC - Potentially Contaminated Areas

A portion of the TSB HVAC services potentially contaminated rooms in the TSB. These rooms

are the Decontamination Workshop, Cylinder Preparation Room, and Ventilated Room. The

HVAC system for these rooms consists of two 50-percent-capacity Air-Handling Units. Airflow

from the potentially contaminated rooms is exhausted through two 50-percent-capacity BagIn/Bag-Out HEPA filters by one of two 100-percent-capacity filtration exhaust fans. The exhaust

air is then discharged to the exhaust stack and monitored for alpha radiation and HF. Some

potentially contaminated areas have fume hoods that are connected to the TSB GEVS.

Centrifuge Test and Post-Mortem Facilities Exhaust Filtration System

The Centrifuge Test and Post-Mortem Facilities Exhaust Filtration System consists of a duct

network connected to one filter station and one of two 100-percent-capacity fans. One fan will

normally be in standby. The Centrifuge Test and Post-Mortem Facilities Exhaust Filtration

System filter train consists of: (a) an 85-percent-efficient prefilter; (b) a 99-percent-efficient

activated charcoal filter for removal of HF; and (c) a 99.97-percent-efficient HEPA filter. 

Cleaned air effluent is discharged through a roof-top vent stack on the Centrifuge Assembly

Building.

Staff Evaluation of Air Effluent Controls

The staff evaluated the air effluent controls, including the TSB and Separation Building GEVS,

the TSB HVAC systems, and the Centrifuge Test and Post-Mortem Facilities Exhaust Filtration

System, the placement of air effluent controls at the proposed facility, and the type of controls

selected for the hazards that are present in various processes. The applicant’s proposed use of

HEPA filters is standard industry practice for control of air effluents that potentially contain

airborne particle matter. Also, the use of activated charcoal for capture and control of HF is

well-established. The applicant will also prepare and maintain operational procedures that limit

activities with dispersible forms of uranium to areas with appropriate air effluent controls. In

some cases, these designated areas will also be fitted with local control devices (flexible hoses

attached to a GEVS, hoods, and glove boxes). Therefore, based on the above, the staff finds

that the applicant has demonstrated that its air effluent controls will reduce releases to assure

adequate protection of the environment and of health and safety of the public.

9.3.1.3 Liquid Effluent Controls to Maintain Public Doses ALARA
9-6

Liquid effluent controls at the proposed facility will be located in the Liquid Effluent Collection

and Treatment System in the TSB. The applicant has identified seven major sources of liquid

waste from processes in the TSB and Separations Building. The system is designed to use

combinations of in-tank precipitation, centrifuge-based oil/water separation, a filter press, an

evaporator/dryer, and mixed-bed demineralizers, to decontaminate liquid wastes before

evaporation of decontaminated aqueous wastes in the outdoor TEEB.

The proposed liquid effluent control systems will reduce the quantity of uranium entering the

TEEB by about two orders of magnitude, from 56.7 kilograms (kg) (125 lbs) to less than 0.6 kg

(1.3 lbs) per year. The uranium removed by the liquid effluent control systems inside the TSB

will be disposed as Class A low-level radioactive waste (LLW) at one or more licensed disposal

facilities. As described in the staff’s Safety Evaluation Report (SER) Section 9.3.1.1, the dose

to the maximally exposed offsite member of the public from the resuspension and atmospheric

dispersion of uranium-bearing dust from a dry TEEB is less than 0.17 :Sv (0.017 mrem) per

year. This dose is less than one percent of the regulatory limit. A reasonable initial ALARA

goal for liquid effluent described in NRC Regulatory Guide 8.34 (NRC, 1992) is 10-20 percent of

the regulatory limit. Therefore, based on the above, the staff finds that the applicant has

demonstrated that it will reduce releases to adequately assure protection of the environment

and of health and safety of the public.

9.3.1.4 ALARA Reviews and Reports to Management

In Section 4.2 of its Safety Analysis Report (SAR), the applicant describes an ALARA program

for the proposed facility. The ALARA program would include annual review of the content and

implementation of the radiation protection program, including the effluent control program. The

Radiation Protection Manager is responsible for implementing the ALARA program and

ensuring that adequate resources are committed to make the program effective. The Radiation

Protection Manager prepares an annual ALARA program evaluation report. The report reviews: 

(1) radiological exposure and effluent release data for trends; (2) audits and inspections; (3)

use, maintenance, and surveillance of equipment used for exposure and effluent control; and

(4) other issues, as appropriate, that may influence the effectiveness of the radiation protection

and ALARA programs. Copies of the report are submitted to the Plant Manager and the Safety

Review Committee. As described more fully in chapter 4 of this SER, the ALARA program is

acceptable to the staff.

9.3.1.5 Waste Minimization

In Section 4.13.4 of the ER (LES 2005b), the applicant described facility features and systems

that will minimize the generation of radioactive waste. These features and systems are based

on principles of control, conservation, reprocessing, and recovery. Specific examples include: 

(a) the Fomblin Oil Recovery System; (b) the Decontamination System; (c) a General

Decontamination process; (d) Sample Bottle Decontamination System; (e) Flexible Hose

Decontamination System; and (f) the Laundry System. These systems are consistent with the

guidance provided in NRC Information Notice 94-23, “Guidance to Hazardous, Radioactive, and

Mixed-Waste Generators on the Elements of a Waste Minimization Program” (NRC, 1994b) and

are, therefore, acceptable to the staff.

9.3.1.6 Safe Handling of Radioactive Wastes
9-7

The applicant has identified design features and procedures for safe handling of air and liquid

effluent and solid wastes from both construction and operation of the proposed facility. The

staff’s evaluation of air and liquid effluent is described in SER Sections 9.3.1.2 and 9.3.1.3.

The applicant has described a solid waste management program at the proposed facility for

industrial (nonhazardous), radioactive and mixed, and hazardous wastes. In addition, the

applicant will segregate solid radioactive and mixed waste according to the quantity of liquid that

is not readily separable from the solid material. The applicant does not propose to process

and/or treat solid waste.

The applicant proposes to dispose of all solid radioactive wastes as Class A LLW. Industrial

waste, including miscellaneous trash, vehicle air filters, empty cutting oil cans, miscellaneous

scrap metal, and paper will be shipped offsite for minimization and then sent to a permitted

waste landfill.

Radioactive waste will be collected in labeled containers in each Restricted Area and

transferred to the Radioactive Waste Storage Area for inspection. Suitable waste will be

volume-reduced and all radioactive waste disposed of at a licensed LLW disposal facility.

Hazardous wastes (e.g., spent blasting sand; empty spray-paint cans; empty propane-gas

cylinders; solvents such as acetone and toluene; degreaser solvents; diatomaceous earth;

hydrocarbon sludge; and chemicals such as methylene chloride and petroleum ether) and some

mixed wastes will be generated at the facility. These wastes will also be collected at the point of

generation, transferred to the Waste Storage Area, inspected, and classified. Any mixed waste

that may be processed to meet land disposal requirements may be treated in its original

collection container and shipped as LLW for disposal.

As noted in Section 3 of the applicant’s ER, the operation of the facility would yield an annual

production of 625 cylinders of depleted UF6 per year, or approximately 7800 metric tons (8,600

tons). The Uranium Byproduct Cylinder (UBC) Storage Pad would have a capacity of 15,727

cylinders. The NRC has evaluated the environmental impacts of alternatives for disposition of

depleted UF6 in its “Environmental Impact Statement for the Proposed National Enrichment

Facility in Lea County, New Mexico” (NRC, 2005). During temporary storage of this material on

the UBC Storage Pad, the applicant has committed to a number of mitigation measures to

minimize public and occupational health impacts. Among these measures is a cylinder

management program to monitor storage conditions on the UBC Storage Pad; to monitor

cylinder integrity by conducting routine inspections for breaches; and to perform cylinder

maintenance, as needed. The UBCs will be stored on saddles made of materials that will not

cause corrosion of the cylinders. The storage array of the UBCs will permit easy visual

inspection of the cylinders. The cylinders will be surveyed for external contamination before

being placed on the UBC Storage Pad and before being transported offsite. In addition, the

UBCs will be re-inspected annually for damage or surface coating defects. These inspections

are to insure that the cylinders are free from bulges; dents; gouges; cracks; or significant

corrosion, and that the cylinder plugs are undamaged and not leaking. If significant

deterioration is detected, then the contents of the cylinder will be transferred to another cylinder. 

This will be followed by a root cause assessment of any significant deterioration. 

On the basis of this analysis, the staff find that the applicant’s implementation of its program for

management of solid radiological and nonradiological wastes related to facility operation will 
9-8

reduce unnecessary exposures to adequately assure protection of the environment and of

health and safety of the public. 

9.3.2 Effluent and Environmental Monitoring

9.3.2.1 Air Effluent Monitoring

As described in SER section 9.3.1.1, the staff finds that the expected concentrations of

radioactive materials in airborne effluents would be well below the regulatory limits specified in

10 CFR 20.1302(c). The applicant has proposed to demonstrate compliance with air effluent

limits by calculation of the total effective dose equivalent to the individual who is likely to receive

the highest dose. Such a demonstration of regulatory compliance is in accordance with 10 CFR

20.1302(b)(1).

The staff reviewed the applicant’s assumptions and conclusions used in its calculations (ER

Sections 4.12 and 8.7) and determined that they are reasonable. The staff’s evaluation of

radiation exposures from normal operations is found in Chapter 4 (section 4.2.12.2) of the Final

Environmental Impact Statement (EIS).

The applicant has identified all airborne effluent discharge locations, and will monitor discharges

from potentially contaminated areas in accordance with NRC Regulatory Guide 4.16 (NRC,

1985). The applicant will perform continuous monitoring at the discharge stacks for alphaemitting radioactivity and HF. The applicant will also conduct periodic grab sampling of areas

that contain dispersible uranium, but which are not expected to have airborne contamination. 

Sampling of each contributing source will not be part of the effluent monitoring design. 

However, each potentially contaminated area will be monitored by health physics surveys for

airborne contamination, should such surveys be required for effective process and effluent

control.

A list of HVAC systems, GEVS, and exhaust filtration systems that discharge building ventilation

exhaust is provided in Table 9.3-1.

Weekly samples of stack air effluent will be measured for gross alpha radioactivity. Weekly

samples will be composited quarterly for isotope-specific analyses for U-234, U-235, U-236, and

U-238.

The applicant has proposed measurement sensitivities of 10-15 :Ci/mL for weekly gross alpha

measurements and 10-17 :Ci/mL for uranium isotopes. 

As noted in section 9.3.1.1 above, the applicant committed to developing a program of

corrective actions to be taken when established action levels of radiation are exceeded for any

of the measured parameters. As proposed by the applicant, action levels are divided into three

priorities:
9-9

Table 9.3-1

Building Ventilation Air Discharge Locations

Process Area No. of AirHandling Units

(Capacity, each)

Estimated

Exhaust Flow

Room

Pressurization

Controls S

Cascade Hall(Each of 3 modules)

None None Ambient None 

Process Service Corridor(Each of 3 modules)

8

(12.5%)

34,300 cfm Neutral None 

Link Corridor(Each of 3 modules)

2

(50%)

3,700 cfm Neutral None 

Above Cascade Hall(Each of 3 modules)

4

(100%)

123,000 cfm Neutral None 

UF6 Handling Area

(Each of 3 modules)

3

(33%)

90,600 cfm Neutral None 

Blending & LiquidSampling Workshop

2

(50%)

34,000 cfm Neutral None 

TSB HVACContaminated Areas(Decon. Workshop,Cylinder Prep. and Vent.Room)

2

(50%)

38,100 cfm Negative Roughing filter

Carbon filter

HEPA filter

Y

TSB HVACNon-contaminated Areas

2

(50%)

38,100 cfm Neutral None 

Vacuum Pump Rebuildand ME&I Workshops

(1 exhaust fan)

(100%)

38,100 cfm Neutral None 
Process Area No. of AirHandling Units

(Capacity, each)

Estimated

Exhaust Flow

Room

Pressurization

Controls S

9-10

Chem./Mass Spect.&Environmental MonitoringLab

1

(100%)

38,100 cfm Pos. (EM Lab)

Neg. (others)

None 

Centrifuge AssemblingBuilding Test and PostMortem Facility

2

(50%)

16,100 cfm Neutral / 

Neg.

Roughing filter

Carbon filter

HEPA filter

Y

Cylinder Receipt andDispatch Building

10

(10%)

288,000 cfm Neutral None 

TSB GEVS N/A 11,000 cfm Neg. Roughing filter

Carbon filter

HEPA filter

Y

Separation BuildingGEVS

N/A 6,500 cfm Neg. Roughing filter

Carbon filter

HEPA filter

Y
9-11

1. The sample parameter lowest alarm point will be selected based on conditions of

service, for example, 3 times the normal background level;

2. The sample parameter exceeds any of the existing administrative limits; and

3. The sample parameter exceeds any regulatory limit.

The program of corrective actions will be implemented to ensure that: (a) the cause for the

action-level exceedance can be identified and immediately corrected; (b) applicable regulatory

agencies are notified, if required; (c) lessons-learned are communicated to appropriate

personnel; and (d) applicable procedures are revised accordingly, if needed.

As described in Section 1.3 of the applicant’s SAR, in addition to meeting NRC requirements,

the applicant will also obtain required Federal and State permits for hazardous air pollutants, as

required.

As described in Section 6.1.2 of the applicant’s ER, the applicant’s reporting procedures comply

with the requirements of 10 CFR 70.59 and the guidance specific in Regulatory Guide 4.16

(NRC, 1985).

On the basis of this analysis, including a review of specific applicant commitments noted above,

the staff finds that the applicant’s air effluent monitoring during operation of the facility will

ensure that concentrations of radioactivity in air effluent are below the regulatory limits in Part

20.

9.3.2.2 Liquid Effluent Monitoring

As described in SER section 9.3.1.2, the expected concentrations of radioactive materials in

liquid effluents are ALARA and are well below the limits specified in 10 CFR 20, Appendix B

Table 2. The applicant will demonstrate compliance with liquid effluent limits by calculation of

the total effective dose equivalent to the individual who is likely to receive the highest dose in

accordance with 10 CFR 20.1302(b)(1).

The staff reviewed the applicant’s assumptions and conclusions used in its calculations (ER

Sections 4.12 and 8.7) and determined that they are reasonable. The staff’s evaluation of

radiation exposures from normal operations is found in section 4.2.12.2 of the Final EIS.

The applicant has identified one liquid effluent discharge location, the TEEB. Treated liquid

effluent will be discharged in batches to the TEEB. The Liquid Effluent Treatment System

Monitor Tank, which holds treated effluent before discharge, will be sampled and analyzed for

uranium isotope concentrations before it is discharged to the TEEB. The sample from this tank

is representative by virtue of continuous mixing with mechanical agitators or recirculation.

The applicant has proposed measurement sensitivities of 3 x 10-9 :Ci/mL for uranium isotopes

in liquid discharges. The applicant has proposed action levels that are divided into three

priorities (see SER 9.3.2.1).

The UBC Storage Pad stormwater retention pond for the site have the potential to contain small

amounts of radioactivity as a result of runoff from the UBC Storage Pad. The applicant
9-12

proposes to include sampling and analysis of water and sediment from each of the stormwater

runoff retention-detention basins in the Radiological Effluent Monitoring Program.

Table 9.3-2 is a summary of the air and liquid effluent sampling points proposed by the

applicant from the Radiological Effluent Monitoring Program.

Table 9.3-2

Summary of Air and Liquid Effluent Sampling 

Points in the Radiological Effluent Monitoring Program

Effluent Type Sample Location Sample Type Analysis Frequency

Air Separation Building

GEVS Stack

Continuous,

Particulate

Matter

Gross

Alpha/Beta

Weekly

U-234, 235,

236, 238

Quarterly

Composite

Air TSB GEVS Stack Continuous,

Particulate

Matter

Gross

Alpha/Beta

Weekly

U-234, 235,

236, 238

Quarterly

Composite

Air TSB HVAC Stack Continuous,

Particulate

Matter

Gross

Alpha/Beta

Weekly

U-234, 235,

236, 238

Quarterly

Composite

Air Centrifuge Test and

Post-Mortem Facilities

Exhaust Filtration

System Stack

Continuous,

Particulate

Matter

Gross

Alpha/Beta

Weekly

U-234, 235,

236, 238

Quarterly

Composite

Air Process Areas Continuous,

Particulate

Matter

Gross

Alpha/Beta

Weekly

U-234, 235,

236, 238

Quarterly

Composite

Air Non-Process Areas Continuous,

Particulate

Matter

Gross

Alpha/Beta

Quarterly

Liquid Monitor Tank Representative

Grab Sample

U-234, 235,

236, 238

Before

Discharge

As described in Section 1.3 of the applicant’s ER (LES, 2005b), in addition to meeting NRC

regulatory requirements, the applicant has committed to obtaining required Federal and State

permits relating to liquid discharges (primarily related to groundwater protection), as required. 
9-13

For example, the applicant will obtain a discharge permit for use of a septic system at the

Proposed facility. The permit is required by New Mexico Water Quality Control Commission

Regulations, under Section 20.6.2.3.3104 of the New Mexico Administrative Code. In addition

to having direct permitting authority, the State of New Mexico also has oversight responsibility

for EPA water permits granted through the New Mexico Water Quality Bureau.

As described in Section 6.1.2 of the applicant’s ER, the applicant has developed reporting

procedures that when implemented will assure compliance with the regulatory requirements of

10 CFR 70.59 and the guidance provided in Regulatory Guide 4.16 (NRC, 1985).

On the basis of this analysis, the staff finds that the applicant’s liquid effluent monitoring during

operation of the facility ensures that concentrations of radioactivity in liquid effluent will be below

the limits in Part 20.

9.3.2.3 Laboratory Quality Control

The applicant has a quality assurance program that includes controlled written procedures for

sample collection, lab analysis, chain of custody, reporting of results, and corrective actions. An

on-site laboratory and any contractor laboratory used to analyze samples will participate in thirdparty laboratory inter-comparison programs appropriate to the media and analytes being

measured. Examples the applicant cites are the Mixed Analyte Performance Evaluation

Program and the U.S. Department of Energy (DOE) Quality Assurance Program that DOE

administers. All radiological and non-radiological laboratory vendors will be certified by the

National Environmental Laboratory Accreditation Conference, or an equivalent State laboratory

accreditation agency, for the analytes being tested.

The staff finds that these procedures are adequate to validate the analytical results produced by

the Radiological Environmental Monitoring Program (REMP), which is described in the

applicant’s ER (Section 6.1.2).

9.3.2.4 Environmental Monitoring

The applicant has establish its REMP for the facility. The REMP is a major part of the

applicant’s effluent compliance program. The effectiveness of the applicant’s effluent controls

will be confirmed through implementation of the REMP. The purpose of the REMP is to verify

confinement integrity at the facility and to support the primary means of demonstrating

compliance with applicable radiation protection standards, which is effluent monitoring (see

SER Sections 9.3.2.1 and 9.3.2.2).

As part of the REMP, the applicant will establish background and baseline concentrations of

radionuclides in environmental media through sampling and analysis. The REMP will be

initiated at least 2 years before plant operations in order to develop a sufficient environmental

baseline. The types of samples to be collected under the REMP are listed in Table 9.3-3. Staff

evaluation of proposed corrective action levels, which apply to both effluent monitoring and the

REMP, is described in SER Section 9.3.2.1. The scope of the applicant’s REMP meets the

environmental monitoring criteria found in NUREG-1520, Section 9.4.3.3.2(2). As noted in the

applicant’s ER (Section 6.1.2), the REMP sampling locations are based on NRC guidance found

in NUREG-1302 (NRC, 1991).
9-14

Table 9.3-3

Types of Samples Collected under the REMP

Sample Type No. of Sample

Locations

Quantity and

Collection Frequency

Type of Analysis

Continuous Airborne

Particulate Matter

7 Biweekly Gross alpha/beta

Quarterly U-234, 235, 236, 238

Vegetation 8 1- to 2-kg samples,

semiannually

U-234, 235, 236, 238

Groundwater 5 4 liters, semiannually U-234, 235, 236, 238

Basins 1 from each of 3

basins

4 liters water and

1- to 2-kg sediment,

quarterly

U-234, 235, 236, 238

Soil 8 1- to 2-kg samples,

semiannually

U-234, 235, 236, 238

Septic Tank(s) 1 from each tank 1- to 2-kg sludge

sample from the

affected tank(s)

before pumping

U-234, 235, 236, 238

TLD 16 Quarterly Gamma / neutron

The applicant has an adequate and timely program to collect information to determine baseline

concentrations of radionuclides, which information will be used to demonstrate compliance with

applicable radiation protection standards. The staff finds that the applicant’s environmental

monitoring program adequately addresses applicable regulatory requirements and is

acceptable.

9.3.3 ISA Summary

In SER Chapter 3, the staff provides its evaluation of the Integrated Safety Assessment (ISA)

Summary, and documents its conclusion that the ISA Summary 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. SER Chapter 11 contains the staff’s

evaluation of management measures used to ensure that IROFS will satisfactorily perform their

intended safety functions. In this section, the staff evaluates whether the ISA uses acceptable

methods to estimate environmental effects that may result from accident sequences.

Under 10 CFR Part 70, Subpart H, an applicant is to assure, among other things, compliance

with various performance requirements. Section 70.61(c)(3) identifies the environmental

performance requirement that the applicant apply controls such that a credible intermediateconsequence event is unlikely to occur or that the consequence of such an event will not

exceed a 24-hour averaged release of radioactive material outside the restricted area in
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concentrations 5000 times the values in Table 2 of Appendix B to Part 20. The restricted areas

are defined by the applicant in Figure 4.7-2 of the SAR (LES, 2005a). The restricted area is one

contiguous area encompassing the main processing buildings and the UBC Storage Pad. 

Specifically, main processing buildings included within the restricted area are: (1) the

Centrifuge Assembly Building; (2) the Cylinder Receipt and Dispatch Building; (3) the six

Cascade Halls; (4) the six UF6 Handling Areas; (5) the TSB; and (6) the Blending and Liquid

Sampling Area.

The applicant derived enhanced definitions of consequence severity levels based on the

following rationale. For releases to the atmosphere of soluble uranium compounds (i.e., UF6),

the applicable value of Table 2 of Appendix B to Part 20 is 3 x 10-12 :Ci/mL. This is the value for

uranium isotopes U-234, -235, and -238. The performance requirement is, therefore, a 24-hour

averaged concentration of any mixture of these isotopes that will not exceed 3 x 10-12 x 5000, or

1.5 x 10-8 :Ci/mL total radioactivity. The staff assumed a bounding enrichment of 6 percent. 

Using footnote 3 of Part 20, Appendix B, the specific activity is approximately 2.8 :Ci per gram

U. Therefore, the environmental performance requirement, after conversion to a mass

concentration, is a 24-hour averaged concentration of 5.4 mg U/m3

. The environmental

performance requirement, expressed as the enhanced severity level of 5.4 mg U/m3

, was

evaluated at the restricted-area boundary. In sections 3.7 and 3.8 of the ISA Summary, the

applicant has shown that it has adequately reduced the risks to the environment from accidents

for which the consequences could otherwise exceed the environmental-consequence severity

level.

In its ISA Summary, the applicant identified various sequences for radiological and nonradiological accidents which were evaluated to assure adequate protection of worker health and

safety. By assuring that all credible high-consequence events are rendered highly unlikely and

that all intermediate-consequence events are unlikely, the applicant also assured that the

environmental performance requirements of section 70.61(c)(3) will be met. In addition, the

staff independently evaluated the accident sequences to identify whether a credible accident

sequence could occur in which an environmental hazard could be created without also creating

a worker-related hazard. Even when postulating conservative, multiple independent equipment

failures combined with human error, the staff did not identify any accident sequence that would

fail to meet the environmental performance requirements of section 70.61(c)(3). 

The applicant’s approach to risk reduction is to be accomplished through a combination of

preventive and mitigative measures, with an emphasis on preventive measures. A more

complete discussion is found in Chapter 3 of the SER, which addresses accident sequences for

high and intermediate consequences. It also addresses preventive and mitigative measures.

9.4 EVALUATION FINDINGS

The applicant has developed a program to implement adequate environmental protection

measures during operation, which 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
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complies with regulatory requirements imposed by the Commission in Parts 20, 30, 40, 51, and

70.

The NRC staff issued a Final EIS in June 2005, for this licensing action, as required by 10 CFR

51.20. After weighing the environmental impacts of the proposed construction, operation, and

decommissioning of the proposed facility and comparing alternatives, the NRC staff

recommends in its Final EIS that, unless safety issues mandate otherwise, the proposed license

be issued to LES.

9.5 REFERENCES

(LES, 2005a) Louisiana Energy Services (LES). “National Enrichment Facility Safety Analysis

Report,” Revision 6, 2005.

(LES, 2005b) Louisiana Energy Services (LES). “National Enrichment Facility Environmental

Report,” Revision 5, 2005.

(NRC, 1985) U.S. Nuclear Regulatory Commission (NRC). Regulatory Guide 4.16, “Monitoring

and Reporting Radioactivity in Releases of Radioactive Materials in Liquid and Gaseous

Effluents from Nuclear Fuel Processing and Fabrication Plants and Uranium Hexafluoride

Production Plants,” 1985.

(NRC, 1991) U.S. Nuclear Regulatory Commission (NRC). NUREG-1302, “Offsite Dose

Calculation Manual Guidance: Standard Radiological Effluent Controls for Boiling Water

Reactors,” 1991.

(NRC, 1992) U.S. Nuclear Regulatory Commission (NRC). Regulatory Guide 8.34, “Monitoring

Criteria and Methods to Calculate Occupational Radiation Doses,” 1992.

(NRC, 1993) U.S. Nuclear Regulatory Commission (NRC). Regulatory Guide 8.37, ALARA

Levels for Effluents From Materials Facilities,” 1993.

(NRC, 1994a) U.S. Nuclear Regulatory Commission (NRC). NUREG-1491, “Safety Evaluation

Report for the Claiborne Enrichment Center, Homer, Louisiana,” 1994.

(NRC, 1994b) U.S. Nuclear Regulatory Commission (NRC). Information Notice 94-23,

“Guidance to Hazardous, Radioactive, and Mixed-Waste Generators on the Elements of a

Waste Minimization Program,” 1994.

(NRC, 2002) U.S. Nuclear Regulatory Commission (NRC). NUREG-1520, “Standard Review

Plan for the Review of a License Application for a Fuel Cycle Facility,” 2002.

(NRC, 2005) U.S. Nuclear Regulatory Commission (NRC). NUREG-1790, “Environmental

Impact Statement for the Proposed National Enrichment Facility in Lea County, New Mexico,”

June 2005.
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10.0 DECOMMISSIONING

The purpose of this review of the applicant’s decommissioning plan is to determine that the

applicant will be able to decommission the facility safely and in accordance with U.S. Nuclear

Regulatory Commission (NRC) requirements.

At the time of the initial license application for a uranium enrichment 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 NRC with financial assurance to cover the cost of those activities in the

future. The DFP, therefore, should contain an overview of the proposed decommissioning

activities, the methods used to determine the cost estimate, and the financial assurance

mechanism. This overview must contain sufficient details to enable the reviewer to determine

whether the decommissioning cost estimate is reasonably accurate.

10.1 REGULATORY REQUIREMENTS

The following NRC regulations require planning, financial assurance, and record-keeping for

decommissioning, as well as procedures and activities to minimize waste and contamination:

10 CFR 20.1401-1406 “Radiological Criterial for License Termination” (Subpart E)

10 CFR 30.35 “Financial Assurance and Recordkeeping for Decommissioning”

10 CFR 30.36 “Expiration and Termination of Licenses and Decommissioning of

Sites and Separate Buildings or Outdoor Areas”

10 CFR 40.36(d) “Decommissioning Funding Plan”

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 CFR 70.22(a)(9) “Decommissioning Funding Plan”

10 CFR 70.25 “Financial Assurance and Recordkeeping for Decommissioning”

10 CFR 70.38 “Expiration and Termination of Licenses and Decommissioning of

Sites and Separate Buildings or Outdoor Areas”

10.2 REGULATORY ACCEPTANCE CRITERIA

The “Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility,”

NUREG-1520 (NRC, 2002a) and “Consolidated NMSS Decommissioning Guidance,” NUREG1757 (NRC, 2003), define relevant regulatory guidance and appropriate acceptance criteria for

decommissioning and DFPs contained in license applications.

10.3 STAFF REVIEW AND ANALYSIS
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10.3.1 Conceptual Decontamination and Decommissioning Plan

10.3.1.1 Decommissioning Program

10.3.1.1.1 Radioactive Contamination Control

The applicant states, in Section 10.1.5.2 of the safety analysis report (SAR) (LES, 2005) that the

following features will primarily serve to minimize the spread of radioactive contamination during

operation and, therefore, simplify eventual plant decommissioning. As a result, worker

exposure to radiation and radioactive waste volumes are minimized as well.

• Certain activities during normal operation are expected to result in surface and airborne

radioactive contamination. Specially designed rooms are provided for these activities to

preclude contamination spread. These rooms are isolated from other areas and are

provided with ventilation and filtration. The Solid Waste Collection Room, Ventilated

Room, and the Decontamination Workshop meet these specific design requirements.

• All areas of the plant are sectioned off into Unrestricted and Restricted Areas. 

Restricted Areas limit access for the purpose of protecting individuals against undue

risks from exposure to radiation and radioactive materials. Radiation Areas and

Airborne Contamination Areas have additional controls to inform workers of the potential

hazard in the area and to help prevent the spread of contamination. All procedures for

these areas fall under the Radiation Protection Program and serve to minimize the

spread of contamination and simplify the eventual decommissioning.

• 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. Fume hoods filter contaminated air in these areas.

• Curbing, pits, or other barriers are provided around tanks and components that contain

liquid radioactive wastes. These serve to control the spread of contamination in case of

a spill.

10.3.1.1.2 Worker Exposure and Waste Volume Control

The applicant states, in Section 10.1.5.3 of the SAR (LES, 2005), that the following features will

primarily 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.

• During construction, a washable epoxy coating is applied to floors and walls that might

be radioactively contaminated during operation. The coating will serve to lower waste

volumes during decontamination and simplify the decontamination process. The coating

is applied to floors and walls that might be radioactively contaminated during operation

that are located in the Restricted Areas.
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• Sealed, nonporous pipe insulation is used in areas likely to be contaminated. This will

reduce waste volume during decommissioning.

• Ample access is provided for efficient equipment dismantling and removal of equipment

that may be contaminated. This minimizes the time of worker exposure.

• Tanks are provided with accesses for entry and decontamination. Design provisions are

also made to allow complete draining of the wastes contained in the tanks.

• Connections in the process systems provided for required operation and maintenance

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 the planning and

implementing of decontamination procedures. This, in turn, will shorten the durations

that workers are exposed to radiation.

• Worker access to contaminated areas is controlled to assure that workers wear proper

protective equipment and limit their time in the areas.

10.3.1.2 Decommissioning Steps

The plan for decommissioning is to promptly decontaminate or remove all materials, from the

site, that prevent release of the facility for unrestricted use. This approach, referred to in the

industry as DECON, avoids long-term storage and monitoring of wastes on site.

The applicant has briefly described the decommissioning approach to be employed at the

facility. The applicant states that implementation of the DECON alternative for

decommissioning the facility may begin immediately after Separations Building Module

equipment shutdown. The applicant estimates that the DECON alternative will take

approximately 9 years to complete in three phases (3 years/module).

Decommissioning activities will generally include: (1) installation of decontamination facilities;

(2) purging of process systems; (3) dismantling and removal of equipment; (4) decontamination

and destruction of Confidential and Secret Restricted Data material; (5) sales of salvaged

materials; (6) disposal of wastes; and (7) completion of a final radiation survey. Credit is not

taken for any salvage value that might be realized from the sale of potential assets (e.g.,

recovered materials or decontaminated equipment) during or after decommissioning.

Overview

Decommissioning, using the DECON approach, requires residual radioactivity to be reduced

below specified levels so the facilities may be released for unrestricted use. Current NRC

guidelines for unrestricted release serve as the basis for decontamination costs estimated by

the applicant. The applicant intends to remove all enrichment-related equipment from the

buildings in such a manner that only the building shells and site infrastructure remain. The

equipment to be removed by the applicant will include: all piping and components from systems

providing uranium hexafluoride (UF6) containment; systems in direct support of enrichment

(such as refrigerant and chilled water); radioactive- and hazardous-waste-handling systems;
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and contaminated filtration systems, etc. The remaining site infrastructure will include services

such as: electrical power supply; treated water; fire protection; ventilation systems; plant

cooling water; communications; and sewage treatment.

The applicant will install two new facilities dedicated for decontamination of plant components

and structures. Existing plant buildings are assumed to house the facilities. One facility will be

specially designed to accommodate repetitive cleaning of thousands of centrifuges, and the

other will serve as a general-purpose facility used primarily for larger components. The two new

facilities will be the primary locations for decontamination activities. The small decontamination

area in the Technical Services Building (TSB), used during normal operation, may also handle

small items at decommissioning. The applicant estimates that the time for installation is

approximately 1 year. The applicant provided details of the facilities in Section 10.1.7 of the

SAR (LES, 2005).

The applicant states, in Section 10.1.6.1 of the SAR (LES, 2005), that decontaminated

components may be reused or sold as scrap. The applicant will decontaminate all equipment

that is to be reused or sold as scrap to a level at which further use is unrestricted. Materials that

cannot be decontaminated will be disposed of by the applicant in a licensed radioactive-wastedisposal facility.

The applicant states in Section 10.1.6.1 of the application that contaminated portions of the

buildings will be decontaminated as required. When decontamination is complete, the applicant

will survey all areas and facilities on the site, to verify that further decontamination is not

required. The applicant will continue decontamination activities until the entire site is

demonstrated to be suitable for unrestricted use. NRC will independently confirm that the site is

suitable to be released for unrestricted use. NRC will not authorize unrestricted release of

materials and equipment unless all release criteria applicable at the time of decommissioning

have been met. NRC will not authorize release of the site for unrestricted use until the applicant

adequately demonstrates that all decommissioning criteria applicable at the time of

decommissioning have been met.

Process System

The applicant states, in Section 10.1.6.3 of the SAR (LES, 2005), that at the end of the useful

life of each Separations Building Module, the enrichment process will be shut down and UF6 will

be removed to the fullest extent possible by normal process operation. This will be followed by

evacuation and purging with nitrogen. The applicant estimates that the shutdown and purging

portions of the decommissioning process will take approximately 3 months.

Dismantling

Dismantling involves cutting out, disconnecting, etc., all components requiring removal. The

operations themselves may be simple, but very labor-intensive. Depending on the level of

contamination, the use of protective clothing may be required. The applicant states, in Section

10.1.6.4 of the SAR (LES, 2005), that the work process will be optimized, considering the

following:

• Minimizing the spread of contamination and the need for protective clothing;
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• Balancing the number of cutting and removal operations with the resultant

decontamination and disposal requirements;

• Optimizing the rate of dismantling with the rate of decontamination facility throughput;

• Providing storage and laydown space required, as affected by retrievability, criticality

safety, security, etc.; and

• Balancing the cost of decontamination and salvage with the cost of disposal.

The applicant states, in Section 10.1.6.4 of the SAR (LES, 2005), that the details of the complex

optimization process will necessarily be decided near the end of plant life, taking into account

specific contamination levels and available waste disposal sites. To avoid laydown space and

contamination problems, dismantling will likely be allowed to proceed no faster than the

downstream decontamination process. The applicant estimates that dismantling and

decontamination will take approximately 3 years per Separations Building Module.

Sale/Salvage

Items to be removed from facilities can be categorized as potentially reusable equipment,

recoverable scrap, and wastes. However, based on a 30-year operating life, the applicant does

not assume that the facility operating equipment has any reuse value. According to the

applicant, wastes will also have no salvage value. With respect to scrap, the applicant states, in

Section 10.1.6.6 of the SAR (LES, 2005), that a significant amount of aluminum will be

recovered, along with smaller amounts of steel, copper, and other metals. For security and

convenience, the uncontaminated material will likely be smelted to standard ingots, then sold at

market price. However, the applicant has not assigned salvage value to scrap, in estimating its

decommissioning funding requirements. Contaminated material will be disposed of as low-level

radioactive waste.

Disposal

The applicant states, in Section 10.1.6.7 of the SAR (LES, 2005), that all wastes produced

during decommissioning will be collected, handled, and disposed of in a manner similar to that

described for those wastes produced during normal operation. According to the applicant,

wastes will consist of normal industrial trash, non-hazardous chemicals and fluids, small

amounts of hazardous materials, and radioactive wastes. The radioactive waste will primarily

be crushed centrifuge rotors, trash, and citric cake. Citric cake will consist of uranium and

metallic compounds precipitated from citric acid decontamination solutions. The applicant

estimates approximately 5000 cubic meters (m3

) [180,000 cubic feet (ft3)] of radioactive waste to

be generated over the 9-year period of facility decommissioning activities. This waste will be

subject to further volume reduction processes before disposal.

The applicant states, in Section 10.1.6.7 of the SAR (LES, 2005), that radioactive wastes will

ultimately be disposed of in licensed low-level radioactive-waste disposal facilities. Hazardous

wastes will be disposed of in permitted hazardous waste-disposal facilities. Non-hazardous and

non-radioactive wastes will be disposed of in a manner consistent with good industrial practice

and in accordance with all applicable regulations. A complete estimate of the wastes and
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effluents to be generated during decommissioning will be provided in the applicant's plan for

completion of decommissioning, to be submitted to NRC at the time of decommissioning.

The applicant states, in Section 10.1.6.7 of the SAR (LES, 2005), that Confidential and SecretRestricted Data components and documents on site will be disposed of in accordance with the

requirements of 10 CFR Part 95. Classified portions of the centrifuges will be destroyed; piping

will likely be smelted; documents will be destroyed; and other items will be handled in an

appropriate manner. Details will be provided in the facility “Standard Practice Procedures Plan

for the Protection of Classified Matter and Information,” submitted separately, in accordance

with Part 95.

Final Radiation Survey

The applicant states, in Section 10.1.6.8 of the SAR (LES, 2005), that it will perform a final

radiation survey, to verify proper decontamination, to allow the site to be released for

unrestricted use. The initial radiation survey performed before initial operation will provide data

on the natural background radiation of the area that can be used to determine any increase in

levels of radiation. The applicant states, in Section 10.1.6.8 of the SAR (LES, 2005), that

radioactivity over the entire site will systematically be measured in the final survey. The

intensity of the survey will vary depending on the location (i.e., the buildings, the immediate area

around the buildings, the controlled fenced area, and the remainder of the site). The survey

procedures and results will be documented in a report. The report will include, among other

things, a map of the survey site, measurement results, and the site's relationship to the

surrounding area. If the results are above allowable residual radioactivity limits, further

decontamination will be performed until the results are determined to be below limits.

10.3.1.3 Management/Organization

The applicant states, in Section 10.1.5.4 of the SAR (LES, 2005), that management of the

decommissioning program will ensure that proper training and procedures are provided to

protect worker health and safety. The programs will focus heavily on minimizing waste volumes

and worker exposure to hazardous and radioactive materials. Contractors assisting with

decommissioning will likewise be subject to facility training requirements and procedural

controls. The NRC staff finds the applicant's general plans acceptable regarding:

• Responsibilities of management of the decommissioning program;

• Minimization of waste volumes and worker exposures; and

• Procedural control and training requirements for contractors assisting in

decommissioning.

Details related to these three items are typically provided in the detailed decommissioning plan

at the time of decommissioning.

10.3.1.4 Health and Safety
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The applicant states, in Section 10.1.5.5 of the SAR (LES, 2005), that, as with normal operation,

during decommissioning, the policy will be to keep individual and collective occupational

radiation exposures as low as is 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.

10.3.1.5 Waste Management

The applicant states, in Section 10.1.5.6 of the SAR (LES, 2005), 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 permitted

hazardous-waste-disposal facilities located elsewhere. Non-hazardous and non-radioactive

wastes will be disposed of consistent with good industrial practice and in accordance with

applicable regulations.

10.3.1.6 Security/Nuclear Material Control

The applicant states, in Section 10.1.5.7 of the SAR (LES, 2005), that requirements for physical

security and for material control and accounting (MC&A) will be maintained, as required, during

decommissioning, in a manner similar to the programs in force during operation. The plan for

completion of decommissioning, submitted near the end of plant life, will provide a description of

any necessary revisions to these programs.

10.3.1.7 Recordkeeping

The applicant states, in Section 10.1.5.8 of the SAR (LES, 2005), that records important for safe

and effective decommissioning of the facility will be kept in the applicant's files. Information

maintained in these records will include:

1. Records of spills or other unusual occurrences involving the spread of contamination in

and around the facility, equipment, or site. These records may be limited to instances

when contamination remains after any cleanup procedures or when there is reasonable

likelihood that contaminants may have spread to inaccessible areas, as in the case of

possible seepage into porous materials such as concrete. These records will include

any known information on identification of involved nuclides, quantities, forms, and

concentrations.

2. As-built drawings and modifications of structures and equipment, in restricted areas,

where radioactive materials are used or stored, and of locations of possible inaccessible

contamination, such as buried pipes, which may be subject to contamination. 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.

3. Except for areas containing only sealed sources, a list contained in a single document

and updated every 2 years, of the following:
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a. All areas designed and formerly designated as Restricted Areas, as defined

under 10 CFR 20.1003;

b. All areas outside of Restricted Areas that require documentation specified in item

1 above;

c. All areas outside of Restricted Areas where current and previous wastes have

been disposed of, as documented under 10 CFR 20.2108; and

d. 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.

4. Records of the cost estimate performed for the DFP and records of the funding method

used for assuring funds.

10.3.1.8 Decontamination

The following paragraphs discuss the facilities, procedures, and expected results of

decontamination, as described by the applicant in Section 10.1.7 of the SAR (LES, 2005).

The applicant states, in Section 10.1.7.1 of the SAR (LES, 2005), that the primary contamination

throughout the plant will be in the form of small amounts of UO2F2, with even smaller amounts of

UF4 and other compounds. Radiological contamination will be characterized by the applicant,

before beginning decommissioning activities. NRC staff review of the final decommissioning

plan will include a review of the nature and extent of contamination present at the time of

decommissioning.

Facilities

The applicant states, in Section 10.1.7.2 of the SAR (LES, 2005), that a decontamination facility

will be required to accommodate decommissioning. A specialized facility is needed for optimal

handling of the thousands of centrifuges to be decontaminated, along with the UF6 vacuum

pumps and valves. Additionally, a general purpose facility is needed for handling the remainder

of the various plant components. The applicant will most likely install these facilities in existing

plant buildings (such as the Centrifuge Assembly Building).

The applicant describes the specialized facility as having four functional areas: a disassembly

area; a buffer stock area; a decontamination area; and a scrap storage area for cleaned stock. 

The general purpose facility may share the specialized facility decontamination area. However,

because of various sizes and shapes of other plant components needing handling, the

disassembly area, buffer stock areas, and scrap storage areas may not be shared.

The applicant assumes that equipment in the decontamination facilities will include:

• Transport and manipulation equipment;
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• Dismantling tables, for centrifuge externals;

• Sawing machines;

• Dismantling boxes and tanks, for centrifuge internals;

• Degreasers;

• Citric acid and demineralized water baths;

• Contamination monitors;

• Wet-blast cabinets;

• Crusher, for centrifuge rotors;

• Smelting and/or shredding equipment; and

• Scrubbing facility.

The applicant states, in Section 10.1.7.2 of the SAR (LES, 2005), that the decontamination

facilities provided in the TSB for normal operational needs would also be available for cleaning

small items during decommissioning.

Procedures

The applicant states, in Section 10.1.7.3 of the SAR (LES, 2005), that procedures for

decontamination will be developed and approved by plant management to minimize worker

exposure and waste volumes, and to assure work is carried out in a safe manner. If, as

expected, European gas centrifuge enrichment facilities are decommissioned before the

proposed facility, then the experience gained will be incorporated into the procedures to be

developed by the applicant. NRC staff will assess the procedures used and the results of the

final survey. A confirmatory survey is expected to be part of NRC's assessment of the final

survey. The facility and site will not be released for unrestricted use unless it is demonstrated

that the residual contamination is within the limits and criteria in place at the time of

decommissioning.

The applicant states, in Section 10.1.7.3 of the SAR (LES, 2005), that contaminated plant

components will be cut up or dismantled, then processed through the decontamination facilities. 

The applicant states, in Section 10.1.7.3 of the SAR (LES, 2005), that the centrifuges will be

processed through the specialized facility. The following operations will be performed:

• Removal of external fittings;

• Removal of bottom flange, motor and bearings, and collection of contaminated oil;

• Removal of top flange, and withdrawal and disassembly of internals;

• Degreasing of items as required;

• Decontamination of all recoverable items for smelting; and

• Destruction of other classified portions by shredding, crushing, smelting, etc.

Results
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The applicant states, in Section 10.1.7.4 of the SAR (LES, 2005), that conventional

decontamination techniques are effective for all plant items, based on their experience in

decommissioning Urenco facilities in Europe. Recoverable items will be decontaminated and

suitable for reuse except for a very small amount of intractably contaminated material. Material

requiring disposal will primarily be centrifuge rotor fragments, trash, and residue from the

effluent treatment systems. The applicant does not anticipate problems that will prevent the site

from being released for unrestricted use.

10.3.1.9 Decommissioning Costs

The applicant submitted decommissioning cost information consistent with the

recommendations in NUREG-1757, Volume 3, “Consolidated NMSS Decommissioning

Guidance - Financial Assurance, Recordkeeping, and Timeliness” (NRC, 2003). The applicant

presented its decommissioning cost estimate breakdown in SAR Tables 10.1-1 through 10.1-14

(LES, 2005). Decommissioning cost information included labor costs, proposed

decontamination methods and unit costs, waste disposal costs, final survey costs, and costs for

dispositioning depleted uranium tails. The decommissioning costs were based on the

decommissioning experience of Urenco, the applicant’s principal general partner, in

decommissioning gas centrifuge enrichment plants in Europe.

The applicant estimates the cost of decommissioning the facility to be approximately $942

million, in 2004 dollars, which includes an estimated cost of $131 million to decommission the

supporting structures, an estimated tails-disposition cost of $622 million, and a 25 percent

contingency factor, equal to $188 million. More than 97 percent of the cost to decommission

the structures are attributed to the dismantling of the centrifuges and other equipment in the

Separation Building Modules.

The cost analysis for decommissioning the centrifuges and equipment supporting the

Separation Building is classified. The staff reviewed the number of centrifuges, the estimated

man-hours to decontaminate the centrifuges, and the estimated volume of material resulting

from the disposal of the centrifuges and supporting equipment; confirmed that the estimate

includes a 25 percent contingency; and also confirmed that no credit is taken for salvage of

materials or equipment. The use of a 25 percent contingency factor and taking no credit for

salvage value is consistent with staff guidance in NUREG-1757, Volume 3 (NRC, 2003).

The unclassified estimate for decommissioning the remainder of the facility was reviewed

considering labor costs, decontamination methods and unit costs, waste disposal costs,

depleted uranium disposition costs, and final survey costs. Both the classified and nonclassified estimates were evaluated and found to be reasonable and consistent with estimates

provided in NUREG/CR-6477, “Revised Analyses of Decommissioning Reference Non-FuelCycle Facilities” (NRC, 2002b). While the cases in NUREG/CR-6477 (NRC, 2002b) do not

include a case for decommissioning a gas centrifuge enrichment plant, some information can be

compared with the cost estimates provided by the applicant. The following other considerations

were also included in comparing the applicant’s estimates with those in NUREG/CR-6477:

1. The proposed facility operates at subatmospheric pressures that minimize the spread of

contamination throughout the plant.
10-11

2. The plant has design features and operating procedures to minimize releases of uranium

hexafluoride (e.g., cylinder feed, withdrawal, blending, and sampling systems). These

design and operating features are principally intended to minimize worker chemical

exposures, but also result in low contamination levels in occupied areas of the plant.

3. Feed material will be restricted to material specifications meeting the requirements of

American Society for Testing and Materials C787, “Standard Specification for Uranium

Hexafluoride for Enrichment” (ASTM, 2003). Through this specification, the applicant

will control the entry of other radioactive contaminants into the process systems.

4. Most of the process equipment in the plant is aluminum, which can be more easily cut

and processed than the steel components assumed in the cases described in

NUREG/CR-6477 (NRC, 2002b).

Based on the staff’s review of the classified and unclassified information, the staff found the cost

estimate for decommissioning the facility to be reasonable.

The applicant conservatively estimated that the facility will generate 132,942 MT of depleted

uranium over a nominal 30 years of production, and did not reduce the estimate of depleted

uranium based on the planned operations approach where production would actually end 5

years earlier. The applicant estimated the waste processing and disposal cost of UF6 tails at

$4.68 per kilogram of uranium (kg U) or $4,680 per metric ton of uranium (MTU). This cost is

based on the total of the three cost components that make up the total disposition cost for DUF6

(i.e., deconversion, disposal, and transportation). The staff reviewed the basis of each of these

three costs components, and has concluded that they are reasonable. 

The deconversion cost was based on proprietary information on a previously proposed private

deconversion plant using the Cogema dry conversion process producing U3O8 and aqueous

hydrogen fluoride (HF) (NRC, 2005). The proposed process was the same as the plant

Cogema has been operating in Pierrelatte, France for 20 years. The cost estimate was adjusted

to account for differences in planned operating capacities, Euros-to-dollars conversion, and

other costs associated with “Americanization.” “Americanization” refers to costs to obtain

regulatory approval and costs to convert European equipment standards to standards used in

the United States. These cost estimates used a proprietary Urenco business study of a

proposed 3,500 Metric Tons (MT) U/year deconversion plant for the Capenhurst site. The study

was based on a Cogema response to a Urenco request for proposal. The applicant modified

the Cogema information to reflect a 7,000 MT U/year capacity by doubling the operating costs

and by adding funds to reflect the increased capital and construction costs of a larger capacity

plant considering the shared nature of some systems. Additional funds were also added for

Americanizing the design and for licensing.

The Cogema proposal assumed that HF would be sold commercially and did not include the

costs to neutralize aqueous HF to calcium fluoride. Staff consider that neutralization would

have no effect on the overall deconversion costs because those costs would be balanced by the

elimination of costs for equipment for storing HF prior to commercial sale. The cost of disposing

the calcium fluoride ($0.02/kg U) was included in the estimate.

The transportation and disposal costs were based on estimates provided by vendors of

transportation and disposal services (LES, 2005a). Transportation costs were based on an
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estimate from Transportation Logistics International. This transportation estimate ($0.85/kg U)

was independent of distance. The disposal cost of $1.14/kg U for depleted uranium oxides was

based on an estimate provided by Waste Control Specialists. Staff compared the Waste

Control Specialists estimate to an estimate for disposal of decommissioning wastes the

applicant had obtained from Envirocare of Utah and found it to be consistent. The Envirocare

disposal estimate for decommissioning waste was $2.12/m3

 ($75/ft3) (LES, 2004). For the

disposal of U3O8, the equivalent disposal cost at Envirocare is $1.07/kg U. 

Further, the applicant submitted an estimate for tails disposition from the U.S. Department of

Energy (DOE) (DOE, 2005) as additional evidence of the reasonableness of their estimate. The

DOE estimate included conversion, transportation, storage, disposal, and decommissioning

costs of the conversion facility and totaled $3.34/kg DUF6 ($4.91/kg U) in 2004 dollars. This is

less than 5 percent of the difference in the applicant’s estimate of $4.68/kg U. Staff considers

that the DOE estimate provides additional assurance that the applicant’s estimate of depleted

uranium disposition costs is reasonable.

Based on the staff’s review of the classified and unclassified information, the staff found that the

cost estimate for decommissioning the facility is reasonable and the cost estimate fulfills the

requirements of 10 CFR 30.35(e), 10 CFR 40.36(d), and 10 CFR 70.25(e) and the evaluation

criteria in Section 4.1 of NUREG-1757, Volume 3 (NRC, 2003) for the following reasons:

• The cost estimate is based on documented and reasonable assumptions;

• The cost estimates for individual facility activities and components are reasonable and,

to the extent possible, consistent with NRC cost estimation reference documents;

• The cost estimate reflects decommissioning under appropriate facility conditions;

• The cost estimate includes costs for labor, equipment and supplies, overhead and

contractor profit, sampling, and miscellaneous expenses;

• The cost estimate includes costs for all major decommissioning activities, including

planning and preparation; decontamination or dismantling facility components;

packaging, shipping, and disposal of wastes; restoration of facility grounds; and the final

radiation survey.

• The computations are correct;

• No credit is taken for salvage value;

• The decommissioning cost estimate includes an adequate contingency factor of 25

percent; and

• The decommissioning cost estimate provides a description of how it will be adjusted

periodically over the life of the facility.

10.3.1.10 Financial Assurance for Decommissioning
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The applicant stated it will utilize a surety bond method to provide reasonable assurance of

decommissioning funding as required by 10 CFR 30.35(f)(2), 10 CFR 40.36(e)(2), and 10 CFR

70.25(f)(2). The applicant provided draft copies of the surety bond and standby trust language. 

Finalization of the specific financial instruments to be utilized will be completed, and signed

originals of those instruments will be provided to the NRC for final confirmation of the instrument

prior to the applicant receiving licensed material at the facility. In addition, the applicant

committed to provide continuous financial assurance through the completion of

decommissioning and termination of the licenses. Although the applicant plans to sequentially

install and operate the separations buildings modules over time, financial assurance for

decommissioning of the full-size facility will initially be set aside, as well as an estimated 3 years

of tails disposition. Thereafter, funding for tails disposition will be provided at a rate in

proportion to the amount of accumulated tails onsite up to the maximum amount of the tails as

described in Sections 10.2.1 and 10.2.2 of the SAR. If the applicant’s schedule for future

enrichment module phase-in changes, the applicant indicated it may reduce the funding for

facility decontamination and decommissioning cost estimate to reflect the actual number of

operating modules.

The surety bond method to be adopted by the applicant will provide a guarantee that

decommissioning costs will be paid in the event the applicant is unable to meet its

decommissioning obligations at the time of decommissioning. The surety bond will be

structured consistent with applicable NRC requirements and in accordance with NRC

regulatory guidance contained in NUREG-1757, Volume 3 (NRC, 2003). Accordingly, the

applicant stated that its surety bond will contain, but not be limited to, the following attributes:

• The surety bond will be open-ended or, if written for a specified term, such as 5 years,

will be renewed automatically unless 90 days or more prior to the renewal date, the

issuer notifies the NRC, the trust to which the surety is payable, and the applicant of its

intention not to renew. The surety bond will also provide that the full face amounts are

paid to the beneficiary automatically prior to the expiration without proof of forfeiture if

the applicant fails to provide a replacement acceptable to the NRC within 30 days after

receipt of notification of cancellation.

• The surety bond will be payable to a standby trust established for decommissioning

costs. The trustee and trust will be ones acceptable to the NRC. For instance, the

trustee may be an appropriate State or Federal government agency or an entity which

has the authority to act as a trustee and whose trust operations are regulated and

examined by a Federal or State agency.

• The surety bond and standby trust will remain in effect until the NRC has terminated the

license.

In accordance with 10 CFR 30.35(e), 10 CFR 40.36(d), and 10 CFR 70.25(e), the applicant will

update the decommissioning cost estimate for the facility and the associated funding levels,

over the life of the facility. These updates will take into account changes resulting from inflation

or site-specific factors, such as changes in facility conditions or expected decommissioning

procedures. These funding level updates will also address anticipated accumulated tails. As

required by 10 CFR 30.35(e), 10 CFR 40.36(d), and 10 CFR 70.25(e), such updates will occur

at least every 3 years.
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In Sections 1.2.5, 10.2.1, and 10.2.2 of the SAR (LES, 2005), the applicant described its

approach to funding the surety bond financial assurance instrument to be used and for

updating the DFP over time. Financial assurance for decommissioning will be provided during

the operating life of the facility. Initially, the applicant will provide funding for decommissioning

the facility and the tails expected to be generated during the first three years of operation. 

Funding for tails dispositioning will thereafter be provided at annual intervals. In Section 1.2.5

of the SAR (LES, 2005), the applicant requested an exemption to fund decommissioning on an

incremental basis. Section 1.2.3.6 of this SER discusses the approval of this exemption

request as required in 10 CFR 40.14 and 10 CFR 70.17.

Updates of the DFP and the financial assurance instrument will be provided as follows:

• In the initial executed financial assurance instrument submitted prior to receipt of

licensed material, the applicant will provide full funding for decontamination and

decommissioning of the full-size facility.

• In the initial executed financial assurance instrument submitted prior to receipt of

licensed material, the applicant will provide funding for the disposition of depleted

uranium tails in an amount needed to disposition the first three years of depleted

uranium tails generation.

• Subsequent updated decommissioning funding estimates and revised funding

instruments for facility decommissioning will be provided at least every three years. If

the applicant reduces the amount of funding for the facility because of a change in

module phase-in, the revisions will be submitted prior to the operation of each facility

module. This will allow the applicant to modify its initial facility decommissioning

funding approach to reflect changes in future enrichment module phase-in schedules.

• Subsequent updated decommissioning cost estimates and revised funding instruments

for depleted uranium disposition will be provided annually on a forward-looking basis to

reflect projections of depleted uranium byproduct generation.

The above DFP update schedule will provides updates at a frequency of at least every 3 years

in accordance with 10 CFR 30.35(e), 10 CFR 40.36(d), and 10 CFR 70.25(e).

The initial financial obligation will be the entire facility decommissioning cost ($131 million), the

cost for dispositioning the first three years of generation of depleted uranium ($22.7 million

based on generating 4,861 MT of depleted uranium in the first 3-year period), and a 25 percent

contingency of $38.5 million giving a total decommissioning obligation for this period of $192

million. These estimates are in 2004 dollars. This approach to funding the financial assurance

instrument is acceptable to NRC staff because the amount of financial assurance will be

sufficient to cover the decommissioning obligation of the licensee at any point in time in the

event that the licensee is unable to complete decommissioning for any reason.

Because final executed copies of the financial assurance mechanism will not be provided to

NRC until prior to receipt of licensed material, NRC staff is imposing the following license

conditions:
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“1. The licensee shall provide final copies of the proposed financial assurance

instruments to NRC for review at least 6 months prior to the planned date for

obtaining licensed material, and provide to NRC final executed copies of the

reviewed financial assurance instruments prior to the receipt of licensed

material. 

2. In addition, the Decommissioning Funding Plan cost estimate shall be updated

as follows:

a. In the first executed financial assurance instrument submitted prior to

receipt of licensed material, the licensee will provide full funding for

decontamination and decommissioning of the full-size facility.

b. In the first executed financial assurance instrument submitted prior to

receipt of licensed material, the licensee will provide funding for the

disposition of depleted uranium tails in an amount needed to disposition

the first three years of depleted uranium tails generation.

c. Subsequent updated decommissioning funding estimates and revised

funding instruments for facility decommissioning will be provided, at a

minimum, every three years. Any proposed reduction based on changes

to module phase-in will be submitted 6 months prior to the scheduled

operation of the facility module.

d. Subsequent updated decommissioning cost estimates and revised

funding instruments for depleted uranium disposition will be provided

annually on a forward-looking basis to reflect projections of depleted

uranium byproduct generation.

3. The Decommissioning Funding Plan cost estimates shall be provided to NRC for

review, and subsequently, after resolution of any NRC comments, final executed

copies of the financial assurance instruments shall be provided to NRC.”

With the above proposed license conditions and the exemption discussed in Section 1.2.3.6,

NRC staff finds the DFP and proposed surety bond method acceptable.

10.4 EVALUATION FINDINGS

The NRC staff has evaluated the applicant’s decommissioning financial assurance plan in

accordance with NUREG-1757 (NRC, 2003). On the basis of this evaluation, the NRC staff

has determined that the applicant’s financial assurance for decommissioning provides sufficient

funding to ensure decommissioning and decontamination of the facility even if the licensee is

unable to meet its financial obligations, and, therefore, provides reasonable assurance of

protection for workers, the public, and the environment.

Because final executed copies of the financial assurance mechanism will not be provided to

NRC until prior to receipt of licensed material, NRC staff is imposing the following license

conditions:
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“1. The licensee shall provide final copies of the proposed financial assurance

instruments to NRC for review at least 6 months prior to the planned date for

obtaining licensed material, and provide to NRC final executed copies of the

reviewed financial assurance instruments prior to the receipt of licensed

material. 

2. In addition, the Decommissioning Funding Plan cost estimate shall be updated

as follows:

a. In the first executed financial assurance instrument submitted prior to

receipt of licensed material, the licensee will provide full funding for

decontamination and decommissioning of the full-size facility.

b. In the first executed financial assurance instrument submitted prior to

receipt of licensed material, the licensee will provide funding for the

disposition of depleted uranium tails in an amount needed to disposition

the first three years of depleted uranium tails generation.

c. Subsequent updated decommissioning funding estimates and revised

funding instruments for facility decommissioning will be provided, at a

minimum, every three years. Any proposed reduction based on changes

to module phase-in will be submitted 6 months prior to the scheduled

operation of the facility module.

d. Subsequent updated decommissioning cost estimates and revised

funding instruments for depleted uranium disposition will be provided

annually on a forward-looking basis to reflect projections of depleted

uranium byproduct generation.

3. The Decommissioning Funding Plan cost estimates shall be provided to NRC for

review, and subsequently, after resolution of any NRC comments, final executed

copies of the financial assurance instruments shall be provided to NRC.”

10.5 REFERENCES

(ASTM, 2003) American Society for Testing and Materials (ASTM). ASTM C787, “Standard

Specification for Uranium Hexafluoride for Enrichment,” 2003.

(DOE, 2005) U.S. Department of Energy (DOE), letter to Louisiana Energy Services,

“Conversion and Disposal of Depleted Uranium Hexafluoride (DUF6) Generated by Louisiana

Energy Services, LP (LES),” March 1, 2005.

(LES, 2004) Louisiana Energy Services (LES), letter to U.S. Nuclear Regulatory Commission,

“Response to NRC Request for Additional Information Regarding Decommissioning Funding

Plan,” December 10, 2004.
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(LES, 2005a) Louisiana Energy Services (LES), letter to U.S. Nuclear Regulatory Commission,

“Clarifying Information Related to Depleted Uranium UF6 Disposition Costs and Request for

License Condition,” March 29, 2005.

(LES, 2005b) Louisiana Energy Services (LES). “National Enrichment Facility Safety Analysis

Report,” Revision 6, 2005.

(NRC, 2002a) U.S. Nuclear Regulatory Commission (NRC). NUREG-1520, “Standard Review

Plan for the Review of a License Application for a Fuel Cycle Facility,” 2002.

(NRC, 2002b) U.S. Nuclear Regulatory Commission (NRC). NUREG/CR-6477, “Revised

Analyses of Decommissioning Reference Non-Fuel-Cycle Facilities,” 2002.

(NRC, 2003) U.S. Nuclear Regulatory Commission (NRC). NUREG-1757, Volume 3,

“Consolidated NMSS Decommissioning Guidance - Financial Assurance, Recordkeeping, and

Timeliness,” 2003.

(NRC, 2005) Johnson, T.C., U.S. Nuclear Regulatory Commission (NRC). April 8, 2005, “InOffice Review Summary: Louisiana Energy Services Decommissioning Funding,” April 29,

2005.
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11-1

11.0 MANAGEMENT MEASURES

Management measures are functions that the applicant performs, generally on a continuing

basis, which are applied to items relied on for safety (IROFS), to ensure compliance with

established performance requirements that the IROFS are available and reliable. Management

measures shall be implemented to assure 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). This

chapter addresses each of the management measures included in the 10 CFR Part 70 definition

of management measures, including: (a) configuration management (CM); (b) maintenance; (c)

training and qualifications; (d) procedures; (e) audits and assessments; (f) incident

investigations; (g) records management; and (h) other quality assurance (QA) elements.

The purpose of this review is to verify whether the applicant provided conclusive information to

ensure that the management measures applied to IROFS, as documented in the ISA Summary,

provide assurance that the IROFS will be available and reliable, consistent with the performance

requirements of 10 CFR 70.61. If a graded approach is used, the review will also determine

whether the measures are applied to the IROFS in a manner commensurate with the IROFS’

importance to safety.

11.1 REGULATORY REQUIREMENTS

The requirements for fuel cycle facility management measures are specified in 10 CFR Part 70,

“Domestic Licensing of Special Nuclear Material.”

1. 10 CFR 70.4 states that management measures include: CM; maintenance; training and

qualifications; procedures; audits and assessments; incident investigations; records

management; and other QA elements.

2. 10 CFR 70.62(a)(3) states that records must be kept for all IROFS failures; describes

required data to be reported; and sets time requirements for updating the records.

3. 10 CFR 70.62(d) requires an applicant to establish management measures, for

application to engineered and administrative controls and control systems that are

identified as IROFS, pursuant to 10 CFR 70.61(e), to ensure they are available and

reliable.

4. 10 CFR 19.12 states requirements for workers instructions that are applicable to

personnel training and qualifications. 

5. 10 CFR 70.22(a)(8) states requirements for license applications to address proposed

procedures to protect health and minimize danger to life and property.
11-2

6. 10 CFR 70.72 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.

7. 10 CFR 70.74(a) and (b) state requirements for incident investigation and reporting. 

11.2 REGULATORY ACCEPTANCE CRITERIA

The acceptance criteria for the U.S. Nuclear Regulatory Commission’s (NRC’s) review of the

applicant’s management measures program are outlined in Section 11.4.3 of NUREG-1520

(NRC, 2002). 

11.3 STAFF REVIEW AND ANALYSIS

11.3.1 Configuration Management Program (CM)

The applicant’s CM function is described in Section 11.2 of the Safety Analysis Report (SAR)

(LES, 2005).

11.3.1.1 CM Policy

The goal of the CM program is to ensure that accurate and current documentation matches the

facility’s physical/functional configuration, and to ensure that IROFS are available and reliable,

and comply with regulatory requirements. The CM program will be implemented throughout

facility design, construction, testing, and operation. The applicant has defined the CM policy for

the facility in Section 11.1.1 of the SAR (LES, 2005). During design and construction, the

Engineering and Contracts organization has primary responsibility for the implementation of the

CM program, through the design control process. As the project transitions from the design and

construction phase to the operations phase, the responsibility for CM transitions to the Technical

Services organization. In all cases, the applicant’s CM program will provide for the control of key

documentation, including the ISA, in accordance with design control, document control, and

records management procedures. All design changes will undergo formal review, including

interdisciplinary reviews, as appropriate. The applicant’s CM program policy statement

addresses the following topics: (1) the scope of SSCs and administrative items that will be

covered under the CM program; (2) CM interfaces with other management measures and QA; (3)

the objectives of the CM program; (4) a description of CM activities; and (5) organizational

descriptions of duties and responsibilities. 

The CM program will be applied to all SSCs that the ISA identifies as IROFS, and any items that

impact the function of IROFS. The CM program will also include all calculations, safety

analyses, design criteria, engineering drawings, system descriptions, technical documents, and

specifications that establish specific design requirements for IROFS. The scope of documents

under CM expands to include relevant data and procedures applicable to IROFS during

construction, initial startup, and operations. CM procedures will provide for evaluation,

implementation, and tracking of changes to IROFS, as well as processes, equipment, computer

programs, and personnel activities that impact the reliability of IROFS.
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The applicant describes how the CM function relates to other management measures, including

QA; records management; maintenance; training and qualifications; incident investigation, audits

and assessments; and procedures. Specifically, the QA Program establishes the framework for

CM, and the records associated with IROFS or items affecting IROFS. Records are generated

and processed in accordance with QA Program requirements. The design basis, which will be

controlled under CM, establishes maintenance requirements.

The applicant clearly states the objectives of the CM function, which will be to ensure design and

operation within the design bases of IROFS. This will be accomplished by identifying and

controlling documentation associated with IROFS, controlling changes to IROFS, and

maintaining the physical configuration of the facility consistent with the approved design. 

Changes to the approved design are subject to review and approval, to ensure consistency

within the design bases of IROFS. The CM function and the design will be subject to periodic

audits and assessments to confirm that the design and physical configuration will be consistent

with the design basis. Identified problems will be subject to the applicant’s corrective action

process which is identified in the applicant’s QA Program Description (QAPD) and associated

implementing procedures. Prompt corrective actions will be developed as a result of incident

investigations or in response to unfavorable audit or assessment results.

The applicant provides a description of CM activities. CM activities will be conducted and

documented under design control provisions and involve the systematic preparation, review, and

approval of design and construction documentation for ensuring that consistency is maintained

between design and design bases for IROFS. CM activities also provide for operation of IROFS

within the limits specified in the ISA and control changes to the facility in accordance with 10

CFR 70.72. Finally, CM activities provide records for certifying that personnel conducting

activities relied on for safety are appropriately trained and qualified and revisions for ensuring

only formally reviewed and approved procedures, specifications, and drawings are used for

activities associated with IROFS.

During design and construction, the CM program will be implemented by the Engineering and

Contracts organization, which is under the direction of the Engineering and Contracts Manager. 

The engineering organization is organized by discipline. Responsible lead engineers will be in

charge of each discipline. In addition to having primary technical responsibility for work

performed by their disciplines, lead engineers will be responsible for conducting interdisciplinary

reviews of design changes. Lead engineers in each discipline report to design managers or

project managers, who report to the Engineering and Contracts Manager. Work performed

within the Engineering and Contracts organization is also subject to reviews, as appropriate, by

construction management, operations, QA, and procurement personnel. The applicant’s various

departments and contractors will perform quality-related activities. The QA Programs developed

by the applicant’s primary contractors will be consistent with the requirements of the applicant’s

QA Program for those contracted activities determined to be within the scope of the QA Program. 

All applicant and contractor personnel are responsible for identifying quality problems. If a

disagreement exists among personnel concerning a quality-related issue, the issue will be

elevated to the appropriate level of management, up to and including the applicant’s President.

11.3.1.2 Design Requirements

During design and construction, the Engineering and Contracts organization establishes and

maintains design requirements and associated design bases. Whereas, during the operations
11-4

phase, this function transitions to the Technical Services Organization. Design requirements will

be documented and maintained in a design requirements document, which also contains a listing

of IROFS. The design requirements documentation and the design-basis documents will be

controlled in accordance with the design control provisions of the CM program. The design

documents associated with IROFS, items that affect the function of IROFS, and items required to

satisfy regulatory requirements, as well as analyses that constitute the ISA of the design bases,

will be subject to interdisciplinary reviews and design verification. Design changes are evaluated

to ensure consistency within the design basis. Qualified individuals will prepare and review

design documents, and the responsible manager approves the corresponding design

documentation. The QA director conducts audits on the design-control process, using

independent, technically qualified staff.

The applicant describes a design-verification process for design documents where emphasis will

be placed on assuring conformance with applicable codes, standards, and license application

design criteria. Design verification ensures that design documents are consistent with the design

criterion used for IROFS. Design verification will be accomplished through design review,

alternative calculations, or qualification testing. Engineering personnel assigned to perform interdisciplinary checks and reviews of design documents have the authority to withhold approval

pending resolution of all questions concerning the design. Design bases will be appropriately

referenced in the design document, and the reviewer will verify proper application. Designverification testing will demonstrate adequate performance under the most adverse design

conditions. Independent design verification will be accomplished before other organizations use

the design document, to support other activities. In all cases, design verification for an item must

be complete, before the item may be relied on to perform its intended function. After verification

and approval, the design document will be distributed to the appropriate parties through the

document control center.

Configuration control will be accomplished, during design, through the use of design-control

procedures, which address preparation, review, verification, approval, release, and distribution of

design documents for use. Engineering documents will be assigned a QA-Level verification that

is based on safety significance. Configuration verification will be accomplished through the

design verification process. All acceptance criteria the designer establishes will be incorporated

into documents used to perform work associated with the design. Documentation will also be

maintained demonstrating that work has been performed and configuration managed.

11.3.1.3 Document Control

The applicant will be required to establish a document control program that will be implemented

and controlled through procedures. Document control procedures govern the preparation and

issuance of, as well as changes to, key documents. In addition to document control procedures,

the applicant will utilize an electronic document management system that will be used to file

project records and to make available the latest controlled copy of design documents. Controlled

documents will be maintained in the system until cancelled or superseded. 

The document control and records management procedures will be a part of the CM program

and as such, they will capture the following documents, which are listed in approved procedures: 

(a) design-requirements documents; (b) design-bases documents; (c) the ISA of the design

bases of IROFS; (d) nuclear criticality safety evaluations and analyses; (e) as-built drawings; (f)

specifications; (g) procedures that are IROFS; (h) training procedures; (i) QA documents; (j)

maintenance documents; (k) audit and assessment reports; (l) emergency operating procedures;
11-5

(m) emergency response plans; (n) system-modification documents; and (o) engineering

documents.

11.3.1.4 Change Control

The applicant is mandated by 10 CFR 70.72 to implement a change-control function, and

associated procedures that control changes to the technical baseline, as part of it’s CM program. 

Review of the applicant’s change control process, the applicant stated that for each change to

the facility or to the activities carried out by personnel, an evaluation of the proposed change will

be performed in accordance with the requirements of 10 CFR 70.72, as applicable. For every

modification to the facility, the applicant has described their process of evaluation of the

modification and how the modification will be performed to determine the need for any required

changes or additions to the facility’s processes, procedures, personnel training, testing program,

or regulatory documents. The applicant provided a descriptive review process which will be

implemented to ensure changes to the facility’s physical/functional configuration, procedures,

and controlled documents are implemented in accordance with the provisions of 10 CFR 70.72. 

Changes to the above items will be controlled through procedures. Before a change is

implemented, the change-control process will require an appropriate level of technical,

management, and safety review and approval. 

During the design phase, changes to the design will include a systematic review of the design

bases. Consistency between documents will be ensured through the interdisciplinary review

process, which will ensure that design changes do not impact the ISA; or are accounted for in

subsequent changes to the ISA; or such design changes will not be approved and/or

implemented. Before the issuance of the license, the applicant has committed to notify the NRC

of potential changes that reduce the level of commitments or the safety margin in the design

bases of IROFS. 

During the construction phase, the applicant has described the documentation, review, and

approval processes that will be used to control all changes to documents issued for construction,

fabrication, and procurement. The review of the applicant’s processes, include the evaluation of

changes to design documents and an evaluation of the design basis of IROFS and the ISA which

will be required to be posted against each affected design document. Vendor drawings and data

will undergo interdisciplinary reviews to ensure compliance with engineering and procurement

specifications. During construction, design changes will continue to be evaluated against the

approved design bases. The applicant is mandated by 10 CFR 70.72 to have a fully

implemented CM program which will be inclusive of reporting and tracking changes made without

prior NRC approval. 

The applicant’s description of design control used during the operations phase, describes the

methodologies used to document, review, and approve design changes before implementation. 

During facility operation, the applicant describes measures that will be implemented to ensure

that the quality of facility SSCs are not compromised by planned changes and modifications. 

Modification procedures will be implemented that state the requirements which will be met to

implement a modification, as well as the requirements for initiating, approving, monitoring,

designing, verifying, and documenting modifications. Modification procedures will be written to

ensure that policies are maintained to satisfy the applicant’s QA requirements. Each

modification will be evaluated for any required changes to the facility’s procedures, training

programs, testing programs, or regulatory requirements described in 10 CFR 70.72. A nuclear
11-6

criticality safety evaluation will be prepared and approved for any change that involves or could

affect uranium on-site. The evaluation of modifications will take into consideration such items as: 

(a) radiation exposure; (b) cost; (c) lessons-learned; (d) QA aspects; (e) operability issues; (f)

maintenance issues; (g) constructability issues; (h) testing requirements; (i) environmental

concerns; and (j) human factors. 

11.3.1.5 Assessments

Review of the applicant’s CM assessment function included description of periodic assessments

of the CM program and design to determine the effectiveness of the CM program and to correct

deficiencies. This review of Assessments will be carried out in accordance with the applicant’s

QAPD, Appendix A, Section 18, “Audit Schedules.” The applicant describes the implementation

strategy used for scheduling internal and external audits/assessments in a manner consistent

with the activities status and importance to on-going work activities. These assessments will be

based upon thorough reviews of the adequacy of documentation and effectiveness of the quality

assurance program and system walk-downs of the as-built facility. As stated in the applicant’s

QAPD, Appendix A, Section 18, “Audits” these periodic assessments will confirm the

implementation of the CM programmatic elements and verification of compliance and

consistency of design with the design bases. If problems or deficiencies are encountered, and

have been evaluated as significant conditions adverse to quality for reportability to the NRC, 

prompt corrective actions will be developed in accordance with the applicant’s QAPD, Appendix

A, Section 16, Corrective Action Procedures and 10 CFR Part 21, “Reporting of Defects and

Noncompliance.”

11.3.2 Maintenance

The maintenance and functional testing programs is described in Section 11.2 of the applicant’s

SAR (LES, 2005). The applicant has outlined those planned and scheduled maintenance and

functional testing programs, for IROFS, that will ensure that equipment and controls will be

maintained in a condition of readiness, and will perform their safety functions when required. 

The applicant has described the performance measures and use of surveillance/monitoring

activities to detect degradation or adverse trends in the performance of IROFS, so that proactive

action may be taken before failure. The applicant has described methods and practices in

correlation with corrective maintenance and the applicant’s Quality Assurance Program function

Section 10, of the applicant’s QAPD, “ Inspection” will be implemented to ensure that equipment

which has unexpectedly degraded or failed will be repaired or replaced. Finally, the applicant

has defined a functional testing program that defines how functional testing will be performed

under pre-operational and operational conditions, and measures for ensuring that IROFS will be

capable of performing their safety functions, when required. The maintenance organization

plans, schedules, and tracks maintenance activities, and maintains records for these activities. 

The applicant has described implementing measures that ensure that the quality of facility SSCs

will be not compromised by planned modifications and/or maintenance activities. Modifications

to facility SSCs will be controlled through the use of facility administrative procedures that the

Technical Services Manager approves, and the QA Manager concurs. Revisions to these

procedures for controlling modifications must also receive approval and concurrence from these

two respective managers. The administrative procedures will contain requirements that will

implement modification requirements, as well as requirements for initiating, approving,
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monitoring, designing, verifying, and documenting modifications. Administrative procedures will

satisfy the QA standards specified in the applicant’s QAPD. 

The applicant has described a set of methods and practices that will be applied to it’s corrective

maintenance, preventive maintenance, and functional-test maintenance elements. The

applicant’s descriptive programs will include the preparation of authorized work instructions that

will contain detailed steps of work sequencing and written procedures associated with the

performance of these methods and practices. The applicant has stated that the methodology

and practices include: (a) parts lists; (b) as-built or redlined drawings; (c) notification to the

Operations organization before conducting repairs and removing IROFS from service; (d)

radiation work permits; (e) replacement with like-kind parts and the control of new or replacement

parts, to ensure compliance with 10 CFR Part 21; (f) compensatory measures while performing

work on IROFS; (g) procedural control of the removal of components from service and for the

return of components to service; (h) ensuring safe operations during the removal of IROFS from

service; and (i) notification to Operations personnel that repairs have been completed. Written

procedures for the performance of maintenance activities will include the methods and practices

listed above. The applicant will require contractors performing maintenance work activities

involving IROFS to follow the same maintenance procedures that the applicant’s personnel are

required to use.

When maintenance activities involve IROFS, the applicant has described the methods listed

below for corrective maintenance, preventive maintenance (PM), functional testing after

maintenance, and surveillance/monitoring activities:

• Pre-maintenance activities require reviews of the work to be performed, including

procedure reviews for accuracy and completeness; 

• New procedures or work activities that involve or could affect uranium on-site require

preparation and approval of a Nuclear Criticality Safety evaluation (NCS) and if required

an NCS analysis; 

• Procedure steps will require notification of all affected parties (operators and appropriate

managers) before performing work, as well as after work is completed (the notification will

include potential degradation of IROFS during the planned maintenance); and 

• Work control will be implemented through comprehensive procedures that maintenance

workers will follow. 

The applicant has established protocols for the review of these procedures which include review

concurrence obtained from the various safety disciplines, including nuclear criticality safety, fire

protection, radiation safety, industrial safety, and chemical process safety. The applicant has

structured these procedures to describe as a minimum the following: (a) the qualifications of

personnel authorized to perform maintenance, functional testing, or surveillance activities; (b)

controls and specifications on the use of replacement components or materials; (c)

post–maintenance testing requirements to verify component operability; (d) tracking and records

management of maintenance activities; and (e) safe work practices (e.g., component tag-out;

confined-space entry requirements; moderation control or exclusion area; radiation or hot work

permits; and criticality, fire, chemical, and environmental safety for workers). 
11-8

11.3.2.1 Corrective Maintenance

The applicant’s corrective maintenance function will involve the repair or replacement of

equipment that has unexpectedly degraded or failed. For IROFS, corrective maintenance will

restore the equipment to acceptable performance through the use of planning, controlling, and

documentation of repair and replacement activities. After performing corrective maintenance on

IROFS, and before the applicable IROFS are declared operable, the necessary functional testing

of the IROFS will be performed to ensure that the IROFS perform their intended safety functions. 

The applicant has indicated that, with respect to corrective maintenance, the applicant’s

Corrective Action Program (CAP) will require facility personnel to determine the cause of

conditions adverse to quality and will also require facility personnel to act to correct these

conditions.

11.3.2.2 Preventative Maintenance (PM)

The applicant describes their Preventative Maintenance program activities as containing

activities that will be inclusive of preplanned, scheduled periodic refurbishment, partial or

complete overhaul or replacement of IROFS to ensure that they are reliable and available to

perform their intended functions. Planning for PM will take into consideration the results of

surveillance and monitoring, including failure history as well as, instrument calibration and

testing. PM tasks, frequencies, and procedures will be initially determined using industry

experience, industry standards, and vendor recommendations, taking into consideration the

appropriate balance between the objectives of equipment failure prevention through PM and the

objective of minimizing unavailability of IROFS, because of PM activities. Feedback from

acquired experience in maintenance activities and incident investigations will be used, as

appropriate, to modify the frequency or scope of PM activities. Deviations from manufacturer

recommendations or industry standards will be documented. All records pertaining to PM will be

maintained in accordance with the applicant’s Records Management System and in accordance

with 10 CFR § 70.62 the results of the applicant’s PM activities related to IROFS reliability and

failure will be evaluated by all safety disciplines to determine any ISA impact and the need for

any updates.

The applicant will develop and implement the necessary post-maintenance functional testing of

SSCs after conducting PM on IROFS and before returning IROFS to service. Post-maintenance

testing will ensure that IROFS that have undergone PM are capable of performing their safety

functions.

11.3.2.3 Surveillance and Monitoring

The applicant has described their surveillance and monitoring system as being proactive, which

will be utilized to detect degradation and adverse trends of IROFS and SSCs identified as

IROFS, prior to component failure. This proactive system primarily consists of periodic

surveillance tests, plant computer monitoring, operator rounds, and walk-downs, and/or

inspections performed at specified intervals for all IROFS, as well for all items that could impact

the function of IROFS and SSCs identified as IROFS. The applicant’s surveillance and

monitoring system utilizes the selection of measurements of monitored performance

specifications to detect degradation or adverse trends in IROFS, whereas, proactive corrective

action may be taken prior to failure. Results of surveillance monitoring will be trended to detect

signs of performance degradation. Any signs of potential performance degradation will be
11-9

addressed by appropriate corrective actions, such as corrective maintenance or adjustment of

preventive maintenance frequencies. Parameters to be monitored or evaluated during

surveillance/monitoring activities will be selected, based on their ability to detect the predominate

failure modes of critical components. Plant performance criteria to be monitored will be

established using industry operating experience, plant equipment operating experience,

operating data, and surveillance data.

The applicant has stated that the surveillance program will adhere to “Inspection, Testing, and

Maintenance Baseline Design Criteria,” as stated in 10 CFR 70.64. The surveillance(s) will be

included in the work control process, and the results will be trended. Frequencies will be

adjusted or other corrective actions taken as necessary.

11.3.2.4 Functional Testing

The applicant’s description of provisions describing their functional testing program will be

applied to IROFS after initial installation, as part of periodic surveillance testing, and after the

performance of corrective maintenance or PM, or calibration to verify that IROFS are capable of

performing safety functions after these maintenance activities. The applicant’s overall functional

testing program will consist of a two primary sets of test programs; pre-operational testing

program, and an operational testing program. The Pre-operational testing program is comprised

of two subsets of tests which include functional tests and initial startup testing. Whereas, the

Operational testing program is comprised of two different subsets of tests, which include periodic

and special testing. The applicant describes the objectives of the preoperational and

operational testing programs are to provide assurance that IROFS meet applicable contractual,

regulatory, licensing, and design requirements, and can be relied on to perform design functions

required to protect the health and safety of workers and the public. Test procedures will be

sufficiently detailed to ensure that qualified personnel can perform the required functions without

direct supervision. Test procedure content will be consistent and will consist of the following

sections: (a) Title; (b) Purpose; (c) References; (d) Time Required; (e) Prerequisites; (f) Test

Equipment; (g) Limits and Precautions; (h) Required Plant Unit Status; (i) Prerequisite System

Conditions; (j) Test Method; (k) Data Required; (l) Acceptance Criteria; (m) Procedure; and (n)

Enclosures. Review of the above noted items (a through n) the applicant has provided broader

detail and elaboration on specific attributes for each test procedure section in the SAR (LES,

2005).

Preoperational Testing Program

The applicant describes their preoperational testing program primarily consisting of functional

testing and initial start-up testing. All preoperational functional tests associated with IROFS will

be completed before the introduction of licensed material at the facility. Before the start of preoperational functional testing, the facility will undergo a constructor turnover process, where

systems or portions of systems will be turned over to the applicant following as-built drawing

verification, purging, cleaning, and initial instrument calibration by the constructor. Functional

testing will begin as systems or portions of systems are turned over to the applicant. Preoperational functional testing at the facility will be conducted to initially determine various facility

parameters and to verify that IROFS are capable of performing their intended functions and that

facility SSCs are capable of meeting performance requirements. These tests are primarily

associated with IROFS (QA Level 1) and certain QA Level 2 SSCs. For SSCs that are not QA

Level 1, acceptance criteria will be established to ensure worker safety and reliable system
11-10

operation. Initial-start-up testing will involve the initial introduction of licensed material and all

subsequent testing through completion of Enrichment Setting Verification. Initial-start-up testing

begins with the initial introduction of licensed material and ends with the start of commercial

operation. Its purpose is to ensure safe introduction of licensed material and to verify

parameters assumed in the ISA. 

The applicant states that only properly reviewed and approved test procedures and plans will be

used for all pre-operational tests. The applicant ensures that copies of approved test procedures

will be made available to NRC staff approximately 60 days before implementation, and not less

than 60 days before the scheduled introduction of licensed material for initial-start-up testing. 

The pre-operational test plan will be made available to NRC at least 90 days before the start of

testing. Records of testing schedules and test results for all IROFS will be maintained for all

required pre-operational testing. The responsible department manager will review and approve

the results of each pre-operational test before the test process continues, where test results form

the basis for continuing the scheduled test program. Retesting will be conducted as necessary. 

All modifications to IROFS that the applicant considers necessary during the testing phase will

be evaluated in accordance with 10 CFR 70.72, and the impact of these modifications will be

evaluated during the 10 CFR 70.72 evaluation process. To ensure the adequacy of operating,

emergency, and surveillance procedures under actual or simulated operating conditions, and

before beginning actual plant operations, the applicant has agreed to the trial use of these

procedures throughout the testing program phases, as practicable. 

Operational Testing Program

The applicant’s operational testing program will consist of periodic testing and special testing. 

Periodic testing consists of testing conducted on a periodic basis, to provide continuing

verification that IROFS are capable of meeting performance requirements. Periodic testing also

includes surveillance tests and post-maintenance tests. Special testing is defined as testing that

is of a nonrecurring nature and does not fall under any of the other testing programs. The

Maintenance Manager will have the overall responsibility for the development and conduct of the

operational testing program. The operational testing program will use test coordinators for

various tests to be performed. The test coordinators are responsible for: (a) verification of all

testing prerequisites; (b) observance of all testing limits and precautions during the test 

evolution; (c) compliance with facility license requirements and other facility directives; (d)

identification and implementation of corrective actions necessary to resolve system conditions

observed during the test evolution; (e) verification of proper data acquisition and evaluation; (f)

ensuring the observance of personnel safety precautions during test evolutions; and (g)

coordinating additional test support from other organizations. Periodic and special testing

procedures will be sufficiently detailed such that qualified personnel can perform the required

functions without direct supervision. Procedures will be reviewed and approved before

implementation.

The applicant’s periodic testing program, will be a component of the operational testing program

and will be used to verify that the facility is capable of operating in a reliable manner and will

ensure that the health and safety of workers, the public, the environment, and compliance with all

regulatory requirements is achieved. The periodic testing program begins during the preoperational testing phase and continues throughout the life of the facility. A periodic testing

schedule will be established to ensure timely performance of testing requirements and evaluation

of testing results. The applicant describes the use of a computer database within the SAR (LES
11-11

2005) which will be used for scheduling and organizing surveillance test requirements,

procedures, and activities, and for ensuring that surveillance requirements are met and are within

the required test interval. The applicant’s database will include all periodic testing, calibrations,

or inspections required by regulations or license conditions. For each surveillance requirement,

the applicant’s database will provide information concerning the specific test to be performed,

test equipment, plant cascade, test frequency, and testing dates for the preceding and next

required surveillance test. Testing will be scheduled such that the safety of the plant is not

dependent upon the performance of IROFS that have not been tested within specified testing

intervals. If system or plant-unit conditions preclude the performance of a test within its required

interval, the applicant’s Health, Safety, and Environment (HS&E) Manager, Operations Manager,

and Maintenance Manager will be notified, in writing, by the responsible department, and the

responsible department will be required to ensure that the test will be performed as soon as

practical, once the required conditions are met. All periodic testing associated with QA Level 1

and QA Level 2 SSCs will be performed in accordance with approved written procedures.

The applicant’s special testing program encompasses all testing that is not pre-operational

testing, periodic testing, post-modification testing, or post-maintenance testing. Special testing

will be non-recurring testing conducted to determine facility parameters and/or to verify the

capability of IROFS to meet performance requirements. Special testing will be conducted for

such purposes as data acquisition and/or information-gathering, for special analyses or incident

investigations, verification of the effectiveness of corrective actions, and confirmation of results

expected from facility modifications, where such confirmatory testing falls outside the realm of

routine post-modification or post-maintenance testing. The applicant’s Plant Manager will have

the authority to determine when certain tests will be conducted as special tests. This

determination will be made based on evaluations that consider such factors as: (a) whether the

proposed testing evolution involves unusual operational configurations for which no prior

experience exists; (b) whether a potential exists for conducting an improper test that could have

an impact on primary plant parameters; and (c) whether the test activity involves seldomperformed evolutions, where any previous experience with such evolutions is no longer relevant

to the overall test activity.

11.3.3 Training and Qualifications

The applicant discusses training and qualification programs for the operations phase of the

facility in Section 11.3 of the SAR (LES, 2005). Operations phase training programs include

training for pre-operational functional testing and initial-start-up testing. The applicant’s purpose

and objective of the training program is to ensure that training is administered and job proficiency

is maintained for all facility personnel who perform activities relied on for safety. The applicant

states that qualification will be indicated by both successful completion of prescribed training,

demonstration of the ability to perform assigned tasks, and, where required by regulation,

maintaining an active and valid license issued by the regulatory agency. Training will involve

formal training to establish knowledge of the principles in the relevant job field and on-the-job

training to develop work performance skills. Continuing training or periodic retraining will be

provided to maintain the knowledge, skills, and proficiency of personnel in the relevant fields. 

The applicant’s QAPD, Appendix A, “Quality Assurance Training“ provides training and

qualification requirements for QA personnel during the design, construction, and operations

phases. The Quality Assurance Training detailed by the applicant in Appendix A of the QAPD

identifies QA Indoctrination Training provisions for personnel who perform QA Level 1 activities

which are applied exclusively to IROFS or any items which are determined to affect the function
11-12

of IROFS. The applicant’s QA Indoctrination and Training provisions include general criteria

inclusive of applicable codes, standards, QA implementing procedures, QA Program elements

and job responsibilities and authorities. In addition to this training, personnel who perform QA

Level 1 activities are required to complete training in job-specific QA activities prior to the

performance of QA Level 1 work. 

11.3.3.1 Organization and Management of the Training Function

In Section 11.3.1 of the SAR (LES, 2005), the applicant states that line management will be

responsible for the content and for conducting training of personnel. The applicant describes

their organization and management structure in detailed position descriptions in Chapter 2 of the

SAR (LES, 2005). The training organization has been delineated to provide support to line

management through planning, directing, analyzing, developing, conducting, evaluating, and

controlling of systematic performance-based training. For personnel performing activities relied

on for safety, the relevant indoctrination and training requirements are established through facility

administrative procedures. Exceptions from training requirements as stated by the applicant,

will require formal documented justification and approval by appropriate management. The

Human Resources Manager will be responsible for training programs at the facility. The

applicant will develop, implement and use lesson plans for all classroom, on-the-job training, and

maintenance of training records on each employee’s qualifications, and experience.

11.3.3.2 Analysis and Identification of Functional Areas Requiring Training

The applicant has described the methods that will be employed to perform an analysis of job

training needs and to ensure that personnel who work on tasks related to IROFS are provided

the appropriate training. Furthermore, the applicant states that employees will participate in Job

Hazard Analysis (also referred to as Job Safety Analysis), which will be used as part of on-thejob training, and will provide employees with the skills required to safely conduct job activities. 

The Job Hazard Analysis process will be used to enhance personnel recognition for defining the

work scope, identifying the hazards, mitigation of the hazard, completion of the task and

providing feedback. In developing the list of training tasks for specific job training activities, the

training organization will consult the appropriate technical and management personnel. During

the consultation process, the applicant will employ and list specific tasks pre-selected for training

which are reviewed and compared to training materials used to systematically evaluate training. 

The task list will be refined and updated when changes occur to procedures, processes, plant

systems, equipment or job scope.

11.3.3.3 Position Training Requirements

The applicant’s Position Training Requirements provide a description for developing minimum

training requirements for positions that involve activities which are relied on for safety or who

perform functions that prevent or mitigate an accident sequence described in the ISA Summary . 

The initial identification of job-specific training requirements will be based on experience, and the

level at which an employee initially enters the training program will be based on the employee’s

past experience, level of ability, and qualifications. 

The applicant has stated that facility personnel may receive training in the areas of General

Employee Training, Technical Training, and/or Employee Development/ManagementSupervisory training. The applicant’s General Employee Training will encompass: (a) QA
11-13

policies and procedures, (b) nuclear safety (inclusive of dosimetry, protective clothing and

equipment), (c) industrial safety, (d) health and first aid, (e) facility security, (f) emergency

planning and implementing procedures, (g) administrative controls and procedure use, (h) facility

systems and equipment, (i) chemical safety, (j) fire protection and fire brigade, (k) new employee

orientation, and (l) applicable regulations. General Employee Training will be required for all

personnel and contractors who report to facility management. Technical training will help facility

employees gain a technical understanding of the fundamental concepts, practices, and

procedures common to a gas centrifuge uranium enrichment facility and will develop the skills

necessary to competently perform assigned work. Technical training will consist of the following: 

(a) Initial Training for providing an understanding of fundamental principles, concepts, and

procedures; (b) On-the-Job Training and Qualifications for providing the necessary job-related

skills for the position; (c) Continuing Training for maintaining and improving job-related

knowledge and skills; and (d) Special Training, which involves subjects, required for a specific

job, that are unique in nature.

Nuclear Safety Training

The applicant ‘s Nuclear Safety Training Programs will be broad based and encompass various

types of job functions such as, (production operator, radiation protection technician and

contractor personnel) which will be used in conjunction with criticality safety and radiation safety

responsibilities associated with each position. The applicant describes their nuclear safety

training program as being structured in a manner that emphasizes the high level of importance

placed on radiological, criticality and chemical safety of plant personnel and the public. Nuclear

Safety Training will be structured as follows:

Personnel access procedures are developed to ensure the completion of formal nuclear safety

training prior to permitting unescorted access into Controlled Access Areas. Comprehensive

training sessions will be administered covering such topics as criticality safety, radiation

protection, and emergency procedures. The applicant’s training sessions include the following

course topics and subject matter: 

a. Notices, reports and instructions to workers,

b. Practices designed to keep radiation exposures ALARA

c. Methods of controlling radiation exposures

d. Contamination control methods ( including decontamination)

e. Use of monitoring equipment

f. Emergency procedures and actions

g. Nature and sources of radiation

h. Safe use of chemicals

i. Biological effects of radiation

j. Use of personnel monitoring devices

k. Principles of nuclear criticality safety

l. Risk to pregnant females

m. Radiation protection practices 

n. Protective clothing (donning and doffing practices)

o. Respiratory protection

p. Personnel surveys
11-14

The applicant’s Criticality Safety Training is described as being developed and implemented in

accordance with ANSI/ANS 8.19, “Administrative Practices for Nuclear Criticality” (ANSI/ANS,

1996) and ANSI/ANS 8.20, “Nuclear Criticality Safety Training” (ANSI/ANS, 1991). The

applicant’s nuclear safety training program will be evaluated and updated annually through

roundtable safety meetings held by supervisors and managers. 

Fire Brigade Training

The applicant’s descriptive scope and intent of Fire Brigade Training is to develop a core group

of facility personnel skilled at fire prevention, fire fighting techniques, first aid procedures, and

emergency response. Fire Brigade Training Program requirements will be used for initial

training of all new fire brigade members, semi-annual class room training and emergency

response drills, annual classroom training and drills, annual practical training, and leadership

training for fire brigade leaders. 

Technical Training

Technical Training as described by the applicant will be designed, developed and implemented

to provide assistance to facility employees in retaining a comprehensive knowledge base of the

fundamentals, procedures and practices employed in a gas centrifuge uranium enrichment

facility. Technical training will be also used to develop manipulative skills that are required to

demonstrate competency. Technical training will be divided into four segments: Initial Training,

On-the-Job Training and Qualifications, Continuing Training, and Special Training. 

Initial Training

The applicant describes initial training and qualification programs as being used to provide an

understanding of the fundamentals, basic principles, and procedures used to control work

evolution or activity. Initial training will be administered in various formats, which include live

lectures, taped and filmed lectures, self-guided study, demonstrations, laboratories and

workshops, and on-the-job-training. Initial job training and qualification programs will be

developed for operations, maintenance and technical services classifications. Logical

sequencing or block modules will be presented in a objective manner that human behavioral

objectives will be identified. The applicant has identified initial training programs which will be

inclusive of; Operations Initial Training, Mechanical Maintenance Initial Training, Instrumentation

and Electrical and Maintenance Initial Training, Health Physics and Chemistry Initial Training,

and Engineer/Professional Initial Training 

Operations Initial Training (General/Specific Systems, Nuclear Preparatory, Plant Familiarization)

General Systems training, as described by the applicant, is comprised of training and instruction

that will be developed to provide a trainee with basic knowledge on concepts and fundamentals

of mathematics, physics, chemistry, heat transfer, and electrical theory. The applicant has

augmented this training with on-the-job-training associated with detailed systems and component

operation supplemented by elementary process instrumentation and control.

Specific Systems training is described by the applicant as providing basic instruction in system

and component identification and basic system operating characteristics. This training module
11-15

will provide a general overview of enrichment plant equipment, nomenclature, and instruction

describing basic system operations.

Nuclear Preparatory training will be provided by the applicant to operations personnel preceding

completion of Specific Systems training. Training objectives described in this training module

include basic nuclear physics, plant chemistry, thermodynamics, radiation protection, and

enrichment theory.

Plant Familiarization will be provided by the applicant as orientation of employees to plant layout,

systems, and practical laboratory and equipment work at the facility.

 

Mechanical Maintenance Initial Training (General Systems, Fundamental Shop Skills, Plant

Familiarization)

General Systems Training is the same course module as described in Operations Initial Training

and will also be provided to Mechanical Maintenance personnel during initial training.

Fundamental Shop Skills as described by the applicant will provide instruction in fundamental

mechanical maintenance performance characteristics. This training module will emphasize

academic instruction combined with hands-on-training familiarizing trainees with design

operational and physical characteristics of enrichment facility components and the basic skill sets

used to perform mechanical repairs or equipment replacement. Factors applied by the applicant

include task training lists, which will be integrated into the module to provide assurance that each

prospective trainee attains a minimum level of performance. Practical tasks will be assigned to

trainees and correlated to the use of work procedures as a guide to task completion with

emphasis placed on radiological and industrial safety.

Plant Familiarization Training is the same course as previously noted, the exception as noted by

the applicant is that this training module will also be provided to Mechanical Maintenance

personnel during initial training. 

Instrumentation and Electrical and Maintenance Initial Training (General Systems, Basic

Instrument and Electrical, Basic Performance, Plant Familiarization)

General Systems Training is the same course module as described in Operations Initial Training

and will also be provided to Instrumentation and Electrical and Maintenance personnel during

initial training. 

Basic Instrument and Electrical training is described by the applicant as providing prospective

trainees with refresher training in electrical and electronic fundamentals, digital techniques and

application, instrumentation and control theory and application and instruction in measuring and

test equipment use. The applicant describes this training module as providing a general

overview and working knowledge of nuclear and non-nuclear instrumentation systems, overall

integrated plant operation and control, and the hazards of calibration errors and calibration

during plant operation.

Basic Performance Training as explained by the applicant, describes fundamental performance

familiarization and instruction provided to trainees regarding plant test procedures, test

equipment, testing, plant records and reports, as well as data collection during operations. 
11-16

Training will be provided by the applicant in the basic application and use of thermodynamics

utilized in plant heat transfer.

Plant Familiarization Training is the same course as previously noted, the exception as noted by

the applicant is that this training module will also be provided to Instrumentation and Electrical

Maintenance personnel during initial training. 

Health Physics and Chemistry Initial Training (General Systems, Fundamental Health Physics,

Fundamental Chemistry, Plant Familiarization)

General Systems Training is the same course module as described in Operations Initial Training

and will also be provided to Health Physics and Chemistry personnel during initial training. 

Fundamental Health Physics training is described by the applicant as providing prospective

trainees with comprehensive and theoretical understanding of nuclear processes. Training

objectives include practical presentations demonstrating the proper techniques and application of

theory in the use of non-automated counting and spectrographic equipment and use of portable

survey instruments . The applicant describes this training module as providing a detailed

overview of administrative materials.

Fundamental Chemistry training is provided by the applicant to familiarize trainees with training

pertaining to chemistry, chemistry theory, techniques, and procedures. The applicant’s

emphasis will be placed on familiarization of trainees with chemical safety hazards in the

laboratory and during plant operation.

Plant Familiarization Training is the same course as previously noted, the exception as noted by

the applicant is that this training module will also be provided to Health Physics and Chemistry

personnel during initial training. 

Engineer/Professional Initial Training (Facility Orientation, Basic Engineer/Professional Training,

Enrichment/Chemical Engineer/ Professional Training, Engineer/ Professional Systems Training) 

The applicant’s Engineer Professional training is for the applicant’s technical staff and managers. 

Professional Engineer training is described as follows:

Facility Orientation training is provided by the applicant as an orientation to each section of the

enrichment facility. Training will be provided based upon on-the-job-task lists, which detail

training objectives that will be accomplished while working in the section. 

Basic Engineer/Professional Training is described by the applicant as providing prospective

trainees with a basic understanding of how uranium is enriched. Training objectives will include

practical presentations demonstrating the systems and components required for producing the

final product and the interrelationship and interface requirements with the various facility

organizations used to achieve facility objectives.

Enrichment/Chemical Engineer/ Professional Training is described by the applicant as providing

chemical engineer’s with comprehensive and theoretical understanding of enrichment

operations. Training objectives which will be addressed specifically by the applicant include;

thermal science, nuclear physics, and application of theory in an enrichment facility.
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Engineer/ Professional Systems Training will be provided by the applicant as an overview of

plant systems, components and procedures necessary to operate an enrichment facility safely

and efficiently.

On-the Job Training and Qualifications

On the Job Training and Qualifications has been described by the applicant in a systematic

manner which takes into consideration job related skills and knowledge of the position. The

foundation of the OJT/qualifications program will comprise applicable tasks and related

procedures for each respective technical area which the applicant states will be designed to

supplement and compliment training received through formal classroom, laboratory and

simulator training. The applicant concludes by stating that the objective of the OJT Training

program is to assure the trainee’s ability to perform job specific tasks as described in detailed

task descriptions and Training and Qualification Guides.

Continuing Training

Continuing Education is defined and described by the applicant as any training that is not

provided as initial qualification and basic training which maintains and improves job-related

knowledge and skills. Continuing training will be used by the applicant to supplement and

provide additional training when warranted. Categorization of continuing training will be as

follows:

a. Facility systems and component changes;

b. OJT/Qualification program testing;

c. Policy and procedure changes;

d. Operating experience program documents review to include Industry and in-house

operating experiences;

e. Continuing training required by regulation (e.g., emergency plan training);

f. General employee, special, administrative, vendor, and/or advanced training topics

supporting tasks that are elective in nature;

g. Training identified to resolve deficiencies (task-based) or reinforce seldom used

knowledge skills;

h. Refresher training on initial training topics;

i. Structured pre-job instruction, mock-up training, walk-throughs; and

j. Quality awareness.

Components of the applicant’s continuing training process is comprised of both formal and

informal delivery systems which will be administered at frequencies based on the need to

maintain on the job proficiency. A systematic approach will be developed by department or
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section and results from data gathered from job performance and safe facility operation will be

used for determining the content and context of continuing training. 

Special Training

Special Training has been described and defined by the applicant as involving training subjects

of unique nature required for a particular area of work. The applicant only administers special

training to select personnel based on specific needs not directly related to discipline specific

qualifications.

11.3.3.4 Basis for Training and Training Objectives

The applicant’s Basis for Training and Training Objectives program content will be established

based on an analysis of job training needs and position-specific requirements. Overall training

objectives will include the demonstration of knowledge, skills, and abilities, as well as the

achievement of performance standards. The applicant states that the basis for safety training

will be derived from compliance with applicable sections of the Occupational Safety and Health

Administration (OSHA) regulation 29 CFR 1910 (Occupational Safety and Health Standards),

1910.1200 (Hazard Communication), 10 CFR Part 20 (Standards for Protection Against

Radiation) and 10 CFR 19 (Notices, Instructions, and Reports to Worker: Inspection and

Investigations). The applicant will provide continuing training in these areas to maintain

employee proficiency. Training basis and the topics covered will be provided to all persons 

under the direct supervision of facility management (including contractors) and will be required to

participate in General Employee Training. Provisions have been stipulated for certain facility

support personnel which is dependent upon normal work assignment, which will exclude

participation in all topics of some training. Temporary maintenance and service personnel

receive General Employee Training to the extent necessary to assure safe execution of their

assigned work duties. In some cases, the applicant may include portions of General Employee

Training with New Employee Orientation.

11.3.3.5 Lesson Plans and Other Training Guides

Lesson plans will be used for classroom training and on-the-job training, as required, and include

trainee performance standards. Lesson plans will be developed from the learning objectives,

which will be based on job performance standards. Lesson plans and training guides will be

developed and reviewed by the training function organization, as well as the organization that is

cognizant in the subject matter. 

11.3.3.6 Evaluation of Trainee Accomplishment

The applicant states that observation/demonstration or oral or written tests will be used, as

appropriate, to evaluate and measure a trainee’s understanding and command of learning

objectives and knowledge and skill sets acquired for job performance. Evaluations will be

performed by individuals qualified in the training subject matter. 

11.3.3.7 Conduct of On-the-Job Training

On-the-job training will be used in conjunction with classroom training for activities identified as

IROFS. Successful completion of on-the-job training requirements involves the performance of
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either the actual task or a simulation of the task. On-the-job training using task simulation will be

carried out with the trainee explaining task actions using conditions encountered during the

actual performance of the task.

11.3.3.8 Evaluation of Training Effectiveness

The applicant will periodically conduct systematic evaluations of the training program as a means

to measure the program’s effectiveness in ensuring the job proficiency and competency of all

facility personnel is maintained. These evaluations will identify strengths and weaknesses in the

training program and will determine whether the program meets current job needs. The need for

any corrective actions will be determined. The applicant has outlined specific methods that may

be used to collect training program evaluation data, including, among other things, testing to

determine trainee accomplishment of objectives, trainee evaluation of the instruction, supervisory

evaluation of job performance after training, and supervisory evaluation of the instruction. The

QA Department will audit the facility training and qualification program. QA assessments of

training program or topical areas may include the following elements:

a. Management and administration of training and qualification programs;

b. Development and qualification of the training staff;

c. Position training requirements;

d. Determination of training program content, including facility change control interface with

configuration management systems;

e. Design and development of training programs;

f. Conduct of training;

g. Trainee examinations and evaluations; and

h. Training program assessments and evaluations.

The applicant describes the evaluation process and documentation of the results with emphasis

placed on highlighted program strengths and weakness. Surveys, questionnaires, performance

appraisals, staff evaluation, and overall effectiveness evaluations will be used as instruments for

verification of implementation of the program.

11.3.3.9 Personnel Qualification

The applicant states that qualification and training requirements for process operator candidates

will be established and implemented in plant procedures. The applicant describes the

qualification requirements for key management positions in Chapter 2 of the SAR (LES, 2005). 

QA personnel also have specific training and qualification requirements, which are specified in

the applicant’s QAPD provided in Appendix A to Chapter 11 of the SAR (LES, 2005).
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11.3.3.10 Periodic Personnel Evaluations

The applicant states that personnel who perform activities related to items relied on for safety

(IROFS) will be evaluated at least biennially, to determine whether they are competent to

continue performing these activities. The applicant’s evaluation process will use written tests,

oral tests, or on the job performance appraisals to conduct and document periodic personnel

evaluations. Retraining will be provided when the results of periodic evaluation dictate the need. 

Retraining will also be administered when facility modification, procedure changes, and QA

Program changes, result in new or revised information. 

11.3.4 Procedure Development and Implementation

The applicant discusses it’s programs for the development and implementation of procedures in

Section 11.4 of the SAR (LES, 2005). The applicant states that all activities involving licensed

materials or IROFS will be conducted in accordance with approved procedures, which will be

used to control activities to ensure that they are conducted in a safe manner and in accordance

with regulatory requirements. Procedures will be made available to NRC for inspection before

the start of enrichment operations at the facility.

The applicant identifies four types of plant procedures that will be used to control activities at the

facility: (1) operating procedures, (2) administrative procedures, (3) maintenance procedures,

and (4) emergency procedures. Operating procedures will be used to directly control process

operations at the facility. The applicant will implement a methodology for identifying, developing,

approving, implementing, and controlling operating procedures. The identification of needed

procedures will take into consideration the results of the ISA. Operating procedures will include,

among other things: (a) the purpose of the activity; (b) governing regulations and policies; (c)

steps for each operating process phase; (d) hazards and safety considerations; (e) operating

limits; (f) precautions to prevent exposures; (g) measures to be taken in case of exposure; (h)

IROFS associated with the process; and (i) the time frame for which the procedure is valid. The

applicant states that applicable safety limits and IROFS will be clearly identified in the operating

procedures.

Administrative procedures will be used to perform activities that support process operations. 

These procedures will control activities that include management measures such as: (a) CM; (b)

nuclear criticality safety, radiation safety, chemical safety, and fire safety; (c) QA; (d) design

control; (e) training and qualification; (f) audits and assessments; (g) incident investigations; (h)

record-keeping and document control; (i) reporting; and (j) procurement. Administrative

procedures will also be used for implementing facility security and emergency plans, including

the Fundamental Nuclear Material Control Plan, the Emergency Plan, the Physical Security Plan,

and the Standard Practice Procedures Plan for Protection of Classified Information. 

Maintenance procedures will address: preventive and corrective maintenance of IROFS;

surveillance; functional testing of IROFS; and pre-maintenance activity requirements. 

Emergency procedures will address the preplanned actions of operators and other plant

personnel, in case of an emergency.

The applicant has described a process for creating, reviewing, and approving procedures. 

Procedure development will be carried out by the appropriate members of the facility staff or

facility contractors. Initial “Draft Procedures” will be reviewed by other appropriate members of
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the facility staff, by personnel from the centrifuge supplier, and other vendors, as warranted. 

These reviews will verify the technical accuracy of the procedure and will include either a walkdown of the procedure in the field or a table-top walk-through. An independent review will be

carried out for procedures that will be written for the operation of IROFS. The Plant Manager or

designee will approve all procedures. If the procedure directly involves elements of QA, the QA

Manager shall approve the procedure. The need for cross-disciplinary reviews will be

considered during the procedure review process. The designated approving official will be

responsible for determining the need for cross-disciplinary reviews.

The applicant states that procedures will be subject to re-verification and re-validation through

periodic reviews, as a means to ensure continued accuracy and usefulness. Periodic reviews of

operating procedures will be conducted at a minimum frequency of once every 5 years. Periodic

reviews of radiation protection procedures and emergency procedures will be conducted at a

minimum of once every year. Furthermore, in response to unusual incidents, including

accidents, unexpected transients, operator errors, or equipment malfunction, the affected

procedures will be reviewed. The applicable procedures will be reviewed after any modification

to a system.

The applicant has defined a set of guidelines that describes how changes to procedures will be

processed. The process for changes to procedures will first involve documentation of the

proposed change, as well as the reason for the change. An evaluation of the proposed change

will be performed in accordance with 10 CFR 70.72, as appropriate. If after the applicant’s

evaluation and determination is made that an amendment to the license is required to implement

the proposed changes, the change will not be implemented until prior approval is received from

NRC. The revised procedure, with the proposed changes, will then be reviewed by a qualified

reviewer. The Plant Manager, a Department Manager, or a designee approved by the Plant

Manager will be responsible for approving procedure changes, and for determining whether a

cross-disciplinary review is necessary. The need for cross-disciplinary reviews will, as a

minimum, be considered for proposed changes having a potential impact on chemical safety,

radiation safety, and nuclear criticality safety, as well as material control and accounting.

The applicant states that the Document Control organization will distribute originally issued

approved procedures and approved procedure revisions in a controlled manner. An index of the

distribution of copies of all facility procedures will be established and maintained by document

control. Revisions to procedures will be controlled and distributed in accordance with this index. 

Indexes will be reviewed and updated on a periodic basis or as required. Department managers

or their designees will be responsible for ensuring that all personnel carrying out work activities,

which require the use of procedures and have ready access to controlled copies of the

procedures. On-the-job training and continuing training will be used to ensure that personnel are

qualified to use the latest procedures. Both these training categories are components of the

applicant’s technical training program.

11.3.5 Audits and Assessments

In Section 11.5 of the SAR (LES, 2005), the applicant is required to implement a system of audits

and assessments to help ensure that facility activities are conducted in a safe manner and for

verifying compliance with regulatory requirements, licensing commitments, and procedures. The

applicant describes a system of audits and assessments that consists of two distinct levels of

activities: an audit activity structured to monitor compliance with regulatory requirements,
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licensing commitments, and facility procedures; and an assessment activity oriented toward

determining the effectiveness of the activities in ensuring that IROFS are reliable and available to

perform their intended safety functions. The applicant’s approach to implementing the system of

audits and assessments will involve performing audits and assessments on critical work activities

associated with facility safety, environmental protection, and other areas.

Audits and assessments will be performed to assure that facility activities are conducted in

compliance with regulatory requirements, licensing commitments, and written procedures, and

that the processes and activities reviewed, effectively meet the applicant’s specified objectives

for ensuring the continued availability and reliability of IROFS. The applicant will implement a set

of functions for which audits and assessments will be conducted. Audits and assessments will

be conducted for the following areas: (a) nuclear criticality safety, radiation safety, and chemical

safety; (b) industrial safety and fire protection; (c) environmental protection; (d) emergency

management; (e) QA; (f) CM; (g) maintenance; (h) training and qualification; (i) procedures; (j)

CAP and incident investigation; and (l) records management. The applicant has also specified

actions that will be carried out in performing typical audits and assessments. 

These specified actions include: (a) using approved audit and assessment checklists; (b)

interviewing responsible personnel; (c) performing plant area walk-downs; (d) reviewing

controlling plans and procedures; (e) observing work in progress; and (f) reviewing completed

QA documentation.

The frequency of audits and assessments will be based on the status, safety significance, and

history of the activities being performed. All major activities will be audited or assessed annually.

Nuclear criticality safety audits will be conducted and documented quarterly, such that all aspects

of the Nuclear Criticality Safety Program will be audited at least every 2 years. The operations

organization will be assessed periodically to ensure that nuclear criticality safety procedures are

being followed and the process conditions have not been altered to adversely affect nuclear

criticality safety. Assessments will be conducted at least semiannually. A schedule will be

established that identifies the audits and assessments to be performed and the responsible

organization assigned to conduct the activity. This schedule will be reviewed periodically and

revised as necessary to ensure the appropriate coverage, based on current and planned

activities.

The applicant states that audits and assessments will be performed routinely by qualified staff

who are not directly responsible for production activities. The QA department, under the

direction of the QA Manager, will be responsible for conducting audits. The conduct of the audit

will be carried out by an audit team, working under the direction of a lead auditor. The audit

team will consist of one or more certified auditors. Audit team members will not have direct

responsibility for the function and area being audited; however, audit team members will have

technical expertise or experience in the area being audited. Auditors and lead auditors will be

responsible for ensuring the performance of audits in accordance with applicable QA procedures

and will hold certifications as required by the QA Program. The auditor certification process

involves training in the following areas: (a) QA Program; (b) audit fundamentals, including audit

scheduling, planning, performance, reporting, and follow-up action involved in conducting audits;

(c) objectives and techniques of performing audits; and (d) on-the-job training. 

Certification of auditors is based on the evaluation of the QA Director or Manager. This

evaluation will take into consideration, among other things, such factors as: (a) education; (b)
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experience; (c) completion of the required training; (d) professional qualifications; (e) leadership;

(f) sound judgment; (g) maturity; (h) analytical ability; and (I) past performance. A lead auditor

must have participated in a minimum of five QA audits or audit equivalents within a period of time

not to exceed 3 years before the date of certification. 

Assessments will fall into two categories, which will both be owned and managed by line

organizations. Management assessments will be conducted by line organizations that are

responsible for the work activity. Independent assessments will be conducted by individuals not

involved in the area being assessed. Personnel performing assessments will not require

certification; however, they will be required to complete QA orientation training, as well as

training on the assessment process. Nuclear criticality safety assessments will be performed

under the direction of the criticality safety staff. The personnel who perform these assessments

will not report to the production organization, and will have no direct responsibility for the function

and/or area being assessed.

The applicant will use approved procedures for conducting internal and external audits and

assessments. These procedures will meet the QA Program requirements and will provide for the

following audit and assessment activities: (a) scheduling and planning of audits and

assessments; (b) certification requirements for audit personnel; (c) development of audit plans

and audit and assessment checklists, as applicable; (d) audit and assessment performance; (e)

reporting and tracking of findings to closure; and (f) closure of audits and assessments.

The applicant will implement a system that captures the planned use of the results of audit and

assessment activities, including the planning and implementation of corrective actions, based on

audit and assessment findings and recommendations. 

11.3.6 Incident Investigations

Section 11.6 of the SAR (LES, 2005) describes the applicant’s incident investigation and

corrective action process for investigating abnormal events and completing the appropriate

corrective actions. The applicant will implement an incident investigation process for reporting

deficiencies, abnormal events, and potentially unsafe conditions or activities. The applicant’s

overall incident investigation process will provide for incident identification, investigation, rootcause analysis, environmental protection analysis, recording, reporting, and follow-up. These

activities will be performed according to written corrective action process procedures. Each

event or condition will be evaluated to determine the level of investigation required. Guidance for

evaluating the significance of occurrences will be contained in corrective action process

procedures, and the extent of the investigation will depend on the significance of the incident,

with respect to the levels of uranium released and/or the potential for exposure to workers, the

public, or the environment. 

The investigation process for abnormal events will implement the requirements for event

notification to government agencies in accordance with regulations. The HS&E Manager will be

responsible for maintaining a list of agencies to be notified, determining if a report to an agency

is required, and notifying the agency when required. The applicant has described their process

for reporting events to the NRC, as required by 10 CFR 70.50 and 70.74. Abnormal events that

have the potential for actual HS&E consequences will be appropriately identified and reported to

the HS&E Manager, who will be responsible for incident investigations into such events. The

HS&E Manager or designee will be responsible for maintaining a record of corrective actions to
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be implemented as a result of incident investigations, in accordance with corrective action

process procedures and for tracking corrective actions to completion. The applicant stated that it

will monitor and document corrective actions through completion.

The applicant has described several specific features of its incident investigation process. These

features include a commitment to establish processes to determine the specific or generic root

cause(s) of abnormal events, the generic implications of such events, the recommendation of

corrective actions, and reporting events to NRC, as required by 10 CFR 70.50 and 70.74. The

investigation process will include a prompt risk-based evaluation of the event, to determine it’s

risk significance. Investigations will begin within 48 hours of the abnormal event, or sooner,

depending on the event’s safety significance.

The number of investigators for a given event will vary according to the complexity and severity

of the event. Investigators will be independent from the line function(s) involved with the incident

under investigation and investigators have been assured of no retaliation for participating in the

investigations. Qualified internal or external investigators will be appointed to serve on

investigating teams when required. The investigating teams will include at least one process

expert and at least one team member trained in root-cause analysis.

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 other

operations at the facility. For each abnormal event, an incident report will be written which

includes: a description of the event; contributing factors; a root-cause analysis; findings; and

recommendations. Relevant findings will be reviewed with all affected personnel. Details of the

event sequence will be compared with accident sequences already considered in the ISA, and

the ISA Summary will be modified to include evaluation of the risk associated with accidents of

the type actually experienced.

The applicant will develop a corrective action process and implementing procedures for

conducting incident investigations. These procedures will include the following elements: (a) a

documented plan for investigating an abnormal event; (b) a description of functions,

qualifications, and responsibilities of the investigative team manager and team members; (c) the

scope of the investigative team’s authorities and responsibilities; (d) assurance of management

cooperation with the investigation effort; (e) assurance of the team’s authority to obtain all the

necessary information; (f) assurance of the investigative team’s independence from the

functional area involved in the incident being investigated; (g) retention of documentation relating

to abnormal events for 2 years or for the life of the operation, whichever is longer; (h) guidance

for investigative team members on how to apply a reasonable, systematic, structured approach

to determining the root cause(s) and generic implications of the event; (i) requirements to make

original investigation reports available to NRC, on request; and (j) a system for tracking the

completion of appropriate corrective actions.

The applicant discusses a corrective action process where individuals involved in quality

activities are responsible for identification and documentation of conditions adverse to quality, as

well as, analysis of these conditions to determine the nature of corrective actions and the

implementation of corrective actions in accordance with corrective action process procedures. 

The QA Program contains the corrective action process procedures for identifying, reporting,

resolving, documenting, and analyzing conditions adverse to quality. Conditions adverse to

quality may involve failures, malfunctions, deficiencies, deviations, defective materials and
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equipment, and nonconformances. Reports of conditions adverse to quality will be analyzed to

identify trends in quality performance. Significant conditions and significant trends adverse to

quality, along with root-cause analyses and corrective actions, will be documented and reported

to senior management for review and assessment, in accordance with corrective action process

procedures. The QA Manager will take follow-up action to verify proper and timely

implementation of corrective action. The QA Program will require regularly scheduled audits and

assessments of the corrective action process, to ensure that the needed corrective actions are

identified. 

All employees will have the authority and responsibility to initiate the corrective action process if

they discover deficiencies.

11.3.7 Records Management

The applicant’s records management function is described in Section 11.7 of the SAR (LES,

2005). Procedures for reviewing, approving, handling, identifying, retaining, distributing,

retrieving, controlling, and maintaining QA records are required by the applicant’s QA Program. 

QA records will include all completed records that furnish documentary evidence of the quality of

items and/or activities affecting quality, including records related to health and safety. Records

related to health and safety will be maintained in accordance with the requirements of Title 10,

Code of Federal Regulations. The applicant has specified examples of the types of records that

will be retained under records management procedures, including: (a) operating logs; (b)

procedures; (c) supplier QA documentation; (d) nonconforming item reports; (e) test

documentation/test results; (f) facility-modification records; (g) drawings; (h) specifications; (i)

procurement documents; (j) nuclear material control and accounting reports;(k) maintenance

activities including calibration records; (l) inspection documentation; (m) audit reports; (n)

reportable occurrences and compliance records; (o) completed work orders; (p) license

conditions records; (q) software verification records; (r) system descriptions; (s) as-built designdocumentation packages; (t) regulatory reports; (u) corrective action documents; and (v) any

other QA documentation required by specifications or procedures. These and other such

records will be retained for at least the periods specified in the records management procedures. 

Other retention times will be specified for other facility records, as necessary, to meet applicable

regulatory requirements. 

That applicant states that for computer codes and computerized data used for activities relied on

for safety, as specified in the ISA Summary, procedures will be established for maintaining the

readability and usability of older codes and data as computing technology changes.

The applicant describes a Master File that will be maintained by the facility for storage and

categorization of records and documents that fall within the scope of the applicant’s records

management function. The Master File storage system will provide for efficient and accurate

retrieval of information. Access to and use of the Master File will be controlled by the facility. 

Documents in the Master File will be legible and identifiable by subject, and they will be

considered valid only if stamped, initialed, signed or otherwise authenticated and dated by

authorized personnel. Documents in the Master File may be originals or reproduced copies. 

Records stored in the Master File will be firmly attached in binders or placed in folders or

envelopes; records will not be stored loosely. All records should be stored in steel file cabinets. 
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Computer storage of records will be permitted and will be done in a manner to preclude

inadvertent loss and to ensure accurate and timely retrieval of data. Dual facility records storage

will use an electronic data management system and storage of backup tapes in a fireproof safe. 

Records that are light-sensitive, pressure-sensitive, or temperature sensitive will be packaged

and stored as recommended by the manufacturer of these media. Examples of such records

include radiographs, photographs, negatives, and microfilm.

Written instructions will be prepared concerning the storage of records in the Master File, and the

responsibility for implementing the requirements of the instructions will reside with a supervisor. 

These instructions will include, but will not necessarily be limited to, the following elements:

a. description and identification of the locations of the Master File and the various record

types within the Master File;

b. the filing system to be used;

c. a method for verifying that records received are in agreement with any applicable

transmittal documents and are in good condition;

d. a method for maintaining a record of the documents received;

e. the criteria governing access to and control of the Master File;

f. a method for maintaining control of and accountability for records removed from the

Master File; and

g. a method for filing supplemental information and for disposing of superseded records.

A qualified Fire Protection Engineer will evaluate record storage areas to assure records are

adequately protected from damage.

11.3.8 Other QA Elements

The applicant addresses other QA elements that will be applied to IROFS and other

management measures in Section 11.8 of the SAR (LES, 2005). The applicant has included a

complete description of its application of QA elements to IROFS. The applicant’s QA Program

specifies mandatory requirements for performing activities affecting quality. These requirements

are implemented through procedures. QA procedures and revisions to QA procedures are

distributed on a controlled basis to organizations and individuals responsible for quality.

Before undertaking an activity, the applicant will establish implementation programs and

procedures to document, approve, and implement the applicable portions of the QA Program. 

The applicant states that a management assessment of the QA Program will be performed at

least 6 months before the scheduled receipt of licensed material at the facility. QA Program

items identified as needing completion or modification as a result of this assessment will be

entered into the applicant’s corrective action process, and corrective action will be completed on

such items before the scheduled receipt of licensed material at the facility. The QA Program will

be monitored by management before the initial management assessment, through project review
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meetings and annual assessments. These management oversight and assessment activities will

ensure that the QA Program is in place and effective before the receipt of licensed material.

The applicant states that the QA Program and supporting manuals and procedures will be

reviewed periodically to ensure they are in compliance with applicable regulations, codes, and

standards. New or revised regulations, codes, and standards will be incorporated into the QA

Program and supporting manuals and procedures, as necessary. The applicant’s President will

assess the scope, status, adequacy, and regulatory compliance of the QA Program through

regular meetings and correspondence with the Plant Manager and the QA organization. The QA

Director will periodically inform the applicant’s President and Plant Manager of quality concerns

that require management resolution.

Personnel performing activities covered by the QA Program will perform work in accordance with

approved procedures and will demonstrate suitable proficiency in their assigned tasks. The

applicant commits to establishing formal training programs for QA policies, requirements,

procedures, and methods. Ongoing training will be provided to ensure continuing proficiency as

procedural requirements change. New employees will be required to attend an introductory QA

class that covers QA authority, organization, policies, manuals, and procedures. Additional

formal training will be conducted in specific topics such as NRC regulations and guidance,

procedures, auditing, and applicable codes and standards. Supplemental training will be

performed as required. On-the-job training will be performed by the employee’s supervisor in QA

area-specific procedures and requirements. Training records will be maintained for each person

performing quality-related job functions.

The applicant states that the QA Program for design, construction, and pre-operational testing

will continue simultaneously with the QA Program for the operations phase while construction

activities are in progress. The applicant will develop and implement a system to plan and

schedule system turnover as construction is completed. Before system turnover, written

procedures will be developed for controlling the transfer of SSCs and associated documentation. 

These procedures will include checklists, marked drawings, documentation lists, system status,

and receipt control. Major work activities performed by the applicant’s contractors will be

identified and controlled. Principal contractors will be required to comply with the applicable

portions of 10 CFR Part 50, Appendix B, “Quality Assurance Criteria for Nuclear Power Plants

and Nuclear Fuel Reprocessing Plants.” The applicant will formally evaluate contractor

performance, in accordance with the safety importance of the contracted activities.

The applicant will assign QA levels to facility SSCs and associated processes, based on their

safety significance. Each component and document will receive a categorization of QA Levels 1,

2, or 3. The applicant’s QA Program and it’s supporting manuals, procedures, and instructions

are applicable to items and activities designated as QA Levels 1 and 2. The QA-Level

assignment will apply throughout the life of the facility, and will be based on the following

definitions:

QA Level 1 Requirements

The applicant states that it’s QA Level 1 Program will conform to the criteria established in 10

CFR Part 50, Appendix B. These criteria will be met by mandates to follow the guidelines of

American Society of Mechanical Engineers (ASME) NQA-1-1994 (ASME, 1994), including

supplements as revised by the ASME NQA-1a-1995 Addenda (ASME, 1995), as specified in the
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applicant’s QAPD, which is provided in Appendix A to the SAR (LES, 2005). The QA Level 1

Program will be applied to those SSCs and administrative controls that have been determined to

be IROFS; items that affect the functions of IROFS; and items required to satisfy those

regulatory requirements applicable to the QA Level 1 Program.

QA Level 2 Requirements

The applicant’s QA Level 2 Program is an owner-defined QA Program that uses the ASME NQA1-1994 standard (ASME, 1994), including supplements as revised by the ASME NQA-1a-1995

Addenda (ASME, 1995), as guidance. General QA Level 2 requirements are described in

Section 2.0, “Quality Assurance Program for QA Level 2 Activities,” in the applicant’s QAPD,

provided in Appendix A, to the SAR (LES, 2005) “Identification and Application of QA Controls.” 

For contractors, the QA Level 2 Program will be described in documents that the applicant will

approve. The QA Level 2 Program will be applied to owner-designated SSCs and activities. An

International Organization for Standardization 9000 series QA Program may be acceptable for

QA Level 2 applications provided it complies with QAPD requirements. The QA Director shall

review and accept the QA Program manual.

QA Level 3 Requirements

The applicant’s QA Level 3 Program is defined as standard commercial practice. A documented

QA Level 3 Program is not required. As such, QA Level 3 will govern all activities not designated

as QA Levels 1 or 2.

The QAPD, Appendix A to Chapter 11 of the applicant’s SAR (LES, 2005), provides additional

details and criterion to other QA elements that will be implemented to support the Management

Measures described in Chapter 11. NRC approved the QAPD on April 9, 2004 (NRC, 2004).

11.4 EVALUATION FINDINGS

11.4.1 CM

The NRC staff has reviewed the CM function for the applicant’s proposed uranium enrichment

facility in accordance with the regulatory acceptance criteria of NUREG-1520 (NRC, 2002). The

staff’s evaluation 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 implementation strategy for a CM

program that meets the requirements of 10 CFR 70.72 and provides adequate assurance that

facility changes are identified and controlled. IROFS are required to be identified and

documented, organizational responsibilities are identified and defined, and administrative

controls, policies, and procedures will be established to maintain the design configuration. 

Management-level policies and procedures are described that will provide the assurance of

consistency among design requirements, physical configuration, and facility documentation. 

These policies and procedures include requirements to perform an analysis and independent

safety review of any proposed activity involving IROFS, which ensures that consistency is
11-29

maintained for new activities or for those changes to existing activities involving licensed

material. The applicant’s management measures will include the following CM elements:

CM Policy

The applicant will develop and implement policies, procedures, organizational structure, and

personnel responsibilities necessary to effectively implement its CM program.

Design Requirements

The applicant’s design requirements and design bases will be documented and be appropriately

supported by analyses. All design documentation will be current and maintain consistency with

the physical configuration of the facility.

Document Control

Documents, including drawings, will be appropriately controlled, stored, and maintained. 

Drawings and related documents captured by the document control system will include those

documents necessary to adequately describe IROFS.

Change Control

The applicant will achieve and maintain strict consistency among design requirements,

documentation, and the physical configuration of the facility. The applicant will develop and

implement the policies, procedures, organizational structure, and personnel responsibilities

necessary to ensure consistency in design and design bases. These requirements include

adequate analysis, review, approval, and implementation of identified changes to IROFS.

Assessments

The applicant will conduct both initial and periodic assessments of its CM program, to assure its

effectiveness in meeting its goals and to correct deficiencies.

11.4.2 Maintenance

The applicant ensures the maintenance of IROFS. The applicant’s maintenance program

implementation strategy adequately address the basic elements required to ensure the

availability and reliability of IROFS. These elements include corrective maintenance, PM,

functional testing, equipment calibration, and work control for maintenance activities. The

applicant’s maintenance program will provide assurance that equipment performance will be

adequately monitored and assessed, using surveillance testing, functional testing, and

maintenance records. The surveillance/monitoring, PM, and functional testing activities

described by the applicant provide assurance that the IROFS identified in the ISA Summary will

be available and reliable to prevent or mitigate accidents.

The applicant’s maintenance program will be based on approved procedures. It will employ work

control methods that properly consider personnel safety, awareness of facility operations, QA,

and the requirements for corrective maintenance. The ISA Summary will be used to identify

IROFS requiring maintenance, and the determination of initial PM intervals will be based on
11-30

industry experience, with due consideration to the applicant’s equipment reliability goals. The

maintenance program will provide for training that emphasizes the importance of IROFS

identified in the ISA Summary, as well as regulations, codes, and personnel safety. 

Based on the above, the staff concludes that the applicant’s maintenance functions meet the

requirements of Part 70, and provide assurance of protecting the health and safety of workers,

the public, and the environment.

11.4.3 Training and Qualification

Based on review of the applicant’s strategy for development and implementation with respect to

training and qualification and comparison of these requirements to the review acceptance criteria

guidance in NUREG-1520 (NRC, 2002), the NRC staff has concluded that the applicant has

adequately described and assessed its personnel training and qualification in a manner that: (1)

satisfies regulatory requirements, and (2) is consistent with the guidance in NUREG-1520 (NRC,

2002). The staff’s evaluation found that the applicant’s description of its training program

appropriately covered: (a) training organization and management; (b) analysis and identification

of functional and position training requirements; (c) training basis and objectives; (d) organization

of instruction; (e) evaluation of trainee learning; (f) conduct of on-the-job training; (g) evaluation

of training program effectiveness; (h) personnel qualification; and (i) personnel evaluations. 

Based on this evaluation, the staff finds the applicant’s training program acceptable. 

There is assurance based on the NRC’s review that implementation of the applicant’s training

program, as described in the SAR (LES, 2005), will result in personnel who are qualified and

competent to design, construct, start up, operate, maintain, modify, and decommission the facility

safely. The staff concludes that the applicant’s plan for personnel training and qualification

meets the requirements of Part 70.

11.4.4 Procedures

Based on the review of the SAR (LES, 2005) and comparison of the applicant’s implementation

strategy to the review acceptance criteria in NUREG-1520 (NRC, 2002), the NRC staff has

determined that the applicant has described a suitably detailed process for the development,

approval, and implementation of procedures. The applicant’s strategy for developing

implementing procedures appropriately cover IROFS, as well as all other items important to

protecting the health and safety of workers, the public, and the environment. The staff concludes

that the applicant’s strategic plan for procedure development will meet the requirements of 10

CFR Part 70.

11.4.5 Audits and Assessments

Based on the review of the SAR (LES, 2005) and comparison of the applicant’s detailed plan and

processes to the review acceptance criteria in NUREG-1520 (NRC, 2002), NRC staff has

concluded that the applicant has adequately described its audit and assessment functions. The

staff has reviewed the applicant's implementation process used to describe audits and

assessments and the description of its policy directives, plans, and procedural requirements for

audits and assessments, taking into consideration such factors as: (1) the general structure of

typical audit and assessment activities; (2) facility procedures to be used to control audit and

assessment activities; (3) the use of qualified and independent audit and assessment personnel;
11-31

(4) the planned use of the results of audit and assessment activities; and (5) documentation,

planning, and implementation of corrective actions, based on the findings and recommendations

of audits and assessments. Based on the above, the staff finds that the applicant’s

implementation plan describing audits and assessments are acceptable.

The staff concludes that the applicant's plan for audits and assessments meets the requirements

of Part 70 and provides assurance of protection of the health and safety of workers, the public,

and the environment.

11.4.6 Incident Investigations

The applicant will develop processes to perform incident investigations. The applicant’s process

which describes the development and implementation methodologies with respect to its incident

investigation program include: (1) performing incident investigations of abnormal events that

may occur during operation of the facility; (2) determining the root cause(s) and generic

implications of the event; (3) recommending corrective actions for ensuring a safe facility and

safe facility operations; (4) monitoring and documenting corrective actions to completion; and (5)

maintenance of documentation so that “lessons learned” may be applied to future operations at

the facility. NRC staff has determined that these developmental and implementational methods

adequately follow the review acceptance criteria outlined NUREG-1520 (NRC, 2002). 

Accordingly, NRC staff concludes that the applicant’s description of the incident investigation

process complies with applicable NRC regulations and provides assurance of protection of the

health and safety of workers, the public, and the environment.

11.4.7 Records Management

The staff has reviewed the applicant’s implementation process concerning its records

management system against the acceptance criteria in NUREG-1520 (NRC, 2002). Based on

this review, NRC staff has concluded that the system will be effective in collecting, verifying,

protecting, and storing information concerning the facility, its design, operations, and

maintenance and the records management processes will provide for retrieval of the information

in readable form for the designated lifetime of the records. The staff verified the adequacy of the

applicant’s developmental and implementation protocols for using a records storage system that

is capable of protecting and preserving health and safety records that are stored at the facility

during the mandated periods, as well as the capabilities of the storage system for protecting

stored records from loss, theft, tampering, or damage during and after emergencies. The

applicant has also provided assurance that any deficiencies in the records management system

or its implementation will be detected and corrected in a timely manner. Based on this

evaluation, the staff finds that the applicant’s description and implementation processes for

records management are acceptable. 

11.4.8 Other QA Elements

Based on its review of the SAR (LES, 2005) the NRC staff has concluded that the applicant has

adequately described the application of other QA elements to IROFS, management measures, 

and other safety-related items. The NRC staff concludes that the applicant’s strategic approach 

in this area adequately address the implementation of the QA Program before beginning

operations at the facility. The staff has also determined that the applicant provides assurance

that personnel performing quality-related activities will perform work in accordance with approved
11-32

procedures and must demonstrate suitable proficiency in their assigned tasks. Additional

conclusions the staff reached through its evaluation of the applicant’s QA process, procedures,

and methods include the following:

1. The applicant has established, documented, and developed an organizational structure

responsible for developing, implementing, and assessing the management measures for

providing assurance of safe facility operations, in accordance with the acceptance criteria

in Section 11.4 of NUREG-1520 (NRC, 2002).

2. The applicant has established and documented a program to develop and implement QA

elements and administrative measures for staffing, performance, assessing findings, and

implementing corrective actions.

3. The applicant has developed a process for preparation and control of written plant

procedures, including a process for evaluating changes to procedures, IROFS, and tests. 

The process for review, approval, and documentation of procedures will be implemented

and maintained.

4. The applicant will develop and implement a program of surveillance, tests, and inspection

that will provide assurance of satisfactory in-service performance of IROFS. Specified

standards, acceptance criteria, and testing steps have been described or provided.

5. Periodic independent audits will be conducted to determine the effectiveness of

management measures. Management measures will provide for documentation of audit

findings and implementation of corrective actions.

6. Training requirements have been established and documented for ensuring that

employees are provided with the skills necessary to perform their jobs safely. 

Management measures have been provided for evaluation of the effectiveness of training

against predetermined objectives and criteria.

7. The organizations and personnel responsible for performing QA 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.

8. QA elements cover the IROFS, as identified in the ISA Summary, and measures are

established to prevent hazards from escalating into higher-risk events or accidents.

Accordingly, the staff concludes that the applicant’s application of other QA elements meets the

requirements of 10 CFR 70.62(d), and other applicable regulations (Appendix B of Part 50), and

provides the assurance of protection of public health and safety and protection of the

environment.

11.5 REFERENCES
11-33

(ANSI/ANS, 1996b) American National Standards Institute/American Nuclear Society

(ANSI/ANS). ANSI/ANS-8.19, “Administrative Practices for Nuclear Criticality Safety,” 1996.

(ANSI/ANS, 1991) American National Standards Institute/American Nuclear Society

(ANSI/ANS). ANSI/ANS-8.20, “Nuclear Criticality Safety Training,” 1991.

(ASME, 1994) American Society of Mechanical Engineers (ASME). ASME NQA-1, “Quality

Assurance Requirements for Nuclear Facility Applications,” 1994.

(ASME, 1995) American Society of Mechanical Engineers (ASME). ASME NQA-1a, “Addenda

to ASME NQA-1-1994 Edition, Quality Assurance Requirements for Nuclear Facility

Applications,” 1995.

(LES, 2005) Louisiana Energy Services (LES). “National Enrichment Facility Safety Analysis

Report,” Revision 6, 2005.

(NRC, 2002) U.S. Nuclear Regulatory Commission (NRC). NUREG-1520, “Standard Review

Plan for the Review of a License Application for a Fuel Cycle Facility,” 2002.

(NRC, 2004) U.S. Nuclear Regulatory Commission (NRC), letter to Louisiana Energy Services,

“Louisiana Energy Services Quality Assurance Program Description for the National Enrichment

Facility,” April 9, 2004.
11-34

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12-1

12.0 MATERIAL CONTROL AND ACCOUNTING

The Nuclear Regulatory Commission (NRC) staff’s review of the applicant’s material control and

accountability (MC&A) program contains information that has been marked as “Proprietary

Information” by the applicant, pursuant to 10 CFR 2.390.

The staff concluded that the applicant provided an acceptable Fundamental Nuclear Materials

Control Plan (FNMCP) for the proposed facility that will meet the applicable Part 74

requirements. 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

staff determined that the applicant meets the requirements in the area of MC&A to operate the

proposed facility, under Part 74.
12-2

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13-1

13.0 PHYSICAL PROTECTION

The Nuclear Regulatory Commission (NRC) staff’s review of the applicant’s Physical Security

Plan contains information that has been marked as “Proprietary Information” by the applicant,

pursuant to 10 CFR 2.390.

NRC staff reviewed the applicant’s Physical Security Plan. The methods and procedures

outlined in the Physical Security Plan satisfy the performance objectives, systems capabilities,

and reporting requirements specified in 10 CFR 73.67 and 73.71. The Physical Security Plan for

the facility is acceptable and meets the NRC requirements for physical protection of SNM of low

strategic significance.

In addition to a Physical Security Plan, the applicant submitted Guard Force Training and the

Safeguards Contingency Plans. However, Guard Force Training and the Safeguards

Contingency Plans are not required for Category III facilities in accordance with 10 CFR 73.67. 

These plans are only required for Category I licensees in accordance with 10 CFR 73.20 and

73.46. The staff performed an informal review of the LES Guard Force Training and the

Safeguards Contingency Plans to ensure there was not information in them that would contradict

a requirement in the Physical Security Plan. Since they are not required, the Guard Force

Training and the Safeguards Contingency Plans are not considered part of the licensing basis

and were not considered as part of the staff's review of the Physical Security Plan.

The protection of classified matter is described in the “National Enrichment Facility Standard

Practice Procedures Plan for the Protection of Classified Matter.” Evaluation of this plan is

discussed in Section 1.2.3.7 of this Safety Evaluation Report.
13-2

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14-1

14.0 PHYSICAL SECURITY OF THE 

TRANSPORTATION OF SPECIAL NUCLEAR MATERIAL 

OF LOW STRATEGIC SIGNIFICANCE

The Nuclear Regulatory Commission (NRC) staff’s review of the applicant’s Transportation

Security Plan contains information that has been marked as “Proprietary Information” by the

applicant, pursuant to 10 CFR 2.390.

The NRC staff reviewed the applicant’s Transportation Security Plan for SNM-LSS shipments

originating from, or arriving at, the facility. The approaches and procedures outlined in the

Transportation Security Plan satisfy the performance objectives, systems capabilities, and event

and advance notification requirements specified in 10 CFR 73.67(c); 73.67(g)(1)-(5); 73.71;

73.73; and 73.74. The NRC staff has concluded that the facility Transportation Security Plan is

acceptable and meets the NRC requirements for physical protection of SNM-LSS in transit.
14-2

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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. U.S. Nuclear Regulatory Commission (NRC) staff directed the

effort and contributed to the technical evaluations. Staff also used contractor support from the

Center for Nuclear Waste Regulatory Analyses and IEF Consulting in the preparation of this

document.

U.S. Nuclear Regulatory Commission Contributors

Timothy C. Johnson, Office of NRC Project Manager

 Nuclear Material Safety and Safeguards (NMSS)

Brian W. Smith, NMSS Section Chief, Gas Centrifuge Facility

Licensing Section

Paul Bell, NMSS Management Measures

Philip Brochman, NSIR Transportation Security

David Brown, NMSS Health Physics and Environmental

Protection

Oleg Bukharin, Office of Nuclear Security and Transportation Security

 Incident Response (NSIR) 

Frederick Burrows, NMSS Electrical Engineering and

Instrumentation and Controls

Sherri Cross, NSIR Material Control and Accounting

Michael Dusaniwskyj, Office of Nuclear Financial Qualifications

 Reactor Regulation (NRR)

Stanley Echols, NMSS Environmental Protection

J. Keith Everly, NSIR Information Security

Harry Felsher, NMSS Nuclear Criticality Safety

Alan Frazier, NSIR Physical Protection

Thomas Fredricks, NMSS Decommissioning Financial

Assurance
15-2

Herman Graves, Office of Nuclear Geotechnical Engineering

 Regulatory Research (RES)

Edward Johannemann, NSIR Physical Protection

Joel Klein, NMSS Health Physics and Emergency

Management

Joel Kramer, RES Human Factors Engineering

Amar Pal, NRR Electrical Engineering

Julius Persensky, RES Human Factors Engineering

Thomas Pham, NSIR Material Control and 

Accounting

Clayton L. Pittiglio, NRR Decommissioning Financial

Assurance

Roman Shaffer, RES Instrumentation and Controls

Wilkins Smith, NMSS Management Measures

William Troskoski, NMSS Integrated Safety Analysis and

Chemical Safety

Rex Wescott, NMSS Fire Safety

Center for Nuclear Waste Regulatory Analyses Contributors

Asadul H. Chowdhury Geotechnical Engineering

Sarah H. Gonzalez Seismic Engineering

Sui-Min (Simon) Hsiung Geotechnical Engineering

John Stamatakos Seismic Engineering

IEF Consulting Contributors

Craig Dean Decommissioning Financial

Assurance

Brantley Fry Decommissioning Financial

Assurance
15-3

Elizabeth Gormsen Decommissioning Financial

Assurance

Jennifer Mayer Decommissioning Financial

Assurance
15-4

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A-1

APPENDIX A - ACCIDENT ANALYSIS

The operation of the uranium enrichment facility that Louisiana Energy Services (LES) proposes

to be located in Lea County, New Mexico, would involve risks to workers, the public, and the

environment from potential accidents. The regulations in 10 CFR Part 70, Subpart H, “Additional

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

Material,” require that each applicant or licensee evaluate, in an “Integrated Safety Analysis”

(ISA), its compliance with certain performance requirements. The purpose of this section of the

Safety Evaluation Report is to summarize the methods and results the U.S. Nuclear Regulatory

Commission (NRC) staff uses to independently evaluate the consequences of potential accidents

identified in the applicant’s ISA. The accidents the staff has evaluated are a representative

selection of the types of accidents that are possible at the proposed facility.

A.1 ACCIDENT ANALYSIS METHODOLOGY

The analytical methods used in this consequence assessment are based on staff guidance for

analysis of nuclear fuel cycle facility accidents (NRC, 1990, 1991, 1998, 2001). With the

exception of the criticality accident, the hazards the NRC staff evaluated involve the release of

uranium hexafluoride (UF6) vapor from process systems that are designed to confine UF6 during

normal operations. As described below, UF6 vapor poses a chemical and radiological risk to

workers, the public, and the environment.

A.1.1 Selection of Representative Accident Scenarios

The Safety Analysis Report (SAR) (LES, 2005a) and Emergency Plan (EP) (LES, 2005c)

describe potential accidents that could occur at the facility. The applicant has provided accident

descriptions for two groups, according to the severity of the accident consequences: highconsequence events and intermediate-consequence events. The accident types are

summarized in Table A.1-1.

NRC staff selected, for detailed evaluation, a subset of the potential accident scenarios that is

intended to encompass the range of possible accidents. The accident sequences the staff

selected vary in severity from high- to low-consequence events, and include accidents initiated

by natural phenomena, operator error, and equipment failure. The list of accident sequences the

staff evaluated are listed in Table A.1-2.

A.1.2 Source Term Methodology

NRC staff evaluated the chemical and radiological hazard to workers, the public, and the

environment from accidental releases of UF6 vapor at the facility (See Section A.2.1 for the

source term methodology for the generic inadvertent criticality accident). For most accidents, the

UF6 vapor is assumed to escape its primary confinement system and enter an occupied room at

the facility. Staff assumed that UF6 would mix instantaneously with the air in the room.
A-2

Table A.1-1

Accident Types

High-Consequence Events Intermediate-Consequence Events

• Natural Phenomena • Carbon Trap Failure

- Earthquake • Pump Exhaust Plugged - Public

- Tornado • Spill of Failed Centrifuge Parts

- Flood • Dropped Contaminated Centrifuge

• Inadvertent Nuclear Criticality • Fire in Ventilated Room

• Fires Propagating Between Areas

• Fires Involving Excessive Transient

Combustibles

• Heater Controller Failure

• Over-filled Cylinder Heated to Ambient

• Product Liquid Sampling Autoclave

Heater Failure Followed by Reheat

• Open Sample Manifold Purge Valve and

Blind Flange

• Pump Exhaust Plugged - Worker

• UF6 Sub-sampling Unit Hot Box Heater

Controller Failure

• Empty UF6 Cold Trap (UF6 Release)

• Cylinder Valve/Connection Failure During

Pressure Test

• Chemical Dump Trap Failure

• Worker Evacuation (UF6 Release due to a

seismic event, fire, or other unplanned

release)

Table A.1-2

Accident Scenarios Representative of Potential Facility Events

• Generic Inadvertent Nuclear Criticality

• Hydraulic Rupture of a UF6 Cylinder in the Blending

and Liquid-Sampling Area

• Natural Phenomena Hazard - Earthquake

• Fire in a UF6 Handling Area

• Process Line Rupture in a Product Low-Temperature

Takeoff Station
A-3

For a constant release rate of UF6, the time-dependent concentration, C(t), of UF6 in a room or

workshop at the facility is (NRC, 1990):

Eq. 1 dC t

dt

R

V

QfCt

V

v v ( )

'

( )

' = −

where R = constant UF6 release rate, grams/second

V’ = k x f x V, the effective room volume, m3

V = actual room volume, m3

k = mixing efficiency (from National Fire Protection Association (NFPA) NFPA 69,

Appendix D (NFPA, 2002)), unitless

f = room free air fraction, unitless

Qv = room ventilation rate, m3

/second

fv = the fraction of Qv exhausted to the atmosphere (1-fv is recycled back into the

room), and

t = time elapsed since start of release, seconds

The values of mixing efficiency, k, and room free air fraction, f, are assumed to be 0.3 and 0.8,

respectively. The mixing efficiency is conservatively based on Table D-1 of NFPA 69 (NFPA,

2002), and is for ventilation systems with forced air supplies and single exhaust openings

comprised of grills and registers. The staff assumes that the room free air fraction value of 0.8

accounts for the volume of equipment that replaces free air inside the facility. The applicant

provided room volumes and ventilation flow rates (LES, 2004). The fraction of air exhaust is 10

percent, which is consistent with the Heating, Ventilation and Air Conditioning (HVAC)

descriptions in Chapters 3 and 4 of the SAR (LES, 2005a).

A solution to Equation 1 is:

C t Eq. 2 R

Q f

e

v v

Q f

V t v v

1( ) 1 ' = − ⎡

⎣

⎢ ⎤

⎦

⎥ −

Equation 2 defines the concentration, C1(t), during the period that UF6 is released at a steadystate rate, R, into a room. After T1 = 30 minutes, staff assumes that either the entire material at

risk would be released or the release would be stopped when operators intervene. The

assumption that operators or affected individuals downwind would respond within 30 minutes is

consistent with conservative self-protective criteria NRC uses to evaluate emergency

preparedness (NRC, 1988). After T1 = 30 minutes, the room would be ventilated until UF6 is

cleared from the room and exhausted to the environment. The room concentration, C2(t), after

all the material escapes to the room, or the release is stopped, is:

C t Eq. 3

R

Q f

e e

v v

QfT

V

Q f

V t vv vv

2 1 1

( ) ' ' = − ⎡

⎣

⎢ ⎤

⎦

⎥ − −

For the seismic event, the applicant has proposed safety-related equipment [i.e., items relied on

for safety (IROFS)] that shut down the HVAC systems in certain process areas. With no forced

ventilation, the primary means by which UF6, uranyl fluoride (UO2F2) particulate matter, and

hydrogen fluoride (HF) vapor enters the environment would be from small cracks and openings

in the building.
A-4

The volumetric leak rate from small cracks and openings in a building is calculated by evaluating

Poiseuille’s Law (Baker, 1987):

Q Eq. 4 dL

C W

dL

C W

C vW L

C L

S S pa S = −⎛

⎝

⎜ ⎞

⎠

⎟ +

⎛

⎝

⎜ ⎞

⎠

⎟ +

12 12 2 2 22 η

ρ

η

ρ

,

where QL = volumetric leak rate, m3 s-1

Ls = perimeter length of all exterior doors, m

W = width of the opening between door and frame, m

p1 = pressure of air indoors, N —2

p2 = pressure of air on building exterior, N —2

0 = coefficient of viscosity of air = 1.81x10-5 N s —2 @ T = 20 C

d = thickness of doors, m

C = 1.5

D = density of air = 1.183 kg —3 at T=25oC

v = windspeed, m s-1

Further, the values of Cp,a depend on the location of the door or opening relative to the direction

of the wind (Blevins, 2003):

Cp,a = 0.9 windward side of the building

Cp,a = -0.3 for leeward side of the building

Cp,a = -0.4 for building sides orthogonal to the wind direction

For this assessment, each exterior door in affected process areas of the building is assumed to

have a W = 0.16 cm (1/16-in.) opening around both sides and the top, and a W = 0.32cm (1/8-

in.) opening at the bottom. The thickness of all doors, d, is estimated to be 5 cm (2-in.). The

perimeter length of doors is estimated from drawings in the SAR (LES, 2005a).

The windspeed, v, assumed for the building-leakage calculations, was chosen with consideration

of the windspeed and stability class assumed in the derivation of the maximum atmospheric

dispersion factor, P/S. The staff calculated the highest P/S for the controlled-area boundary

(CAB) is 5.4x10-5 s —3. With corrections for building wake and low wind-speed plume meander,

the windspeed for F class stability conditions for which a P/S = 5.4x10-5 s —3 would be derived is

1.8 m s-1. Therefore, a bounding value of v = 2 m s-1 is used to estimate building leakage.

Solid UO2F2 produced by the reaction of UF6 with water vapor (i.e., humidity) forms a fine powder

that will settle by gravity. Therefore, in addition to removal by exfiltration through door cracks to

the environment, solid UO2F2 will also be removed from the air by settling on the floor and

equipment of the affected process area (NRC, 1998). The concentration in the building is

calculated as:

 Eq. 6 ( ) Ct C e L LO

V Q v At L d ( ) , ' = − + 1

where vd = settling velocity of UO2F2 particles in air, m s-1

A = floor area of the affected process area, m2
A-5

From Table 12.4 of U.S. Department of Energy (DOE)/TIC-27601 (DOE, 1984), staff estimated

the settling velocity of fine uranium compounds to be approximately 0.0001 cm s-1. The floor

areas of the affected process areas are estimated from drawings in the SAR (LES, 2005a).

A.1.3 NRC Performance Requirements

The performance requirements in Part 70, Subpart H, define acceptable levels of risk of

accidents at nuclear fuel cycle facilities, such as the proposed facility. The regulations in

Subpart H require that the applicant reduce the risks of credible high-consequence and

intermediate-consequence events, and assure that under normal and credible abnormal

conditions, all nuclear processes are subcritical. Threshold consequence values that define the

high- and intermediate-consequence events are described in Table A.1-3 from the facility SAR

(LES, 2005a).

Table A.1-3

Definition of High- and Intermediate-Consequence Events

Receptor Intermediate Consequence High Consequence

Worker - Radiological > 25 rem (0.25 Sv) > 100 rem (1 Sv)

Worker - Chemical

(10-minute exposure)*

> 19 mg U / m3

> 78 mg HF / m3

> 146 mg U / m3

> 139 mg HF / m3

Environment at the Restricted

Area Boundary

> 5.4 mg U / m3,

or 24-hour average release

greater than 5000 times the

values in Table 2 of Appendix

B of 10 CFR Part 20

N/A

Individual at the Controlled

Area Boundary - Radiological

> 5 rem (0.05 Sv) > 25 rem (0.25 Sv)

Individual at the Controlled

Area Boundary - Chemical

(30-minute exposure)

> 2.4 mg U / m3

> 0.8 mg HF / m3

> 13 mg U / m3

> 28 mg HF / m3

* Limits on uranium intake are also defined for workers in the immediate proximity of the

release. These limits are 10 mg and 40 mg uranium for intermediate- and highconsequence events, respectively.

A.1.4 Consequence-Assessment Approach for Acute Health Effects

The staff evaluated accident consequences for releases of UF6 for the facility worker, the

environment outside the restricted area boundary (RAB), an individual at the CAB, and the public

beyond the CAB. As stated above, the analytical methods used in this consequence

assessment are based on staff guidance for analysis of nuclear fuel cycle facility accidents

(NRC, 1990, 1991, 1998, 2001).
A-6

Facility Worker Exposure to Uranium and HF

The accident consequences to a facility worker include the risks of toxicological effects of

uranium intake, radiation dose from uranium intake, and exposure to HF concentration in air. 

The amount of uranium a facility worker could inhale (uranium intake) is calculated by assuming

the worker is exposed to C1(t) until T1 = 10 minutes after the start of the release (LES, 2005b). 

By T1 = 10 minutes, a worker is conservatively assumed to successfully escape the affected

room. Staff calculated uranium concentration for comparison with the proposed levels in Table

A.1-3. Uranium intake can be determined by multiplying the uranium concentration, time of

exposure, and breathing rate. The worker is assumed to inhale at a constant breathing rate of

3.33 x 10-4 m3

/s (20 liters per minute), which is consistent with the breathing rate NRC used in

Part 20, Appendix B, for Reference Man performing “light work.” Similarly, the HF concentration

to which a facility worker could be exposed is calculated by evaluating the time-averaged HF

concentration during the first T1 = 10 minutes.

For the uranium and HF exposure calculations, staff assumed that sufficient moisture (i.e.,

humidity) is present in the room to completely convert released UF6 gas to UO2F2 particulate

matter and HF vapor. This assumption results in a conservative estimate of the concentration of

HF vapor that would be present in both the affected room of the facility and downwind.

RAB 24-hour Average Uranium Concentration

In accordance with Part 70, Subpart H, the applicant must reduce the environmental risks of

accidents. The environmental consequences of accidents are evaluated at the RAB. At the

facility, the RAB is a fenced area inside the controlled area that includes the process buildings

and the uranium byproduct cylinder pad (LES, 2004). To evaluate whether accidents exceed the

environmental performance requirement, staff calculated the 24-hour average uranium

concentration at the restricted area boundary. Staff assumed that points of release are the

stacks on the roof of the technical services building (TSB).

The total source term for the first phase of the event (before the release is stopped) is S1. The

residual source term from the time that the release is stopped, T1, until the source is either

depleted, or until 24 hours has elapsed, is S2.

Eqs. 7, 8

S S t dt C t dt Q f R T V

Q f e for t T

S S t dt C t dt Q f R e V

Q f e

T T

v v

v v

Q fT

V

T

T

T

T

v v

Q fT

V

v v

Qf T T

V

v v

vv vv

1 1

0

1

0

1

1

1

22 2

1 1 1

1

2

1

2 1 21

1 0

1 1

= = ××= − − ⎧

⎨

⎩

⎫

⎬

⎭

⎡

⎣

⎢

⎢

⎤

⎦

⎥

⎥

< ≤

= = ××= − ⎡

⎣

⎢ ⎤

⎦

⎥ − ⎧

⎨

∫ ∫

∫ ∫

−

− − −

() () ' ,

() () '

'

'

( )

'

⎩

⎫

⎬

⎭

⎡

⎣

⎢

⎢

⎤

⎦

⎥

⎥

, for T t T 1 2 < ≤

To compare downwind concentrations with the applicable performance requirement, staff

calculated the uranium concentration downwind as a 24-hour average. For the RAB and the

CAB, staff calculated the atmospheric dispersion factor (P/S) for various distances from the

process buildings to the boundary in each downwind sector, using ARCON96 (NRC, 1997). The

distance to the RAB and CAB in each compass sector, the persistence of the wind in each

direction, and P/S values calculated using ARCON96, are presented in Table A.1-4. The highest

P/S at the RAB, which would result in the highest downwind concentration, occurs directly east of

the TSB. Therefore, the concentration at the RAB is calculated for wind blowing to the east.
A-7

Table A.1-4

Values of Atmospheric Dispersion Factors for the Facility Boundaries

Direction

from

Facility

Distance from Facility Frequency

of Wind

(percent)

RAB

P/S

(s/m3

)

CAB

P/S

(s/m3 RAB )

meters

(feet)

CAB

meters (feet)

S 160 (524) 417 (1368) 5.66 2.64E-04 4.84E-05

SSW 168 (552) 417 (1368) 3.98 2.40E-04 4.80E-05

SW 210 (690) 422 (1384) 4.91 1.69E-04 5.37E-05

WSW 261 (856) 503 (1650) 4.87 1.14E-04 4.08E-05

W 261 (856) 769 (2522) 6.29 1.14E-04 2.37E-05

WNW 278 (911) 1071 (3513) 5.52 9.96E-05 1.46E-05

NW 757 (2484) 1072 (3516) 7.52 2.12E-05 1.34E-05

NNW 639 (2098) 995 (3264) 10.80 2.35E-05 1.13E-05

N 589 (1932) 995 (3264) 20.40 2.67E-05 1.18E-05

NNE 530 (1739) 754 (2473) 7.35 3.08E-05 1.77E-05

NE 463 (1518) 581 (1906) 5.46 3.78E-05 2.61E-05

ENE 362 (1187) 540 (1771) 4.68 4.96E-05 2.61E-05

E 109 (359) 540 (1771) 4.45 4.49E-04 2.68E-05

ESE 101 (331) 540 (1771) 2.42 4.26E-04 2.54E-05

SE 143 (469) 487 (1597) 2.69 2.76E-04 3.10E-05

SSE 185 (607) 417 (1368) 3.04 1.70E-04 3.95E-05

The downwind concentration at the RAB is calculated for the downwind sector with the highest

P/S|RAB using Equation 9.

U Eq. 9 mg

m

S t dt S t dt

dt

g

s S

s

m

mg

g

mgU

RAB mgUF

T

T

T hr

T hr

RAB

,

() ()

,, . 3

1

0

2

24

0

24 3

3

6

1

1

2

2

= 10 0 68

+

⎡

⎣

⎢

⎢

⎤

⎦

⎥

⎥ × ××

∫ ∫

∫

=

=

Χ
A-8

CAB Uranium and HF Exposure

The accident consequences to an individual at the CAB include the risks of toxicological effects

of uranium intake, radiation dose from uranium intake, and exposure to HF concentration in air. 

The uranium concentration at the CAB is calculated for the downwind sector with the highest 

atmospheric dispersion factor (P/S|CAB). The highest P/S at the CAB, which would result in the

highest downwind concentration, occurs southwest of the TSB. Therefore, the accident

consequences at the CAB are calculated for wind blowing to the southwest.

The 30-minute average uranium concentration at the CAB is calculated using Equation 10.

 Eq. 10 [ ] U

S t dt S t dt g

s

g

s S

s

m

mg

g

mgU

mgUF

T

T

T hr

CAB

, min

() () ,

, 30 ,, . 1800

10 0 68

1

0

2

24

3

3

6

1

1

2

=

+

⎡

⎣

⎢

⎢

⎤

⎦

⎥

⎥ × ××

∫ ∫

=

Χ

Similarly, the unmitigated 30-minute average HF concentration is:

 Eq. 11

[ ] HF

S t dt S t dt g

s

g

s S

s

m

mg

g x mg HFmgUF

T

T

T hr

CAB

, min

() () ,

, 30 ,, . 1800

10 0 23

1

0

2

24

3

3

6

1

1

2

=

+

⎡

⎣

⎢

⎢

⎤

⎦

⎥

⎥ × ×

∫ ∫

=

Χ

A.1.5 Consequence-Assessment Methodology for Chronic Health Effects

Staff reviewed the potential, during a release of UF6, for acute exposures to HF to result in

chronic toxicological effects. Earlier studies have indicated that if fatality from suffocation caused

by edema (swelling) in the lungs does not occur, the swelling resulting from HF exposure will

subside and recovery should be complete. Thus, acute sublethal inhalation of HF is not

expected to have long-term effects (NRC, 1991). Therefore, the post-accident chronic health

effects the staff evaluated are limited to the toxicological and radiological health effects to

members of the public offsite resulting from exposure to uranium compounds.

The risk for chronic health effects is discussed in terms of latent cancer fatalities (LCFs). LCFs

are a measure of the expected number of additional cancer deaths in a population (or people

dying of cancer) as a result of exposure to radiation. Death from cancer induced by exposure to

radiation may occur at any time after the exposure takes place. However, latent cancers would

be expected to occur in a population from one year to many years after the exposure takes place

(NRC, 2005).

The staff has also previously reviewed human toxicological effects of exposure to soluble

uranium compounds (NRC, 1991). The staff concluded that a single acute intake of 10 mg of

soluble uranium would produce either minimal or nondetectable effects, either short-term or longterm, in humans. Therefore, if an accident could not result in acute intakes above 10 mg of

soluble uranium in any individual at or just beyond the site boundary (CAB), then no long-term

health effects would be expected among the exposed population further downwind. At the

facility, only one type of event is capable of causing toxicological effects from exposure to soluble

uranium among the offsite public -- the rupture of a large UF6 cylinder from inadvertent

overheating or overfilling. The protective measures the applicant proposed to prevent this type

of event are described in Section A.2.2.
A-9

The staff used GENII v. 1.485 (PNL, 1988) to estimate collective radiation doses (person-rem) to

members of the public resulting from post-accident inhalation and ingestion of soluble uranium

compounds. The exposure pathways the staff analyzed include inhalation of soluble uranium

carried by wind, external radiation from radioactivity deposited on the ground downwind of the

facility, and ingestion of contaminated food (produce, meat, and dairy products). The ingestion

parameters the staff used to estimate radiological doses to the public from accidents are

described in Table A.1-5. Staff used meteorological data for nearby Midland-Odessa (LES,

2005d). Staff obtained demographic information using SECPOP2000, a sector population, land

fraction, and economic estimate program prepared for NRC (NRC, 2003). Population data for

each of 16 downwind sectors at 10 distance intervals are shown in Table A.1-6. For releases of

uranium compounds, staff found that the north sector would have the highest collective doses

because Hobbs, New Mexico, is a large population center in the prevailing downwind direction.

Table A.1-5

Ingestion Parameters Used in GENII to Calculate 

Collective Radiological Dose to the Public from Accidents

Parameter Values for Consumption of Terrestrial Food

Food Type

General Population

Growing

Time (days)

Yield

(kg/m2)

Holdup Time

(days)

Consumption

Rate (kg/yr)

Leafy Vegetables 90 1.5 14 15

Root Vegetables 90 4.0 14 140

Fruit 90 2.0 14 64

Grains/Cereals 90 0.8 180 72

Parameter Values for Consumption of Animal Products

Food

Type

Consumption

Rate (kg/yr)

Holdup

Time

(days)

Stored Feed Fresh Forage

Diet

Fraction

Growing

Time

(days)

Yield

(kg/m2

)

Storage

Time

(days)

Diet

Fraction

Growing

Time

(days)

Yield

(kg/

m2

)

Storage

Time

(days)

Beef 70 34 0.25 90 0.80 180 0.75 45 2 100

Poultry 8.5 34 1 90 0.80 180 --- --- --- ---

Milk 230 3 0.25 45 2 100 0.75 30 1.5 0

Eggs 20 18 1 90 0.80 180 --- --- --- ---
A-10

Table A.1-6

Public Population in Sectors Surrounding the Facility

Sector 0-1 mile 1-2 miles 2-3 miles 3-4 miles 4-5 miles

N 00000

NNE 0 0 0 0 0

NE 0 0 0 0 0

ENE 0 0 0 0 0

E 00000

ESE 0 0 0 0 0

SE 0 0 0 0 0

SSE 0 0 0 0 0

S 00000

SSW 0 0 0 4 0

SW 0 0 0 0 0

WSW 0 0 0 0 15

W 0 0 11 53 2099

WNW 0 0 0 0 104

NW 0 0 0 5 2

NNW 0 0 0 0 0

Sector 5-10 miles 10-20 miles 20-30 miles 30-40 miles 40-50 miles

N 9 14637 12616 273 222

NNE 0 69 217 4760 1120

NE 0 49 995 7464 2809

ENE 0 7 430 972 46

E 0 7 45 351 41

ESE 0 0 105 12351 60

SE 0 23 18 20 848

SSE 0 0 19 8 18

S 0 4 37 3369 3754

SSW 6 4 2033 11 12

SW 17 12 3 1 3

WSW 34 9 13 2 8

W 484 13 2 4 21

WNW 35 20 0 9 8

NW 3 223 33 43 83

NNW 0 5044 4543 10565 1391
A-11

For dose calculations to the public, staff assumed that individuals downwind spend 100 percent 

of the time inside the passing plume (i.e., not sheltered). Staff also evaluated exposure of the

offsite population to contaminated soil resuspended by wind using a mass loading factor of 0.1-

mg soil per cubic meter of air.

A.2 ACCIDENT ANALYSIS

A.2.1 Generic Inadvertent Criticality

An inadvertent criticality at the facility would result from the unintended accumulation of enriched

uranium, leading ultimately to a self-sustaining or runaway nuclear chain reaction. A criticality

accident could release large amounts of heat and radiation. An inadvertent criticality could also

produce radioactive fission products, such as isotopes of noble gas like xenon and krypton,

radioiodines, and radiocesium. As a bounding case at the facility, one process area for which

this accident is postulated is the decontamination workshop.

Specifically, the accumulation of uranium in the citric acid tank could cause a criticality accident. 

For this to occur, the operator would have to fail to control the uranium mass in the tank. A

criticality in the solution in the tank could produce an initial burst of 1.0 x 1018 fissions, followed

by 47 bursts of 1.92 x 1017 fissions per burst, for a total of 1.0 x 1019 fissions in 8 hours.

The source term (ST) for the inadvertent nuclear criticality was determined using the five-factor

formula:

ST = MAR × DR × ARF × RF × LPF Eq.12

where MAR = material at risk

DR = damage ratio

ARF = airborne release fraction

RF = respirable fraction

LPF = leak path factor

For the criticality accident, the material at risk (MAR) is the amount of fission product radioactivity

that would accumulate during the event (NRC, 1998). The damage ratio (DR) is one, since all of

the solution in the tank would be involved in the event. The atmospheric release fraction (ARF)

for noble gases is 100 percent. The ARF for radioiodines is 0.25, and the ARF for other fission

products is 5x10-4 (NRC, 1998). The respirable fraction is assumed to be 100 percent. A leak

path factor of 0.001 is used for radioiodines and fission products other than noble gases, since

the gaseous effluent vent system (GEVS) in the TSB is equipped with High-Efficiency Particulate

Air (HEPA) filters and charcoal filters (LES, 2005b).

Exposure at the CAB was determined using the source terms calculated using Eq. 12, the P/S

value for the SW direction in Table A.1-2, and dose conversion factors from Federal Guidance

Report No. 11 (EPA, 1988).

The results of the staff consequence assessment are presented in Table A.2-1. Industry

experience with this type of criticality accident indicates that a worker is unlikely to survive the

accident if the individual were located in the immediate vicinity of the reaction. However, with
A-12

increasing distance from the accident, the radiation doses would be lower, and the probability

that a worker could survive increases. At the facility, workers have direct access to vessels and

other process equipment in which criticality events are possible. Therefore, staff have

qualitatively evaluated the accident as a high-consequence event for the worker.

Table A.2-1

Health Effects Resulting from Inadvertent Criticality

Worker

(egress after 10 min.)

Environment

at RAB

(Ratio)

Individual at CAB,

SW direction

Collective Dose,

West direction

High 0.66 0.14 rem person-rem LCFs

44 0.03

RAB - Restricted Area Boundary

CAB - Controlled Area Boundary

LCFs - latent cancer fatalities.

To convert rem to sievert, multiply by 0.01.

The environmental consequence is evaluated using the sum of the fractions rule. The

concentration at the RAB of each fission product radionuclide generated during a hypothetical

uranium solution criticality event in Table 3-14 of NUREG/CR-6410 (NRC, 1998), which was

used as a model for the generic inadvertent criticality event, is compared to 5000 times the

corresponding values in Appendix B to 10 CFR Part 20. The fractions thus generated (i.e.,

calculated fission product concentrations divided by their Appendix B limits) are added to yield

one value. If that value is less than 1, the accident consequences to the environment are low. 

Since the sum presented in Table A.2-1 is less than 1, the staff estimates that the postulated

criticality event is of low consequence to the environment.

A maximally exposed individual at the CAB in the southwest direction would receive a radiation

dose of 1.4 mSv (0.14 rem) total effective dose equivalent. This is a low consequence to this

individual. Similarly, the low collective dose to the offsite population, as determined using GENII

(PNL, 1988), in the west sector (Eunice, New Mexico) means that the risk of health effects to the

offsite public LCFs from this accident is low. The west sector would have the highest radiation

doses after a criticality accident because Eunice, New Mexico, lies in closer proximity to the

proposed facility than other population centers, and short-lived radionuclides formed during the

criticality accident would not have completely decayed before reaching Eunice, New Mexico. 

Larger population centers in the north sector, such as the city of Hobbs, New Mexico, would

receive lower collective doses because the short-lived fission products would decay during the

time the plume transits from the proposed facility.

In accordance with the performance requirements of Part 70, Subpart H, the applicant has either

identified IROFS to reduce the risk to the facility worker from all criticality accidents or identified

safe-by-design components that meet criteria such that they are highly unlikely to fail.

A.2.2 Hydraulic Rupture of a UF6 Cylinder in the Blending and Liquid-Sampling Area
A-13

At the product blending system in the blending and liquid-sampling area of the separations

building, 30B (2.5-ton) cylinders are filled with product to customer specifications. The transfer of

product to 30B cylinders would begin by heating a 14-ton 48Y cylinder containing product UF6,

inside a blending donor station, to no more than 61o

C (142oF). The heated UF6 gas would be

transferred by piping from the heated 48Y cylinder to a blending receiver station containing a

30B cylinder. The blending receiver station is cooled, which allows the UF6 gas to desublime to

a solid inside the 30B cylinder, completing the transfer.

An accident is postulated wherein the blending donor station heater controller fails, causing the

blending donor heater within the station to remain on. Were this to occur, the 48Y product

cylinder could overheat and the cylinder could hydraulically rupture because of the expansion of

the liquid UF6. On cylinder rupture, the entire contents of the 48Y product cylinder [12,501 kg

(27,560 lbs) of UF6] would be released within the blending donor station. Since the station

enclosure is not air-tight, the UF6 would be released to the blending and liquid-sampling area. 

The UF6, when in contact with air, would produce HF gas and UO2F2. The release into the

building would then be released to the environment. The HVAC is conservatively assumed to be

operating at the maximum ventilation flow rate. Significant quantities of HF and UO2F2 would be

carried by the prevailing wind beyond the CAB.

The results of the staff-consequence assessment are presented in Table A.2-2.

The health and environmental consequences of this accident are high. A facility worker in the

vicinity of the blending donor station would, within seconds, be exposed to lethal UF6, UO2F2, and

HF concentrations. The environmental consequences are higher than the 5.4 U mg/m3

 threshold

for an intermediate consequence. An individual located at the CAB in the southwest sector

would suffer high consequences from both uranium and HF exposure. The collective dose to the

offsite population in the north sector indicates a risk of several LCFs in the population in the

years after the accident.

In accordance with the performance requirements of Part 70, Subpart H, the applicant has

identified IROFS to reduce the risk to the facility workers, the public, and the environment from

the effects of this accident. To prevent this accident, the applicant will rely on fail-safe hardwired high-temperature heater trips and redundant independent fail-safe capillary hightemperature heater trips. Each control will be tested annually to ensure its availability and

reliability to serve its intended safety function on demand. The purpose of these controls is to

ensure that the accident is highly unlikely to occur. In addition, there have been no similar heater

control failures at the Urenco facilities in Europe in over 30 years of operation.

In addition to IROFS, the applicant has committed to an EP that includes certain mitigating

actions to reduce the consequences of the event. For example, in response to an alarm that 

indicates the release of UF6, a control room operator can secure the HVAC system for the

affected area. The action to secure the HVAC within minutes of the accident would considerably

reduce the risk to the public and the environment.

Table A.2-2

Health Effects Resulting from Hydraulic Rupture of a UF6 Cylinder
A-14

Worker

(egress after 10 min.)

Environment

at RAB

Individual at CAB,

SW direction

Collective Dose,

north direction

U mg/m3

(rem)

HF,

mg/m3

U mg/m3 U mg/m3

(rem)

HF,

mg/m3

person-rem LCFs

High 44 250

(0.97)

86 12,000 7

RAB - Restricted Area Boundary

CAB - Controlled Area Boundary

HF - hydrogen fluoride.

LCFs - latent cancer fatalities.

A.2.3 Natural Phenomena Hazard - Earthquake

An earthquake is postulated to breach all UF6 piping systems and lead to a release of

approximately 860 kilograms (kg) (1900 lb) of UF6 (LES, 2005b). Staff evaluated this accident

for the blending and liquid-sampling area; UF6 handling areas; and the cascade halls. The

applicant has committed to ensure that the affected process buildings can withstand the design

basis earthquake. Therefore, for this evaluation, staff assumed that the buildings would remain

intact. The applicant will also install and maintain an electrical trip system for selected HVAC

systems in process areas with large inventories of gaseous UF6. The trip system will detect

earthquakes and secure the HVAC units. Therefore, for this evaluation, staff has assumed that

the HVAC in affected process buildings is shut down.

The results of the staff consequence assessment are presented in Table A.2-3.

The consequences to a facility worker shown in Table A.2-3 are for a worker located in one of

the cascade halls during the earthquake. Depending on the location of the worker when the

event occurs, the large quantity of UF6 which could be released would result in a high

consequence to this individual before the individual could escape the room. However, for

seismic events, the worker is assumed to evacuate the area of concern upon detection of a

seismic event, which results in a reduced exposure time and an acceptable risk. The

consequences to the environment are low. The maximally exposed individual at the CAB in the

southwest direction would not be expected to suffer any observable heath effects. Similarly, the

low collective dose to the offsite population in the north sector means that the risk of health

effects to the offsite public (latent cancer) from this accident is low.

A.2.4 Fire in a UF6 Handling Area

A fire involving transient combustible material is postulated to breach a UF6 transfer manifold

containing feed vapor from five feed stations in a single UF6 Handling Area. The release would

involve approximately 3.4 kg (7.5 lb) of UF6 vapor.

Table A.2-3

Health Effects Resulting from an Earthquake
A-15

Worker

(egress after 10 min.)

Environment

at RAB

Individual at CAB,

SW direction

Collective Dose,

north direction

U mg/m3

(rem)

HF,

mg/m3

U mg/m3 U mg/m3

(rem)

HF,

mg/m3

person-rem LCFs

High 0.11 0.64

(0.0017)

0.22 14 0.008

RAB - Restricted Area Boundary

CAB - Controlled Area Boundary

HF - hydrogen fluoride.

LCFs - latent cancer fatalities.

The results of the staff consequence assessment are presented in Table A.2-4.

Table A.2-4

Health Effects Resulting from Fire in a UF6 Handling Area

Worker

(egress after 10 min.)

Environment

at RAB

Individual at CAB,

SW direction

Collective Dose,

north direction

U mg/m3

(rem)

HF,

mg/m3

U mg/m3 U mg/m3

(rem)

HF,

mg/m3

person-rem LCFs

59

(0.020)

20 0.012 0.070

(0.000072)

0.024 0.92 0.0006

RAB - Restricted Area Boundary

CAB - Controlled Area Boundary

HF - hydrogen fluoride.

LCFs - latent cancer fatalities.

The consequences of this accident are low for the environment, the individual at the CAB, and

the public offsite. For the facility worker, the consequences are intermediate for acute chemical

exposure to uranium. However, for fires, the worker is assumed to evacuate the area of concern

once the fire is detected, which would result in an exposure time much shorter than 10 minutes,

thus resulting in acceptable risk.

In accordance with the performance requirements of Part 70, Subpart H, the applicant has

identified IROFS to ensure the risk of this type of accident remains low. To reduce the

magnitude of fires resulting from the presence of transient combustible material, the applicant will

rely on administrative controls. The purpose of these controls is to prevent large fires that could

result in the release of large inventories of UF6 by administratively limiting transient combustible

loading in areas containing uranic material to ensure that the integrity of uranic material

components/containers is maintained.

A.2.5 Process Line Rupture in a Product Low-Temperature Takeoff Station (LTTS)

Cold traps and chemical traps would be used at the facility to remove residual UF6 and HF from

process lines before discharging exhaust gases from these lines to the GEVS. An accident

could occur if a product vent subsystem carbon trap became saturated with UF6, caused by a

small UF6 leak through a product cold trap valve. Were this to occur, a UF6 plug could form on
A-16

the discharge of the vacuum pump, causing high pressure in the vacuum pump and, thus, failing

seals, leading to a release of approximately 1.0 kg (2.2 lb) of UF6 vapor to the UF6 handling area. 

The results of the staff-consequence assessment are presented in Table A.2-5.

The consequences of this accident are low for the facility worker, environment, the individual at

the CAB, and the public offsite. In accordance with the performance requirements of Part 70,

Subpart H, the applicant has identified IROFS to ensure the risk of this type of accident remains

low. For this accident, a preventive measure is a fail-safe, hard-wired, high-carbon-trap weight

trip of the vacuum pump. This equipment will be tested annually to ensure its availability and

reliability to serve its intended safety function.

Table A.2-5

Acute Health Effects Resulting from Process Line Rupture in a Product LTTS

Worker

(egress after 10 min.)

Environment

at RAB

Individual at CAB,

SW direction

Collective Dose,

NNW direction

U mg/m3

(rem)

HF,

mg/m3

U mg/m3 U mg/m3

(rem)

HF,

mg/m3

personremLCFs

17

(0.022)

5.8 0.0035 0.020

(0.000078)

0.0069 0.97 0.0006

RAB - Restricted Area Boundary

CAB - Controlled Area Boundary

HF - hydrogen fluoride.

LCFs - latent cancer fatalities.

A.3 ACCIDENT ANALYSIS SUMMARY

Results of the accident analysis are summarized in Table A.3-1. Staff selected and evaluated a

representative subset of the potential accidents that could occur at the proposed facility. The

accident consequences vary in magnitude, and include accidents initiated by natural

phenomena, operator error, and equipment failure. Analytical results indicate that accidents at

the facility pose acceptably low risks. The most significant accident consequences are those

associated with the release of UF6 caused by rupturing an over filled or over heated cylinder. 

The facility design reduces the risk (likelihood) of this event by using redundant heater controller

trips. In addition, the facility EP addresses this type of event, and all other lower-risk, high- and

intermediate-consequence events. The NRC staff concludes that through the combination of

plant design, passive and active engineered controls (IROFS), and administrative controls,

accidents at the facility pose an acceptably low risk to workers, the environment, and the public.
A-17

Table A.3-1

Summary of Health Effects Resulting from Accidents

Workera

Environment at

RAB

Individual at CAB,

SW direction Collecti

Accident [U], mg/m3

(rem)

[HF],

mg/m3

[U], mg/m3 [U], mg/m3

(rem)

[HF],

mg/m3

Direction pe

r

InadvertentNuclear Criticality Highb 0.66c (0.14d) --- West 

Hydraulic Ruptureof a UF6 Cylinder Highb 44 250 86 North 12

(0.97)

Earthquake Highb 0.11 0.64 0.22 North 

(0.0017)

Fire in a UF6

Handling Area

59

20 0.012

0.070 0.024 North 0

(0.020) (0.000072)

Process LineRupture

17

5.8 0.0035

0.020 0.0069 North 0

(0.022) (0.000078)

a Worker exits after 10 minutes.b High consequence could lead to a fatality.c Pursuant to 10 CFR 70.61(c)(3), this value is the sum of the fractions of individual fission product radionuclide concentrations over 5000 times the concappear in 10 CFR Part 20, Appendix B, Table 2d The dose to the individual at the CAB is the sum of internal and external doses from fission products released from the Technical Services Building ga

systems stack.

e Though the consequences of the rupture of a liquid-filled UF6 cylinder would be HIGH, redundant heater-controller trips would make this event highly u

RAB - Restricted Area BoundaryCAB - Controlled Area BoundaryHF - hydrogen fluoride.LCF - latent cancer fatalities.mg/m3 - milligrams per cubic meter.

To convert rem to sievert, multiply by 0.01.
A-18

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A-19

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