Safety Evaluation Report for the 

Eagle Rock Enrichment Facility in 

Bonneville County, Idaho

Docket No. 70-7015

AREVA Enrichment Services, LLC

Office of Nuclear Material Safety and Safeguards 

NUREG-1951
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Safety Evaluation Report for the 

Eagle Rock Enrichment Facility in

Bonneville County, Idaho

Docket No. 70-7015

AREVA Enrichment Services, LLC

Manuscript Completed: September 2010 

Date Published: September 2010

Office of Nuclear Material Safety and Safeguards

NUREG-1951
ABSTRACT

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

and safeguards evaluation of AREVA Enrichment Services LLC's (AES's) application for a

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

material, source material, and special nuclear material (SNM). The proposed facility is known

as the Eagle Rock Enrichment Facility (EREF). AES proposes that the EREF be located in

Bonneville County, Idaho, about 32 kilometers (20 miles) west northwest of the city of

Idaho Falls. The EREF will possess natural, depleted, and enriched uranium, and will be

authorized to enrich uranium up to a maximum of 5 weight percent uranium-235.

The objective of the NRC's review is to evaluate the facility's potential adverse impacts 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.

Potential environmental impacts associated with the proposed facility and its reasonable

alternatives will be addressed in a separate NRC document, the Final Environmental Impact

Statement, which is expected to be issued in February 2011.

iii

CONTENTS

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

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

EXECUTIVE SUM MARY ..................................................................................................... xv

ACRONYMS AND ABBREVIATIONS ....................................................................................... xxi

CHAPTER 1.0 GENERA L INFORMATION ............................................................................. 1-1

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

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

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

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

1.1.4 Evaluation Findings ............................................................................... 1-5

1.2 Institutional Information ..................................................................................... 1-5

1.2.1 Regulatory Requirements ..................................................................... 1-6

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

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

1.2.3.1 Corporate Identity ................................................................ 1-7

1.2.3.2 Foreign Ownership, Control, or Influence ............................. 1-8

1.2.3.3 Financial Qualifications ........................................................ 1-9

1.2.3.4 Liability Insurance .............................................................. 1-10

1.2.4 Type, Quantity, and Form of Licensed Material ................................... 1-11

1.2.4.1 Authorized Uses ................................................................ 1-12

1.2.4.2 Special Exem ptions or Special Authorizations ................... 1-13

1.2.4.3 Protection of Classified Matter ........................................... 1-16

1.2.5 Evaluation Findings ............................................................................. 1-17

1.3 Site Description .............................................................................................. 1-18

1.3.1 Regulatory Requirements ................................................................... 1-18

1.3.2 Regulatory Guidance and Acceptance Criteria .................................... 1-19

1.3.3 Staff Review and Analysis ................................................................... 1-19

1.3.3.1 Site Geography .................................................................. 1-19

1.3.3.2 Demographics ................................................................... 1-21

1.3.3.3 Meteorology ....................................................................... 1-22

1.3.3.4 Geology ............................................................................. 1-26

1.3.4 Evaluation Findings ............................................................................. 1-35

1.4 References ..................................................................................................... 1-35

CHAPTER 2.0 ORGANIZATION AND ADM INISTRATION ..................................................... 2-1

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

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

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

2.3.1 Organization ......................................................................................... 2-1

V
2.3.2 Organizational Responsibilities and Q ualifications ................................ 2-3

2.3.3 M anagem ent Control ........................................................................... 2-12

2.4 Evaluation Findings ........................................................................................ 2-15

2.5 References ..................................................................................................... 2-16

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

3.1 Regulatory Requirem ents ................................................................................. 3-1

3.2 Regulatory G uidance And Acceptance Criteria ................................................. 3-3

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

3.3.1 Safety Program and ISA Com m itm ents ................................................. 3-3

3.3.1.1 Process Safety inform ation .................................................. 3-4

3.3.1.2 ISA Com m itm ents ................................................................ 3-5

3.3.1.3 M anagem ent M easures ....................................................... 3-6

3.3.1.4 Safety Program and ISA Commitments Conclusion ............. 3-7

3.3.2 ISA Methodology ................................................................................... 3-7

3.3.2.1 Hazard Identification and Evaluation .................................... 3-7

3.3.2.2 Definition of Receptors for Consequence Evaluations ....... 3-10

3.3.2.3 Likelihood Evaluation M ethod ............................................ 3-11

3.3.2.4 Chem ical Consequences ................................................... 3-11

3.3.2.5 Radiological Consequences .............................................. 3-13

3.3.2.6 Environm ental Consequences ........................................... 3-14

3.3.2.7 ISA M ethodology Conclusion ............................................. 3-14

3.4 Evaluation ....................................................................................................... 3-14

3.5 References ..................................................................................................... 3-15

CHAPTER 4.0 RADIATIO N PROTECTIO N ............................................................................ 4-1

4.1 Regulatory Requirem ents ................................................................................. 4-1

4.2 Regulatory G uidance and Acceptance Criteria ................................................. 4-2

4.3 Staff Review and Analysis ................................................................................ 4-2

4.3.1 RP Program Im plem entation ................................................................. 4-2

4.3.2 ALARA Program ................................................................................... 4-3

4.3.3 O rganization and Personnel Q ualifications ............................................ 4-6

4.3.4 W ritten Procedures ............................................................................... 4-7

4.3.5 Training ................................................................................................. 4-8

4.3.6 Ventilation and Respiratory Protection Program s ................................ 4-10

4.3.7 Radiation Survey and M onitoring Program s ........................................ 4-13

4.3.8 Additional Program Requirem ents ....................................................... 4-20

vi
4.4 Evaluation Findings ........................................................................................ 4-21

4.5 References ..................................................................................................... 4-22

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

5.1 Regulatory Requirem ents ................................................................................ 5-1

5.2 Regulatory Acceptance Criteria ........................................................................ 5-1

5.3 Staff Review and Analysis ................................................................................ 5-1

5.3.1 Industry Standards ................................................................................ 5-1

5.3.2 O rganization and Adm inistration ........................................................... 5-2

5.3.3 Managem ent of the NCS Program ........................................................ 5-3

5.3.4 NCS M anagem ent Measures ................................................................ 5-4

5.3.4.1 Training .............................................................................. 5-4

5.3.4.2 Procedures .......................................................................... 5-4

5.3.4.3 Audits and Assessm ents ...................................................... 5-4

5.3.5 NCS M ethodologies and Technical Practices ........................................ 5-5

5.3.5.1 NCS Methodologies ............................................................. 5-5

5.3.5.2 NCS Technical Practices ..................................................... 5-9

5.3.5.3 Findings ............................................................................. 5-12

5.3.6 CAAS .................................................................................................. 5-12

5.3.6.1 SAR Com m itm ents ............................................................ 5-12

5.3.6.2 ISA Sum m ary .................................................................... 5-12

5.3.6.3 Findings ............................................................................. 5-13

5.3.7 Reporting Requirements ................................ 5-13

5.3.8 ISA ...................................................................................................... 5-13

5.3.8.1 ISA Com m itm ents .............................................................. 5-13

5.3.8.2 ISA Summary Review ......................... 5-16

5.3.8.3 Conclusion ......................................................................... 5-19

5.3.9 Facility Changes and Configuration M anagem ent ............................... 5-19

5.4 Evaluation Findings ........................................................................................ 5-20

5.5 References ..................................................................................................... 5-21

CHAPTER 6.0 CHEM ICAL PRO CESS SAFETY .................................................................... 6-1

6.1 Regulatory Requirem ents ................................................................................. 6-1

6.2 Regulatory G uidance and Acceptance Criteria ................................................. 6-1

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

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

6.3.1.1 G as Centrifuge Process ....................................................... 6-2

6.3.1.2 Chem ical Screening and Classification ................................ 6-3

6.3.1.3 Hazardous Chemicals and Chemical Interactions ................ 6-5

6.3.1.4 UF6 and Interaction Chem icals ............................................ 6-6

6.3.1.5 UF6 and Construction Materials ........................................... 6-7

vii
6.3.1.6 Process Description Conclusion .......................................... 6-8

6.3.2 Chem ical Accident Sequences .............................................................. 6-9

6.3.3 Chem ical Accident Consequences ...................................................... 6-10

6.3.4 Items Relied On For Safety (IROFS) and Management Measures ...... 6-11

6.3.4.1 Chem ical Process IRO FS .................................................. 6-11

6.3.4.2 M anagem ent Measures ..................................................... 6-11

6.3.5 Em ergency Managem ent .................................................................... 6-16

6.3.6 Baseline Design Criteria ................................ 6-16

6.4 Evaluation Findings ........................................................................................ 6-18

6.5 References ..................................................................................................... 6-18

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

7.1 Regulatory Requirem ents ................................................................................. 7-1

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

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

7.3.1 Fire Safety M anagem ent Measures ...................................................... 7-1

7.3.1.1 M anagem ent Policy and Direction ....................................... 7-2

7.3.1.2 Fire Safety Management Measures Conclusions ................. 7-2

7.3.2 Fire Hazards Analysis (FHA) ................................................................. 7-3

7.3.2.1 IRO FS Related to Fire Safety .............................................. 7-4

7.3.2.2 Fire in the CRSB .................................................................. 7-6

7.3.2.3 Fire Involving Cylinder Transporters/M overs ........................ 7-7

7.3.2.4 Fire in the Separations Building ........................................... 7-7

7.3.2.5 Fire in the UF6 Handling Area or Blending, Sampling, and

Preparation Building ............................................................ 7-7

7.3.2.6 Fire in the TSB ..................................................................... 7-8

7.3.2.7 Cylinder Storage Pads ......................................................... 7-9

7.3.2.8 FHA Conclusions ............................................................... 7-10

7.3.3 Facility Design ..................................................................................... 7-10

7.3.3.1 Facility Passive-Engineered Fire Protection Systems ........ 7-10

7.3.3.2 Facility Active-Engineered Fire Protection Systems ........... 7-11

7.3.3.3 Facility Design Conclusions ............................................... 7-13

7.3.4 Process Fire Safety ............................................................................. 7-14

7.3.4.1 Process Descriptions ......................................................... 7-14

7.3.4.2 Process Fire Safety Conclusions ....................................... 7-16

7.3.5 Fire and Em ergency Response ........................................................... 7-16

7.4 Evaluation Findings ........................................................................................ 7-17

7.5 References ..................................................................................................... 7-17

CHAPTER 8.0 EM ERG ENCY MANAG EM ENT ...................................................................... 8-1

8.1 Regulatory Requirem ents ................................................................................. 8-1

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

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

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

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

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

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

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

8.3.6 M itigation of Consequences .................................................................. 8-4

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

8.3.8 Responsibilities ..................................................................................... 8-5

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

8.3.10 Information To Be Comm unicated ........................................ 8-6

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

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

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

8.4 Evaluation Findings .......................................................................................... 8-9

8.5 References ....................................................................................................... 8-9

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

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

Effl uent 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 .. 9-5

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

9.3.1.5 W aste Minim ization .............................................................. 9-6

9.3.1.6 Safe Handling of Radioactive W astes .................................. 9-7

9.3.2 Effl uent 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 ................................................. 9-11

9.3.2.4 Environmental Monitoring .................................................. 9-12

9.3.3 ISA Sum mary ...................................................................................... 9-13

9.4 Evaluation Findings ........................................................................................ 9-15

9.5 References ..................................................................................................... 9-15

CHAPTER 10.0 DECOMM ISSIONING ......................................................................... 10-1

10.1 Regulatory Requirem ents ............................................................................... 10-1

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

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

10.3.1 Request for Alternate Schedule for Decommissioning ......................... 10-2

10.3.2 Conceptual Decontamination and Decommissioning Activities ............ 10-3

10.3.3 Financial Assurance for Decommissioning .......................................... 10-6

10.3.3.1 Description of Incremental Funding Approach and Approach to

Updating the DFP, Cost Estimate, and Financial Instruments

.......................................................................................... 1 0 -7

10.3.3.2 Discussion on the Adequacy of the Cost Estimate ........... 10-13

10.3.3.3 Review of Financial Instruments ...................................... 10-14

10.4 Evaluation Findings ...................................................................................... 10-16

10.5 References ................................................................................................... 10-16

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

11.1 Regulatory Requirements ............................................................................... 11-2

11.2 Regulatory Acceptance Criteria ...................................................................... 11-2

11.3 Staff Review And Analysis .............................................................................. 11-3

11.3.1 Configuration Management (CM) ........................................................ 11-3

11.3.1.1 CM Policy .......................................................................... 11-3

11.3.1.2 Design Requirements ........................................................ 11-4

11.3.1.3 Document Control .............................................................. 11-5

11.3.1.4 Change Control .................................................................. 11-6

11.3.1.5 Assessments ..................................................................... 11-7

11.3.1.6 Evaluation Findings ............................................................ 11-7

11.3.2 Maintenance ....................................................................................... 11-8

11.3.2.1 Surveillance/Monitoring ...................................................... 11-9

11.3.2.2 Corrective Maintenance ..................................................... 11-9

11.3.2.3 Preventive Maintenance ................................................. 11-9

11.3.2.4 Functional Testing ............................................................ 11-10

11.3.2.5 Evaluation Findings .......................................................... 11-12

11.3.3 Training and Qualifications ................................................................ 11-13

11.3.3.1 Organization and Management of the Training Function..11-13

11.3.3.2 Analysis and Identification of Functional Areas Requiring

T ra in ing ............................................................................ 1 1-13

11.3.3.3 Position Training Requirements ....................................... 11-14

11.3.3.4 Basis and Objectives for Training ..................................... 11-19

11.3.3.5 Organization of Instruction, Using Lesson Plans and Other

Training Guides ............................................................... 11-19

11.3.3.6 Evaluation of Trainee Learning ........................................ 11-19

11.3.3.7 Conduct of OJT Training .................................................. 11-19

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

11.3.3.9 Personnel Qualification .................................................... 11-20

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

11.3.3.11 Evaluation Findings .......................................................... 11-21

11.3.4 Procedures Development and Implementation .................................. 11-22

11.3.4.1 Preparation of Procedures ............................................... 11-24

11.3.4.2 Administrative Procedures ............................................... 11-24

x
11.3.4.3 Procedures ...................................................................... 11-25

11.3.4.4 Changes to Procedures ................................................... 11-26

11.3.4.5 Distribution of Procedures ................................................ 11-26

11.3.4.6 Evaluation Findings .......................................................... 11-26

11.3.5 Audits and Assessm ents ................................................................... 11-27

11.3.5.1 Scheduling of Audits and Assessments ............. 11-28

11.3.5.2 Procedures for Audits and Assessm ents .......................... 11-29

11.3.5.3 Qualification and Responsibilities for Audits and Assessments

........................................................................................ 1 1 -2 9

11.3.5.4 Evaluation Findings .......................................................... 11-30

11.3.6 Incident Investigations and Corrective Action Process ...................... 11-30

11.3.6.1 Incident Investigations ..................................................... 11-31

11.3.6.2 Corrective Action Process ................................................ 11-32

11.3.6.3 Evaluation Findings .......................................................... 11-32

11.3.7 Records M anagem ent ....................................................................... 11-33

11.3.7.1 Records M anagem ent program ........................................ 11-33

11.3.7.2 Evaluation Findings .......................................................... 11-34

11.3.8 Other QA Elem ents ........................................................................... 11-35

11.3.8.1 Description of other QA Elem ents .................................... 11-35

11.3.8.2 Evaluation Findings .......................................................... 11-36

11.4 Evaluation Findings ...................................................................................... 11-36

11.4.1 Configuration Managem ent ............................................................... 11-36

11.4.2 M aintenance ..................................................................................... 11-37

11.4.3 Training and Q ualification .................................................................. 11-38

11.4.4 Procedures ........................................................................................ 11-38

11.4.5 Audits and Assessm ents ................................................................... 11-38

11.4.6 Incident Investigations ....................................................................... 11-39

11.4.7 Records M anagem ent ....................................................................... 11-39

11.4.8 Other QA Elem ents ........................................................................... 11-39

11.5 References ................................................................................................... 11-40

CHAPTER 12.0 PHYSICAL PROTECTION .................................................................. 12-1

12.1 Regulatory Requirements ............................................................................... 12-1

12.2 Regulatory Guidance ...................................................................................... 12-1

12.3 Staff Review and Analysis .............................................................................. 12-1

12.4 Evaluation Findings ........................................................................................ 12-2

CHAPTER 13.0

APPENDIX A

APPENDIX B

APPENDIX C

APPENDIX D

SAFETY EVALUATION REPORT PREPARERS ................................ 13-1

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

ACCIDENT ANALYSIS ............................................................ B-1

COST ESTIMATE TO CONSTRUCT AND OPERATE ................... C-1

HUMAN FACTORS ............................................................... D-1

xi
APPENDIX E

APPENDIX F

APPENDIX G

APPENDIX H

ELECTRICAL SYSTEM AND INSTRUMENTATION AND CONTROL... E-1

STRUCTURAL DESIGN .......................................................... F-1

NUCLEAR CRITICALITY SAFETY ................................................ G-1

MATERIAL CONTROL AND ACCOUNTING ................................ H-1

xii
TABLES

Table 1.2-1 Proposed Possession Lim its ............................................................................... 1-12

Table 3.3-1 Consequence Level as Related to AEGL Values and Soluble Uranium Intake .... 3-12

Table 3.3-2 Chemical Consequence Levels Proposed for Hydrogen Fluoride ....................... 3-13

Table 3.3-3 Chemical Consequence Levels Proposed for Uranium Hexafluoride .................. 3-13

Table 3.3-4 Radiological Consequence Levels Proposed by Applicant .................................. 3-14

Table 5.3-1 Comparison of the EREF Safe Values for 6 wt% 235U and the NRC

Endorsed Single Parameter Limits and Safety Factors ......................................... 5-7

Table 7.3-1 Identification of IRO FS .......................................................................................... 7-5

Table 9.3-1 Overview of Gaseous Effluent Monitoring Program ............................................. 9-10

Table 9.3-2 Types of Samples Collected under the REMP .................................................... 9-13

xiii

EXECUTIVE SUMMARY

On December 30, 2008, AREVA Enrichment Services LLC (AES or the applicant) submitted, to

the U.S. Nuclear Regulatory Commission (NRC), an application requesting a license, under

Title 10 of the Code of Federal Regulations (10 CFR) Parts 30, 40, and 70, to possess and use

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

uranium enrichment facility. The applicant proposes that the facility, known as the Eagle Rock

Enrichment Facility (EREF), be located in Bonneville County, Idaho, about 32 kilometers

(20 miles) west-northwest of the city of Idaho Falls. On April 23, 2009, AES submitted a revised

application to increase the facility's nominal capacity from 3 million separative work units

(SWU)/year to 6 million SWU/year. (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 be authorized to 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 staffs safeguards

review involved reviews of the applicant's Fundamental Nuclear Material Control Plan; 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.

In conducting its safety review, the staff assessed, among other things, whether the AES's

proposed equipment, facilities, and procedures will adequately protect public health and safety.

The staff evaluated AES's existing facility designs and procedures, which are in various stages

of completion, together with the applicant's stated commitments to complete certain design and

procedures according to criteria specified by the applicant. If the Commission issues a license,

it may contain conditions and limitations imposed to assure compliance with applicable

regulations. The determination whether there is reasonable assurance that public health and

safety will be adequately protected will be based in part on a comparison of the regulatory

requirements against existing information, applicant commitments, and conditions imposed by

the staff.

Once a license is granted (assuming the NRC's decision is to issue a license), construction of

the facility may begin. In accordance with 10 CFR 70.72(d)(2), the applicant (then licensee) will

submit to the NRC annual updates to the Integrated Safety Analysis Summary, along with a

brief summary of the changes made during the year. The NRC will review these submissions,

as well as any license amendment requests that may be submitted.

Although the applicant (then licensee) can start construction following issuance of the license, it

may not begin operation of the EREF until after it successfully completes a second step. Prior

to operation, the NRC must verify through inspection that the facility has been constructed in

accordance with the requirements of the license (see 10 CFR 70.32(k)). Only after this step is

successfully completed will the enrichment facility be allowed to begin operations.

xv
The staff used several guidance documents to evaluate the applicant's license application and

to complete the Safety Evaluation Report for the EREF. As previously mentioned, the primary

guidance document used by the staff is NUREG-1 520, "Standard Review Plan for the Review of

a License Application for a Fuel Cycle Facility." As needed, the staff used other NUREG and

Regulatory Guidance documents in its review. For the staff's review of the applicant's chemical

safety program, the staff used NUREG-1601, "Chemical Process Safety at Fuel Cycle Facilities"

and NUREG-1 513, "Integrated Safety Analysis Guidance Document" in addition to

NUREG-1520. For the staffs review of the Material Control and Accounting section of the

Safety Analysis Report (SAR), the staff used NUREG/CR-5734 "Recommendations to the NRC

on Acceptable Standard Format and Content for the Fundamental Nuclear Material Control Plan

Required for Low-Enriched Uranium Facilities," in addition to NUREG-1520. For its review of

the security-related portions of the SAR, the staff used Regulatory Guide 5.59, "Standard

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

Material of Moderate or Low Strategic Significance" and NRC Regulatory Issue Summary 2005-

22, "Requirements for the Physical Protection During Transportation of Special Nuclear Material

of Moderate and Low Strategic Significance: 10 CFR Part 73 vs. Regulatory Guide 5.59 (1983)."

NUREG-1520 was the primary document used to assist in defining the scope, level of detail,

and acceptance criteria for reviews. NUREG-1520 provides generic guidance for reviewing and

evaluating the health, safety, and environmental protection aspects of applications for licenses

to possess and use SNM in fuel cycle facilities. The principal purpose of NUREG-1520 is to

ensure the quality and uniformity of reviews conducted by the staff. Because NUREG-1520

describes the scope, level of detail, and acceptance criteria for reviews, it also serves as

regulatory guidance for applicants who need to determine what information to present in a

license application. Because NUREG-1 520 is regulatory guidance, it does not preclude

applicants from identifying portions of NUREG-1 520 that are not directly applicable or from

suggesting alternative approaches to those specified in NUREG-1 520 to demonstrate

compliance with applicable regulations. Should an applicant suggest alternative approaches,

the staff retains the responsibility to make an independent determination concerning the

adequacy of the applicant's suggested approaches.

NUREG-1520 was developed as a generic document for licensing fuel cycle facilities under

10 CFR Part 70, including fuel fabrication facilities and uranium enrichment facilities. Extensive

communication occurred with current fuel cycle licensees to ensure that all necessary safety

and environmental issues were addressed. While it is true that there are differences among

these types of plants and among the relative risks of certain hazards at different fuel cycle

facilities, hazards that will exist at the proposed EREF are similar to the types of hazards at

other fuel cycle facilities for which NUREG-1 520 was prepared. These hazards include

handling of uranium hexafluoride (UF6) cylinders; processing of UF6 as a gas, and sometimes

as a liquid; use of autoclaves or similar devices for feeding and sampling uranium; nuclear

criticality; and equipment decontamination operations.

Based on relative risk, the staff adapts NUREG-1520 to review applications for different types of

10 CFR Part 70 facilities. The relative risk of the applicant's facility depends on the specific

hazards associated with a particular technology (e.g., enrichment facility or fuel fabrication

facility). The staffs review of each type of facility focuses on those specific types of hazards.

The goal of the review is to determine whether applicable regulatory requirements are met to

ensure that an adequate level of safety is provided to protect the health and safety of the

workers, public, and the environment. Specific regulatory requirements for each type of facility

xvi
are found in the applicable sections of NRC's regulations. The staff recognizes that the types

and magnitudes of potential hazards vary greatly between the various types of facilities and

even within each type of hazard.

As part of the license application, AES submitted an Environmental Report, which is under a

parallel review by the NRC staff. The NRC staff will use the Environmental Report to prepare

the Final Environmental Impact Statement, which is expected to be issued in February 2011.

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

General Information

AES provided an adequate overview of the facility layout and a summary description of the

proposed 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.

AES also adequately described and documented the corporate identity, structure, financial

information, and the types, forms, and quantities and proposed purpose and authorized uses of

licensed materials to be permitted at the facility. The staff plans to impose a license condition

concerning funding for each incremental phase of the EREF.

AES also provided information on liability insurance. Because full liability insurance coverage

will not be provided until prior to receipt of licensed material, the staff plans to impose a license

condition for AES to provide proof of full liability insurance prior to obtaining licensed material.

AES also provided information on its plans to protect classified matter, including security

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

reproduced, transmitted, transported, and destroyed appropriately. Because a specific facility

for use and storage of classified matter has not been identified, the staff plans to impose a

license condition prohibiting AES to use, process, store, reproduce, transmit, handle, or allow

access to classified matter, except provided by applicable personnel and facility clearances.

The staff also plans to impose a license condition requiring notification to NRC once AES

designates areas for the routine use and handling of classified information.

Under 10 CFR 40.14 and 10 CFR 70.17, AES requested an exemption from 10 CFR 40.36(d)

and CFR 70.25(e) to provide forward-looking incremental funding for decommissioning. The

staff will grant the exemption and impose a license condition to address AES's commitments for

updating the decommissioning funding plan over time.

In addition, AES requested special authorization to make changes to the license application that

do not decrease the effectiveness of its safety commitments in the license application without

prior NRC approval. The staff will grant the authorization and impose a license condition for

address the criteria, documentation, and reporting of changes made to the license application

without prior NRC approval.

AES also adequately described and summarized information concerning site geography, nearby

population, meteorology, hydrology, and geology, and potential effects of natural phenomena at

the facility.

xvii
Orgqanization and Administration

AES adequately described the applicant's organization, key management positions,

qualifications, and management controls-including 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 SAR provide reasonable assurance

that an acceptable organization, administrative policies, and sufficient, competent resources

have been or will be established in such a manner as will allow for safe operation of the facility.

Integrated Safety Analysis (ISA) and ISA Summary

AES provided sufficient information about the facility, facility processes, hazards, and types of

accident sequences. Area boundaries, including the controlled area boundary and the locations

of restricted areas, are adequately described. AES commits to maintaining process safety

information and to conducting an ISA. AES's ISA methodology uses appropriate methods for

identifying potential accidents, determining consequences, and deriving likelihood. The

likelihood evaluation, including the definition of "not credible," and the determination of chemical

consequences are acceptable. AES has addressed the baseline design criteria and defense-indepth practices required of new facilities.

The staff will impose a license condition to ensure that the boundaries of the items relied on for

safety (IROFS) are developed and available to staff prior to the operational readiness review.

The staff will also impose a license condition concerning the conduct of human factors

engineering reviews of the human-system interfaces supporting IROFS requiring operator

actions. And, the staff will impose a license condition addressing the design of IROFS that may

use software, firmware, microcode, programmable controllers, and/or any digital device.

AES's ISA methodology and ISA Summary meet the requirements of 10 CFR Part 70.

Radiation Protection

AES provided sufficient information to evaluate its Radiation Protection Program. The SAR

adequately describes: (a) the program for ensuring that worker and public doses are as low as

(is) reasonably achievable (ALARA); (b) the qualification requirements for radiation protection

personnel; (c) AES's commitment to approved, written radiation protection procedures and

radiation work permits; (d) the training for all personnel who have access to restricted areas;

(e) the program to control airborne concentrations of radioactive material with engineering

controls and respiratory protection; (e) the radiation survey and monitoring program; and

(f) recordkeeping. AES's Radiation Protection Program meets the requirements of

10 CFR Parts 19, 20, 30, 40, and 70.

Nuclear Criticality Safety

AES provided sufficient information to evaluate its Nuclear Criticality Safety (NCS) Program.

The SAR adequately describes: (a) AES's commitment to have a group of qualified staff to

develop, implement, and maintain the NCS Program; (b) AES's NCS methodologies and

technical practices; and (c) the criticality accident alarm system. AES's NCS methodologies

and technical practices will provide an adequate margin of subcriticality, and the safety

programs and management measures will ensure that the margin of subcriticality is maintained.

AES's NCS Program meets the requirements of 10 CFR Part 70.

xviii
Chemical Process Safety

AES adequately described and assessed accident consequences that can result from the

handling, storage, or processing of licensed materials that can potentially have significant

chemical consequences and effects. AES performed hazard analyses that identified and

evaluated those chemical process hazards and potential accidents and established safety

controls providing reasonable assurance of safe facility operation. AES's plan for managing

chemical process safety and chemical process safety controls meets the requirements for

10 CFR Part 70.

Fire Safety

AES provided sufficient information to evaluate the potential fire hazards, consequences, and

required controls for the proposed processes. The applicant identified a reasonable set of

controls and defense-in-depth protection to meet the performance requirements of

10 CFR 70.61 and the baseline design criteria and defense-in-depth practices required of new

facilities.

Emergqency Managqement

AES provided an adequate Emergency Plan for the facility that meets the regulatory

requirements. The facility will be properly configured to limit releases of radioactive materials in

case of an accident; the capability will be established for measuring and assessing the

significance of accidental releases of radioactive materials; appropriate emergency equipment

and procedures will be provided; a system will be established to notify Federal, State and

County government agencies; and recovery procedures will be established to return the facility

to a safe condition after an accident. AES commits to implement the Emergency Plan through

approved, written procedures.

Environmental Protection

AES has developed a program to implement adequate environmental protection measures

during operation, including: (a) environmental and effluent monitoring and (b) effluent controls to

maintain public doses ALARA as part of the Radiation Protection Program. The applicant's

program is adequate to protect the environment and the health and safety of the public, and

complies with the regulatory requirements in 10 CFR Parts 20, 30, 40, 51, and 70.

Decommissioning

Under 10 CFR 70.38(i), AES requested an alternate schedule for decommissioning since it

expects decommissioning will take about eight years. 10 CFR 70.38(h) requires that

decommissioning be completed no later than 24 months following the initiation of

decommissioning, unless the Commission approves an alternate schedule. The staff reviewed

the request in light of the considerations listed in 10 CFR 70.38(i)(1)-(i)(5) and determined that

the alternative schedule is warranted.

AES provided a conceptual decontamination and decommissioning plan including the initial site

characterization data and plans to collect and analyze additional samples to determine a

background value for the site. The staff will impose a license condition to require AES to

implement its plans to collect and analyze additional samples.

xix
The applicant submitted an adequate decommissioning funding plan cost estimate that is

consistent with NUREG-1557, Volume 3, and satisfies the regulatory requirements of 10 CFR

Parts 30, 40, and 70. The applicant also provides a draft letter of credit, standby trust

agreement, and supporting documentation. The staff will impose a license condition concerning

AES's method of providing decommissioning financial assurance on an incremental, forwardlooking basis.

Management Measures

AES has adequately described the management measures that will be applied to the proposed

facility. The information describes: (a) the organization responsible for developing,

implementing, and assessing the management measures; (b) the commitment to quality

assurance elements and administrative measures for staffing, performance, assessing findings,

and implementing corrective actions; (c) the process for the development, approval, and

implementation of procedures; (d) the commitment to establishing and documenting

surveillances, tests, and inspections to provide reasonable assurance of satisfactory

performance of the IROFS; (e) the audit program; and (f) training requirements. AES's quality

assurance staff has the independence and authority to carry out their function without undue

influence and quality assurance elements cover the IROFS. The proposed management

measures meet the regulatory requirements of 10 CFR Part 70.

Physical Protection

AES provided an adequate Physical Security Plan including information on the policies,

methods, and procedures to be implemented to protect SNM of low strategic significance used

and possessed at the facility. In the Physical Security Plan, AES also provided information on

the policies, methods, and procedures to be implemented to protect SNM of low strategic

significance in transit to and from the facility. The Physical Security Plan is acceptable and

meets the requirements in Part 73.

Materials Control and Accountability

AES provided an adequate Fundamental Nuclear Material Control (FNMC) Plan for the

proposed facility that will meet the applicable 10 CFR 74 requirements. The staff will impose a

license condition requiring AES to maintain and follow the FNMC Plan.

xx
ACRONYMS AND ABBREVIATIONS

ac acres

ACI American Concrete Institute

AEC active engineered control

AEGL Acute Exposure Guideline Level

AES AREVA Enrichment Services LLC

AISC American Institute of Steel Construction

ALARA as low as is reasonably achievable

Al 203 aluminum oxide

ANS American Nuclear Society

ANSI American National Standards Institute

ARF airborne release fraction

ASCE American Society of Civil Engineers

ASME American Society of Mechanical Engineers

ASTM American Society of Testing and Materials

BDC baseline design criteria

BHS Bureau of Homeland Security

BLM Bureau of Land Management

Bq Becquerel

BSPB Blending, Sampling and Preparation Building

BTP branch technical position

0

C degree(s) Celsius

CAA controlled access area

CAAS criticality accident alarm system

CAB Centrifuge Assembly Building

CAP Corrective Action Program

CEDE Committed Effective Dose Equivalent

CFR Code of Federal Regulations

Ci curie

cm centimeter

CM configuration management

CMP Classified Matter Plan

CRSB Cylinder Receipt and Shipping Building

DBE Design Basis Earthquake

DFP Decommissioning Funding Plan

DOE U.S. Department of Energy

DOT U.S. Department of Transportation

DR damage ratio

DU depleted uranium

EBR Argonne National Lab- West

EHS&L Environmental, Health, Safety and Licensing

EIS Environmental Impact Statement

EOC Emergency Operations Center

EP Emergency Plan

EPA U.S. Environmental Protection Agency

xxi
EPIP emergency plan implementing procedure

ER Environmental Report

EREF Eagle Rock Enrichment Facility

ERO emergency response organization

ETC Enrichment Technology Company Limited

OF degree(s) Fahrenheit

FA financial assurance

FHA Fire Hazard Analysis

FNMC Fundamental Nuclear Material Control

FOCI foreign ownership, control, or influence

FTC full tails cylinder

ft feet

g gram

gal gallon

GEVS gaseous effluent ventilation system

gpm gallons per minute

ha hectare

HAZOP hazard and operability

HEPA high efficiency particulate air

HF hydrogen fluoride

HFE human factors evaluation

HPS Health Physics Society

HS&E Health, Safety, and Environment

H/U hydrogen/uranium ratio

HVAC Heating, Ventilating, and Air Conditioning

H20 Water

IBC International Building Code

ICBO International Conference of Building Officials

in inch

INL Idaho National Laboratory

INTEC Idaho Nuclear Technology and Engineering Center

IROFS items relied on for safety

ISA Integrated Safety Analysis

ISO International Organization for Standardization

IT information technology

kg kilogram

km kilometer

L liter

lb pound

LCS local control system

LEL lower explosive limit

LES Louisiana Energy Services

LLW low-level waste

LOC letter of credit

xxii
LPF

Ipm

LSS

LTTS

m

m 3

MAR

MBq

MC&A

MFC

mg

mi

ml

mm

MOU

Mph

m/s

mrem

Ms

mSv

MT

Mw

leak path factor

liter per minute

low strategic significance

Low Temperature Take-off Station

meter

cubic meter

material at risk

megabecquerel

material control and accounting

Materials and Fuels Complex

milligram

mile

milliliter

millimeter

Memorandum of Understanding

miles per hour

meter per second

milli Roentgen equivalent man

surface wave magnitude

millisievert

metric tonne

momentum magnitude

N/A

NaF

NAVFAC

NCS

NEI

NELAC

NELAP

NESHAPS

NFPA

NIOSH

NIST

NMSS

NOAA

NRC

OJT

OSB

OSHA

PAR

PFPE

PM

PNNL

PSC

psia

psig

not applicable

sodium fluoride

Naval Facilities Engineering Command

Nuclear Criticality Safety

Nuclear Energy Institute

National Environmental Laboratory Accreditation Conference

National Environmental Laboratory Accreditation Program

National Emission Standards for Hazardous Air Pollutants

National Fire Protection Association

National Institute for Occupational Safety and Health

National Institutes of Standards and Technology

Office of Nuclear Material Safety and Safeguards

National Oceanic and Atmospheric Administration

U.S. Nuclear Regulatory Commission

on-the-job training

Operation Support Building

U.S. Occupational Safety and Health Administration

protective action recommendation

perfluorinated polyether

preventive maintenance

Pacific Northwest National Laboratory

Process Service Corridor

pounds-force per square inch absolute

pounds-force per square inch gauge

xxiii
PSHA probabalistic seismic hazard assessment

PSP Physical Security Plan

QA quality assurance

QAPD Quality Assurance Program Description

QC quality control

RAI request for additional information

RASCAL Radiological Assessment System for Consequence Analysis

rem Roentgen equivalent man

REMP Radiological Environmental Monitoring Program

RF respirable fraction

RG Regulatory Guide

RP radiation protection

RWP Radiation Work Permit

SAR Safety Analysis Report

SBM Separations Building Module

SER Safety Evaluation Report

SNM special nuclear material

SOP Standard Operating Procedure

SPCC spill prevention, control and countermeasures

SPPP Standard Practice Procedures Plan

SRC Safety Review Committee

SRP Standard Review Plan

SSC structures, systems, and components

ST source term

Sv sievert

SWU separative work unit

Tc technetium

TEDE total effective dose equivalent

TLD thermoluminescent dosimeter

TSB Technical Support Building

TSP Transportation Security Plan

TWA time-weighted average

U uranium

pCi microcurie

pg microgram

UF4 uranium tetrafluoride

UF6 uranium hexafluoride

UHRS uniform hazard response spectra

U0 2F2 uranyl fluoride

UPS uninterruptible power supply

USEC U.S. Enrichment Corporation

USGS U.S. Geological Survey

USL upper subcritical limit

wt weight

xxiv
yd yard

yd3 cubic yard

yr year

xxv

CHAPTER 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 applicant's

facility and process description is to evaluate 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 (AES, 2010b).

1.1.1 Regulatory Requirements

The regulations in Title 10 of the Code of Federal Regulations (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 the applicant will use to protect public health and

safety 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 Guidance and Acceptance Criteria

The guidance applicable to the NRC's review of the facility and process description section of

the Safety Analysis Report (SAR) (AES, 2010a) is contained in Chapter 1 of "Standard Review

Plan for the Review of a License Application for a Fuel Cycle Facility," NUREG-1 520 (NRC,

2002). This chapter is applicable in its entirety. The acceptance criteria applicable to this

review are contained in Section 1.1.4.3 of NUREG-1 520 (NRC, 2002).

1.1.3 Staff Review and Analysis

In Section 1.1 of the SAR (AES, 2010a), the applicant, AREVA Enrichment Services (AES),

provided a summary description of the proposed gas centrifuge uranium enrichment facility and

processes. The summary description includes discussion of the major chemical and

mechanical processes to be used in the facility.

The applicant is proposing to use a gas centrifuge enrichment process to enrich natural uranium

by separating a feed stream containing a natural isotopic concentration of about 0.7 weight

percent uranium-235 (U-235) into a product stream enriched in U-235 at 5 weight percent and a

tails stream depleted in U-235. The proposed plant, known as the Eagle Rock Enrichment

Facility (EREF), will have a plant design capacity of 6.6 million SWUs per year (AES, 201 Oa).

(A SWU is a measure of the effort required to perform isotopic separation.) The plant design

capacity accounts for production margin for centrifuge failures and occasional production losses

during the operational lifetime of the facility, thus the nominal production capacity will be

6 million SWUs per year (AES, 201 Oa).

The feed material for the enrichment process will be uranium hexafluoride (UF6) with a natural

composition of isotopes U-234, U-235, and U-238 (AES, 2010a). The enrichment process will

1-1
be a mechanical separation of isotopes using a fast rotating cylinder (centrifuge) based on a

difference in centrifugal forces. Gaseous natural UF6 will enter a high-speed centrifuge.

Centrifugal forces press the heavier isotope of uranium, U-238, to the outer wall of the

centrifuge. The lighter isotope, U-235, will remain closer to the center, away from the wall of the

centrifuge. Internal scoops will be used to collect the heavier and lighter fractions and circulate

them to other centrifuges piped in a cascade arrangement. No chemical changes or nuclear

reactions will take place. The feed, product, and tails streams will all be in the form of UF6.

The applicant has proposed that the EREF be located in Bonneville County, Idaho

approximately 32 kilometers (km) (20 miles [mi]) west northwest of the city of Idaho Falls. The

facility will be located on piece of property of about 1,700 hectares (ha) (4,200 acres [ac]) (AES,

2010a). The footprint of the facility will be about 172 ha (426 ac) in the north central portion of

the plot (AES, 201 Oa).

The property is currently privately-held by a single landowner (AES, 2010a). This privately-held

land will be purchased by AES. Within the proposed site, there is also a 16 ha (40 ac) parcel

which is administered by the U.S. Bureau of Land Management (AES, 2010a).

The major structures of the facility are described in Section 1.1.2 of the SAR (AES, 201 Oa). The

EREF facility will consist of multiple buildings, each of which will perform a specific function. A

list of selected buildings and their specific functions follows:

Separations Building Modules (SBMs): The EREF will include four identical SBMs.

Each module will consist of two cascade halls. Each hall will have twelve cascades with

each cascade having hundreds of centrifuges.

Technical Support Building (TSB): This building will contain the radiological support

areas for the EREF, including solid waste collection, workshops, liquid effluent collection

and treatment, laundry sorting, laboratories, radiation monitoring, truck bay/shipping and

receiving, ancillary areas, and maintenance for contaminated facility equipment.

Operation Support Building (OSB): This building will contain the non-radiological

support for areas for the EREF, including workshops, medical, lockers, cafeteria, lobby,

ancillary areas, control room, training, security alarm system, and environmental

laboratory.

Centrifuge Assembly Building (CAB): This building will be used to assemble centrifuges

before they are moved into the SBMs.

Administration Building: This building will include general office areas.

Security and Secure Administration Building: This building will contain secure office

areas and the Entry-Exit Control Point for the EREF.

Guard House: This building will function as the security checkpoint for all incoming and

outgoing traffic.

Cylinder Receipt and Shipping Building: This building will be used to receive, inspect,

weigh, and temporarily store cylinders of feed UF 6; to temporarily store, inspect, weigh,

1-2
and ship cylinders of enriched UF 6; to receive, inspect, weigh, and temporarily store

empty and depleted uranium tails cylinders; and to transfer filled depleted uranium

cylinders to the Full Tails Cylinder Storage Pads.

Blending, Sampling,-and Preparation Building: This building will be used to fill cylinders

and to liquefy, homogenize, and sample cylinders prior to shipment to customers.

Cylinder Storage Pads: Outside storage areas will be used for the storage of full

cylinders containing UF6 and of empty cylinders. Feed cylinders, containing natural UF6,

will be temporarily stored on the Full Feed Cylinder Storage Pads, prior to use in the

facility. Tails cylinders, containing depleted UF 6, will be temporarily stored on the Full

Tails Cylinder Storage Pads. Product cylinders, containing enriched UF 6, will be stored

on the Full Product Cylinder Storage Pad, prior to shipment offsite to a fuel fabrication

facility. Empty cylinders will be temporarily stored on the Empty Cylinder Storage Pads.

The Full Feed, Full Tails, Full Product, and Empty Cylinder Storage Pads will be sized to

store approximately 712 feed cylinders, 25,718 tails cylinders, 1,032 product cylinders,

and 1,840 empty cylinders, respectively.

Electrical Services Building: This building will house four standby diesel generators, day

tanks, switchgear, control panels, and building heating, ventilation, and air conditioning

(HVAC) equipment.

Mechanical Services Buildings: There will be two mechanical services buildings; and

they will house air compressors, the demineralized water system, centrifuge cooling

water system pumps, heat exchangers, and expansion tanks.

Electrical Services Building for the CAB: This building will house four transformers and

switchgear.

Gasoline and Diesel Fuel Station: The station will include a fuel pump island for vehicle

fueling and a building for on-site vehicle repair and maintenance.

Visitor Center: This building will be located outside the security fence area.

Natural uranium feed will be shipped to the EREF by truck in cylinders having a nominal

capacity of 12.5-metric tonnes (MT) (14-ton) of UF6. Under ambient conditions, the UF 6 is a

solid. The feed cylinders will be loaded into the solid feed stations, vented to remove the light

gases-primarily air and hydrogen fluoride (HF)-and then heated to sublime the solid UF 6 to a

gas. The light gases and UF6 gas generated will be routed to the feed purification subsystem to

remove any light gases from the UF6 feed prior to introduction into the cascades. After

purification, the gaseous UF 6 from the solid feed stations will be directed to a cascade for

enrichment.

After enrichment in the cascade, both depleted and enriched UF6 will be withdrawn and

desublimed at subatmospheric pressure in the tails take-off system and the product take-off

system, respectively. Tails and product take-off systems will be designed to preclude UF6 from

becoming a liquid. The product take-off system will also contain a system to purge light gases.

1-3
Sampling to verify the product assay level will be performed in the product liquid sampling

system. In an autoclave, UF6 will be heated to a liquid; the cylinder will be tilted so that UF6 can

flow into sample manifold and sample bottles; and the cylinder will be returned to its original

horizontal position. This is the only system in the plant where UF 6 will be in a liquid form.

To produce enriched uranium meeting customer-assay specifications, a product blending

system will be used to mix enriched uranium at two different enrichment levels to meet the

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

The SAR (AES, 201 Oa) and the ISA Summary (AES, 2010b) provide additional descriptions and

process details, including drawings of the plant buildings and the location of plant systems

within the buildings.1 Geographical features of the site are also provided on these drawings.

The EREF will possess natural, enriched, and depleted uranium. The feed, product, and tail

streams will all be in the form of UF6. At full capacity, the EREF will handle, on an annual basis,

1,424 nominal 12.5-MT (14-ton) natural uranium feed cylinders; 1,032 nominal 2-MT (2.5-ton)

enriched uranium product cylinders; and 1,222 nominal 12.5-MT (14-ton) depleted uranium tails

cylinders (AES, 2010a).

Solid waste will be generated at the EREF, including, Class A low-level radioactive waste, lowlevel mixed waste, hazardous waste, and industrial (non-hazardous) waste (AES, 2010a). The

SAR provides estimates of the annual quantities generated (AES, 2010a). Table 1.1-2 of the

SAR provides estimates of the annual quantities of radiological and mixed wastes produced,

including uranium content. Table 1.1-4 of the SAR provides estimates of the annual quantities

of non-radiological wastes produced. Construction wastes will also be generated during

construction of the facility. Table 1.1-5 of the SAR provides estimates of the annual hazardous

wastes produced during construction. All low-level radioactive waste and mixed waste will be

disposed of at a licensed low-level waste disposal facility (AES, 201 Oa). All hazardous chemical

wastes will be transported to permitted treatment and disposal facilities (AES, 201 Oa).

Depleted UF6 tails will be generated at the EREF. Table 10.3-1 of the SAR provides estimates

of the tails production during the operating life of the EREF. AES intends to transfer depleted

UF6 to the U.S. Department of Energy (DOE) for conversion and disposal. Depleted UF6 tails

will be stored onsite in cylinders on the cylinder storage pad until the cylinders are transferred to

the DOE (AES, 2010a). Once removed, depleted UF 6 tails will be converted to a stable form

and disposed of in accordance with the United States Enrichment Corporation Privatization Act

and other statutory authorizations and requirements at DOE's conversion facility (AES, 201 Oa).

Gaseous airborne effluents will be released from the EREF. Table 1.1-1 of the SAR provides

estimates of the annual gaseous effluents. The applicant estimates that less than 20 grams

(g) (0.0441 pounds [Ib]) of uranium and less than 2 kilogram (kg) (4.4 Ib) of HF will be released

annually in 4.13 x 109 cubic meters of air discharge. Sources of air emissions include the

gaseous effluent ventilation system and the HVAC systems. As discussed in Chapter 9 of this

SER, these emissions will be significantly below the limits in 10 CFR Part 20 and the As Low as

Reasonably Achievable (ALARA) program air effluent goals.

The applicant does not anticipate any liquid discharges of licensed radioactive materials from

the EREF. The applicant will collect liquid effluent from plant processes and treat the collected

1Drawings of the facility layout, plant buildings, and plant systems within the buildings as well as the ISA Summary

have been withheld from public release as security-related information under 10 CFR 2.390.

1-4
effluent to remove the uranic material (AES, 2010a). Table 1.1-3 of the SAR provides estimates

of the annual liquid effluent collected from plant processes (AES, 2010a). The applicant

estimated that a total of about 57,100 liters (L) /year (yr) (15,625 gallons [gal]/yr) of liquid will be

generated and treated, containing a total of 114 kg (251 Ib) of uranic material. Sources of the

liquid effluent to be collected include laboratory effluent, floor washings, condensates,

degreaser water, and spent citric acid. The Liquid Effluent Collection and Treatment System for

the EREF will employ precipitation and filtration to remove uranic material. The final stage of

evaporation will release the treated distillate directly into the atmosphere. The applicant

estimated the amount of uranic material in the evaporated distillate to be less than

0.04 grams (g)/yr of total uranium. The uranium removed during treatment will be disposed of

offsite at a licensed disposal facility. As discussed in Chapter 9 of this SER, these emissions

will be significantly below the limits in 10 CFR Part 20 and the ALARA program air effluent

goals.

As described above, the applicant provided information at a level of detail that is appropriate for

general familiarization and understanding of the proposed facility and processes. The

application summarizes the facility information contained in the ISA Summary and includes

descriptions of the overall facility layout on scaled drawings, including the site's geographical

features and facility structural features. The applicant's summary also describes the

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

The major chemical and mechanical processes involving licensable material are described in

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

includes: (a) reference to the building locations of major process components; (b) brief

descriptions of the process steps; (c) the chemical forms of licensable material in process; and

(d) the types, amounts and discharge points of waste materials discharged to the environment

from the processes. The applicant presented a summary of the feed materials, by-products,

waste, and finished products of the facility. Information concerning expected levels of trace

impurities or contaminants appears in Table 1.2-1 of the SAR and is discussed in

Section 1.2.3.4 of this SER. The information the applicant provided meets the guidance in

Section 1.1.4.3(1), (2), (3), and (4) of NUREG-1 520 (NRC, 2002) and is, therefore, acceptable.

1.1.4 Evaluation Findings

The staff has reviewed the proposed general facility and process descriptions for the EREF

according to Section 1.1 of NUREG-1520 (NRC, 2002). 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 the applicant's institutional information is to evaluate whether

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

characteristics, and the proposed activity.

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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 for financial qualifications are found in 10 CFR 70.22(a)(8) and

10 CFR 70.23(a)(5). 10 CFR 70.22(a)(8) requires an applicant to submit information with

respect to its financial qualifications, and 10 CFR 70.23(a)(5) requires that an applicant be

financially qualified to engage in the proposed activities in accordance with the regulations. In

addition, the staff took into consideration the Commission's ruling in Louisiana Energy Services,

L.P. (Claiborne Enrichment Center), CLI-97-15, 46 NRC 294 (1997), which pertains to an

application by Louisiana Energy Services (LES) to construct and operate a uranium enrichment

facility pursuant to 10 CFR Part 70. Among other things, the ruling held that "the NRC is not

required as a matter of law to apply the strict financial qualification provisions of Part 50 to all

Part 70 license applications." Id., 46 NRC at 298. In summary, the Commission concluded that

"the general language of Part 70 leaves the Commission free to review the reasonableness of

an applicant's financial plan in light of all relevant circumstances," which might or might not lead

to application of any or all of the criteria stated in Part 50. Id. at 302.

The staff also considered other regulatory requirements, including the following. The

regulations found in 10 CFR 40.38 and 10 CFR 70.40 place restrictions on the eligibility of

certain applicants to hold licenses. The regulations in 10 CFR 70.22(m) identify the

requirements to protect against unauthorized viewing of classified enrichment equipment.

10 CFR Part 95 contain provisions to protect against unauthorized disclosure of classified

matter and requirements for obtaining a facility security clearance. The regulations in

10 CFR 140.13(b) require applicants for uranium enrichment facilities to provide and maintain

liability insurance.

1.2.2 Regulatory Guidance and Acceptance Criteria

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

application are contained in 10 CFR 30.32, 10 CFR 40.31, 10 CFR 40.38, 10 CFR 70.22, 10

CFR 70.23(a)(5), 10 CFR 70.40, 10 CFR Part 95, 10 CFR 140.13b, and Section 1.2.4.3 of

NUREG-1520 (NRC, 2002). Chapter 1 of NUREG-1 520 is applicable to the EREF in its entirety.

Section 1.2.3.6, "Special Exemptions or Special Authorizations," of this chapter, addresses

exemptions and special authorizations.

In addition, the EREF classified matter plan, "Security Plan for the Protection of Classified

Matter" was reviewed for compliance with the requirements of 10 CFR Part 95 by using

"Standard Practice Procedures Plan Standard Format and Content for the Protection of

Classified matter for NRC Licensee, Certificate Holder, or Other Activities as the Commission

May Determine" (NRC, 2006).

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1.2.3 Staff Review and Analysis

1.2.3.1 Corporate Identity

In Section 1.2.1 of the SAR (AES, 2010a), AES, provided information on the corporate identify

and location and ownership organization.

AES is a Delaware limited liability corporation with a principal location of business in Bethesda,

Maryland. It has been formed solely to provide uranium enrichment services for commercial

nuclear power plants. AES is a wholly owned subsidiary of AREVA NC Inc.

AREVA NC Inc. is a wholly owned subsidiary of AREVA NC SA, which is part of AREVA SA

(AREVA), a corporation formed under the laws of France. The principal owners of AREVA SA

are identified in Section 1.2.1.2 of the SAR (AES, 2010a) and include the Commissariat a

I'Energie Atomique (French Atomic Energy Commission) and the French State.

AES is governed by the AES Management Committee. Section 1.2.3.2 of the SAR (AES,

2010a) identifies the Committee members. The President and Chief Executive Officer of AES is

a naturalized citizen of the United States, as well as a citizen of Canada. Any safety decision

related to the operation of the facility will be made by the President of AES.

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

than where the applicant has contracted such services.

The AES principal office is located at 4800 Hampden Lane, Bethesda, Maryland 20814.

The EREF will be located in Bonneville County, Idaho, along State Highway 20. The following is

the legal description of the EREF site location, as described in the SAR (2010a):

"All of Sections 13, 14 and 15; the Northeast quarter (NE1/4) of Section 21; the North

half (N1/2), and Southeast Quarter of the Southeast Quarter (SE1/4 SE1/4) of Section

22; the North Half (N1/2), the Southeast Quarter (SE1/4), the East Half of the Southwest

Quarter (E1/2 SW1/4), and the Southwest Quarter of the Southwest Quarter (SW1/4

SW1/4) of Section 23; West Half (W1/2), and the West Half of the Southeast quarter

(W1/2 SE1/4), and the Northeast quarter of the Southeast quarter (NE1/4 SE1/4) and

the Northwest quarter of the Northeast quarter (NW1/4 NE1/4) of Section 24; the West

1/2 (W1/2) of Section 25, Less the Highway and that portion of the SW1/4 deeded to the

State of Idaho in a Warranty Deed recorded July 25, 1950, in Book 72 of Deeds, at page

565 and the Northeast quarter (NE1/4); the East Half of the Northwest Quarter (E1/2

NW1/4), the Northeast Quarter of the Southwest Quarter (NE1/4 SW1/4), the Northwest

Quarter of the Southeast Quarter (NW1/4 SEl/4) and that portion of the South Half of

the Southeast Quarter (S1/2 SE1/4) lying north of the centerline of State Highway 20 as

surveyed and shown on the official plat of the Twin Buttes F-1422(2) Highway Survey on

file in the office of the Department of Highway of the State of Idaho, all in Section 26; All

in Township 3 North, Range 34 East of the Boise Meridian, Bonneville County, Idaho,

contains four thousand two hundred and ten (4,210) acres, more or less."

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As stated above, the applicant furnished its full name and address, a full description of the

proposed facility site location, the state where the corporation is incorporated, and the location

of the principal offices. The applicant also described the corporation's control and ownership,

including any control or ownership exercised by a foreign entity. The applicant provided

information on primary ownership and relationships to other components of the organization of

the same ownership. The applicant described the presence and operations of any other

organization on the site. The information the applicant provided meets the guidance in

Section 1.2.4.3(1) of NUREG-1 520 (NRC, 2002) and is, therefore, acceptable.

1.2.3.2 Foreign Ownership, Control, or Influence

With respect to foreign ownership, control or influence (FOCI) for AES's EREF, the NRC staff

has determined that any FOCI mitigation measures placed on AES would provide no additional

benefit to the National Security of the United States. The staff believes that the

recommendation made by the DOE to NRC by letter dated' March 31, 2005 (letter has been

withheld from public release as "Official Use Only") regarding waiving the FOCI requirement for

FOCI mitigation associated with the granting of a facility security clearance to URENCO for the

National Enrichment Facility, applies to AES's EREF. This is based on the fact that both

URENCO and AES use, or will use, the same classified technology supplied by the Enrichment

Technology Company Limited (ETC). The NRC decision is based on an Interagency

Agreement between NRC and DOE dated May 6, 2002 (DOE, 2002). The following

summarizes the basis for the decision to waive the requirement for FOCI mitigation for AES.

The information and technology that will be classified as restricted data in the United States are

already owned and controlled by the European Governments and the foreign-controlled

companies associated with URENCO and AREVA. This information and technology are only

being classified under U.S. law by the fact that it's being introduced into the United States. The

NRC has considered whether the operation of the EREF could generate new restricted data, not

otherwise available to the European Governments or their nationals. All of the parties

associated with URENCO and AREVA agree that little, if any, new restricted data should be

created as a result of the AES facility. In addition, the Pentapartite Agreement (an agreement

under development between the United States and four European Governments) will establish

protocols to be followed if any new restricted data is created at the EREF and would prevent

new restricted data from being disseminated to European nationals.

Thus, the staff finds that any additional FOCI mitigation measures placed on AES would provide

no additional benefit to the National Security of the United States. Under 10 CFR 95, AES is

required to complete the NRC facility clearance process which entails, among other things, an

NRC-approved classified matter plan (CMP), referred to as a Standard Practice Procedures

Plan; on-site inspection; and granting of individual NRC personnel security clearances. Section

1.2.4.3 of this SER evaluates AES's CMP for the protection of classified matter at the proposed

EREF facility.

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1.2.3.3 Financial Qualifications

1.2.3.3.1 Evaluation of Cost Estimate to Construct and Operate EREF

In Section 1.2.2 of the SAR (AES, 201 Oa) and in supplemental information provided on

September 28, 2009 (AES, 2009a), AES estimated the total construction cost of EREF to be

approximately $4.1 billion (2007 dollars) excluding escalation, contingency, interest, tails

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

facility. AES asserted that investment in EREF at 6.6 million SWU will be divided into at least

two (3.3 million SWU) phases. According to AES, the first phase may be broken down into

smaller increments based on calendar time or construction phase (AES, 2009a). The second

phase of construction may vary as well. AES stated that, depending on market conditions, the

second 3.3 million SWU may be constructed in one or more phases timed with demand

(AES, 2009a). AES has provided proprietary information regarding the estimated costs of the

two phases. This proprietary information is discussed in Appendix C to this SER.

As part of the cost estimate analysis, the staff reviewed Section 1.1.2, "Facilities Description," of

the SAR to ensure the support structures and systems required for operation were addressed

and compared this information to NUREG-1 827, "Safety Evaluation Report for the National

Enrichment Facility in Lea County, New Mexico" (NRC, 2005) to ensure consistency. After this

review, the staff finds that the EREF construction cost estimate encompasses the EREF

structures and systems and is consistent with the basis of the cost estimate for a similar facility

and therefore, is reasonable.

Furthermore, in Section 1.2.2 of the SAR (AES, 201 Oa), AES states that it will make available

updated cost estimates for each phase prior to initiating construction. In Section 1.2.2 of the

SAR, AES states that:

Construction of each incremental phase of the EREF shall not commence before funding

for that increment is available or committed. Of this funding, AES must have in place

before constructing such increments, commitments for one or more of the following:

equity contributions from AES or its parents, a commitment from the parent company to

provide the necessary funds for the project, and lending and/or lease arrangements that

solely or cumulatively are sufficient to ensure funding for the particular increment's

construction costs. AES shall make available for NRC inspection, documentation of both

the budgeted costs for each incremental phase and the source of funds available or

committed to pay those costs.

The staff will impose this approach as a license condition.

1.2.3.3.2 Financial Information

In Section 1.2.2 of the SAR (AES, 201 Oa), AES proposed to satisfy the obligation to

demonstrate the financial qualifications to carry out the proposed activities, as set forth in

10 CFR 70.23(a)(5) in a manner consistent with the approach previously accepted by the NRC

staff in Section 1.2.3.3.2 of NUREG-1851, "Safety Evaluation Report for the American

Centrifuge Plant in Piketon, Ohio" (NRC, 2006).

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AES plans to fund Phase 1 of the project with a mix of approximately 62 percent debt and 38

percent equity from AES parent contributions and self-generated cash during pre-production.

The remaining 62 percent of the estimated $3.2 billion Phase 1 construction costs will be

financed through financial institutions (AEs, 2009a). The estimated construction costs for Phase

2 of the project are expected to be financed with a mix of approximately 70 percent debt from

financial institutions and 30 percent equity from AES parent contributions and commitments and

cash flow from AES operation (AES, 2009a). Furthermore, as discussed in Section 1.2.3.3.1 of

this SER, AES has asserted that it shall make available for NRC inspection, documentation of

the source of funds available or committed to pay costs prior to commencement of construction

of each phase (AES, 2009a).

Additionally, AES stated that, as of October 2009, 55 percent of EREF Phase 1 output is

committed from 2014 to 2028 with eight major U.S. utilities with 44 percent already contracted;

and the remaining 11 percent committed through letters of intent and contract negotiations in

the final stages. Furthermore, AES expects to have an additional 35 percent (for a total of 90

percent of Phase l's output) contracted before the expected start of construction in 2011, with

terms sufficient to guarantee a reasonable return on investment for the entire term of the offtake agreements (AES, 2009a).

AES has no reported income statements since its inception. However, AES' parent company,

AREVA, has the assets necessary to provide the aforementioned equity contribution. As

described in Section 1.2.3.3.1 of this SER, AES has proposed an approach for funding which

the NRC staff will impose as a license condition. The proposed approach for funding includes a

statement that equity contributions will be provided by either AES or its parent. In 2008, AREVA

had total assets in excess of $48 billion, total revenue of more than $18 billion, a net income of

$702 million, and cash and cash equivalents of $1.48 billion (AES, 2009a). Therefore, AES

identified sources of debt and equity for construction, and additional financial resources if

necessary.

The NRC staff finds that, based on the financial information submitted in the SAR (NRC, 2010)

and RAI response (AES, 2009a), AES meets the financial qualifications for the proposed

activities in accordance with 10 CFR 70.23(a)(5).

1.2.3.4 Liability Insurance

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

cover public claims arising from any occurrence, within the United States that causes, within or

outside the United States, 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.

According to the application, by letter dated December 22, 2008, the American Nuclear

Insurers documented its expectation to provide nuclear liability insurance for EREF at a

maximum policy amount of $300 million by the time AES takes possession of source

material or special nuclear material (SNM). Because AES states in the SAR (AES,

2010a) that it will provide proof of, and maintain, nuclear liability insurance in the

maximum available amount, the staff finds that the applicant satisfies the regulatory

requirements under 10 CFR 140.13(b).

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Because the liability insurance coverage will not be provided until AES takes possession of

source material or SNM, NRC staff is imposing the following license condition:

The licensee shall provide proof of full liability insurance of $300 million, as required

under 10 CFR 140.13(b), at least 30 days prior to the planned date for obtaining licensed

material.

1.2.4 Type, Quantity, and Form of Licensed Material

Table 1.2-1 of the SAR (AES, 201 Oa) lists the types, quantity, and form of licensed material

proposed for acquisition, delivery, receipt, possession, production, use, transfer, and/or storage.

Table 1.2-1 of this SER lists AES's proposed possession limits for SNM, source material, and

by-product material.

The applicant has included technetium-99 (Tc-99), transuranic isotopes, and other contaminants

in Table 1.2-1 of the SAR (AES, 201 Oa). These radionuclides may exist at the EREF as a

consequence of the historical feed of uranium at other enrichment facilities, for example,

process contaminants and waste or material held in UF 6 cylinders from previous operations. As

indicated in Footnote 1 to Table 1.2-1 of the SAR (AES, 2010a), the applicant will require UF 6

suppliers to provide commercial natural UF6 in accordance with the requirements of American

Society of Testing and Materials (ASTM) C787-03, "Standard Specification for Uranium

Hexafluoride for Enrichment" (ASTM, 2003). This standard contains the purity requirements of

the uranium enrichment feed. In addition, cylinder suppliers will be required to preclude use of

cylinders that, in the past, have contained reprocessed UF6, unless they have been

decontaminated. Periodic audits of suppliers will be performed to provide assurance that these

requirements are satisfied. Because the applicant has indicated that natural uranium supplied

to EREF will meet ASTM Standard C787-03, the staff finds that this is acceptable for ensuring

that the Tc-99 and transuranic possession limits contained in Table 1.2-1 of the SAR are not

exceeded. Thus, the quantities of Tc-99 and transuranic isotopes from residual contamination

are expected to have no significant radiological impact.

In addition, Section 1.2.3 of the SAR (AES, 2010a) states that other source and byproduct

materials will be used for instrument calibration purposes. The applicant intends to identify

these materials during the design phase for the EREF and will submit future license amendment

requests to incorporate the proposed quantities and types for the sealed and unsealed

instrument calibration sources to its possession limits (AES, 2010a).

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Table 1.2-1 Proposed Possession Limits

Source or Special Nuclear Physical and Chemical Form Maximum Amount to

Material be Possessed at Any

One Time

Uranium (natural and Physical Form: Solid, liquid and gas 225,000,000 kg

depleted) and daughter Chemical Form: UF 6, UF 4, U0 2 F2, (496,000,000 Ib)

products oxides and other compounds

Uranium enriched in isotope Physical Form: Solid, liquid and gas 1,750,000 kg

U-235 up to 5 percent by Chemical Form: UF6, UF 4, U0 2 F2, (3,860,000 lb)

weight and uranium oxides and other compounds

daughter products

Tc-99, transuranic isotopes Any Amount that exists as

and other contamination contamination as a

consequence of the

historical feed of

recycled uranium at

other facilities

Note: Tc-99 - technetium-99

UF6 - uranium hexafluoride

UF 4- uranium tetrafluoride

UO 2F2 - uranyl fluoride

As stated above, the applicant identified the elemental name, maximum quantity, and

specifications, including the chemical and physical forms, of the licensed material that the

applicant proposes to acquire, deliver, receive, possess, produce, use, transfer, or store. The

applicant also identified the isotopic content and amount of enrichment by weight percent of the

licensed material. The information provided by the applicant meets the guidance in

Section 1.2.4.3 (3) of NUREG-1520 (NRC, 2002) and is, therefore, acceptable.

1.2.4.1 Authorized Uses

The application is for the issuance of a license for the construction and operation of a gas

centrifuge uranium enrichment facility in accordance with 10 CFR Parts 30, 40, and 70. The

facility will employ centrifuge technology from ETC.

Section 1.2.4 of the SAR (AES, 2010a) lists the authorized uses of the SNM, source, and

byproduct material. The AES proposes to use SNM 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 ultimately be used in the manufacture of fuel for the production of

electricity by commercial nuclear power plants. In Section 1.2.4 of the SAR, AES also proposes

a 30-year license term (AES, 201 Oa).

Since the proposed facility will contain classified information, AES also requested approval of a

classified matter facility clearance for the EREF under 10 CFR Part 95 and has submitted a

"Security Plan for the Protection of Classified Matter" as part of the license application.

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As stated above, the applicant provided a summary, non-technical narrative description for each

activity or process in which the applicant proposed to acquire, deliver, receive, possess,

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

material proposed for the facility are described in Section 1.2.4 of the SAR and are consistent

with the Atomic Energy Act of 1954, as amended. The description is also consistent with more

detailed process descriptions submitted as part of the ISA Summary (AES, 2010b), as reviewed

under Chapter 3 of this SER. The information provided by the applicant meets the guidance in

Section 1.2.4.3(4) of NUREG-1 520 (NRC, 2002) and is, therefore, acceptable.

1.2.4.2 Special Exemptions or Special Authorizations

In Section 1.2.5 of the SAR (AES, 2010a), the applicant requested an exemption from the

regulations. In addition, in subsequent correspondence, the applicant requested a special

authorization (AES, 2010g). The following sections describe the staff's evaluation of these

requests.

1.2.4.2.1 Exemption to Provide Forward-Looking Incremental Funding for Decommissioning

In Section 1.2.5 of the SAR (AES, 2010a), the applicant requested an exemption from

10 CFR 40.36(d) and 10 CFR 70.25(e), "Financial Assurance and Recordkeeping for

Decommissioning," in order to provide forward-looking incremental funding for

decommissioning.

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

10 CFR 70.25(e) and include a requirement that a licensee certify that financial assurance has

been provided in the amount of the cost estimate for decommissioning. AES's cost estimate for

decommissioning is based on a 30-year operating life and includes the estimated

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

disposition of depleted uranium tails. AES intends to sequentially install and operate modules of

the enrichment equipment, thereby phasing in the enrichment capacity for the EREF and the

resultant accumulation of depleted uranium tails. AES has requested the exemption in order to

provide financial assurance for decommissioning during the operating life of the EREF at a rate

that is proportional to the decommissioning liability for these facilities as they are phased in.

Similarly, AES plans to provide financial assurance for the disposition of depleted uranium tails

at a rate proportional to the amount of accumulated depleted uranium tails onsite up to the

maximum amount of the depleted uranium tails produced by the EREF.

AES has requested the exemption on the basis that AES is committed to updating the

decommissioning cost estimates and to providing to the NRC revised funding instruments for

facility decommissioning prior to the operation of each facility module (AES, 201 Oa). AES also

commits to updating the decommissioning cost estimates on an annual forward-looking

incremental basis and to providing the NRC revised funding instruments that reflect these

projections of depleted uranium tails production (AES, 2010a). AES expects to be able to

accurately predict the production of depleted uranium tails based on long-term enrichment

contracts. If any adjustments to the funding assurance were determined to be needed during

the annual period due to production variations, AREVA would make them promptly and provide

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a revised funding instrument to the NRC. The requested exemption would allow AES to satisfy

the applicable decommissioning funding assurance requirements for the EREF without imposing

an unnecessary financial burden on itself.

Under the provisions of 10 CFR 40.14, and 70.17, both titled "Specific Exemptions," the

Commission may, upon application by any interested person or upon its own initiative, grant

exemptions from the requirements of 10 CFR Parts 40 or 70, respectively, when the exemptions

are authorized by law, will not endanger life or property or the common defense and security,

and are otherwise in the interest of the public.

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

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

applicant from providing decommissioning funding assurance on an incremental basis. Staff

also determined that, because the incremental funding approach proposed by the applicant will

provide funding for the all the 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.

The granting of this exemption meets the criteria for categorical exclusion under 10 CFR 51.22.

The categorical exclusion described in 10 CFR 51.22(c)(25) addresses the granting of

exemptions from regulatory requirements. Granting this exemption request meets the criteria

described in 10 CFR 51.22(c)(25), namely, that there are no significant hazard considerations;

no change in the types or amounts of effluents; no increase in public or occupational radiation

exposure; no construction impact; no increase in the potential for or consequences of a

radiological accident; and the exemption involves surety requirements.

Therefore, NRC staff grants the requested exemption. A license condition will be included in the

license that will address AES's commitments for updating the decommissioning funding plan

over time. This license condition is discussed further in Chapter 10 of this SER.

1.2.4.2.2 Authorization to Make Certain License Application Changes Without Prior NRC

Approval

The applicant has requested authorization to make changes to license commitments, without

prior NRC approval, that do not decrease the effectiveness of these commitments (AES,

2010g). The requested authorization has two parts covering (1) changes requiring prior

approval and (2) changes not requiring prior approval. Changes to the license application that

will require prior NRC approval are those that would decrease the effectiveness of any safety

commitments (AES, 2010g). For a change that would decrease the effectiveness of a safety

commitment, the applicant will submit the change to the NRC as a license amendment and will

not implement the change until the NRC has reviewed and approved the amendment (AES,

2010g). Changes to the license application that could be made without prior NRC approval are

those for which there is no degradation in the safety commitments in the license application and

for which the change, test, or activity does not conflict with any condition specifically stated in

the license (AES, 201 Og). Records of such changes shall be maintained, including technical

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justification and management approval, in dedicated records to enable NRC inspection upon

request at the facility (AES, 2010g). A report containing a description of each such change, and

an appropriate revised section of the license application, will be submitted to the NRC within

three months of implementing the change (AES, 2010g).

In evaluating the applicant's request for authorization, the staff considered the guidance in the

Draft Regulatory Guide (RG), DG-3037, "Guidance for Fuel Cycle Facility Change Processes"

(NRC, 2009). This draft RG states that a license condition can be applied to allow changes to

the safety analysis report without prior NRC approval (NRC, 2009). The draft RG also

recommends that a license condition allowing changes to the Safety Analysis Report without

prior NRC approval should contain the following: (1) criteria; (2) a commitment to document the

licensee's evaluation supporting any findings that preapproval is not required; and (3) -the

reporting frequency for providing changes to the NRC after implementation (NRC, 2009). The

staff reviewed the applicant's authorization request and found it to be consistent with the

guidance provided in the draft RG.

The staff will impose the following license condition for this authorization:

The licensee is hereby granted the special authorization as identified in Section 1.2.5

"Special Exemptions and Special Authorizations" of the Eagle Rock Enrichment Facility

Safety Analysis Report:

a. The licensee shall not make changes to the license application that decreases

the effectiveness of safety commitments in the license application, without prior

NRC approval. For these changes, the licensee shall submit to the NRC, for

review and approval, an application to amend the license. Such changes shall

not be implemented until approval is granted.

b. Upon documented completion of a change request for a facility or process, the

licensee may make changes in the facility or process as presented in the license

application, or conduct tests or activities not presented in the license application,

without prior NRC approval, subject to the following conditions:

1. There is no degradation in the safety commitments in the license

application and

2. The change, test, or activity does not conflict with any condition

specifically stated in the license application.

Records of such changes shall be maintained, including technical justification and

management approval, in dedicated records to enable NRC inspection upon request at

the facility. A report containing a description of each such change, and appropriate

revised sections to the license application, shall be submitted to the NRC within three

months of implementing the change.

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1.2.4.3 Protection of Classified Matter

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

conclude that there is an adequate CMP for the protection of classified matter at the proposed

EREF facility and that a facility clearance can be issued.

1.2.4.3.1 Regulatory Requirements

10 CFR Part 70.22(m) requires an application to contain a full description of an applicant's

security program to protect against the unauthorized viewing of classified enrichment equipment

and unauthorized disclosure of classified matter in accordance with the requirements of

10 CFR Part 95.

1.2.4.3.2 Regulatory Guidance and Acceptance Criteria

The EREF CMP was reviewed for compliance with the requirements of 10 CFR Part 95 by using

"Standard Practice Procedures Plan Standard Format and Content for the Protection of

Classified matter for NRC Licensee, Certificate Holder, or Other Activities as the Commission

May Determine" (NRC, 2006).

1.2.4.3.3 Staff Review and Analysis

AES submitted its CMP entitled "Eagle Rock Enrichment Facility Standard Practice Procedure

Plan" on December 30, 2008, and submitted Revision 1 on April 23, 2009 and Revision 2 on

April 30, 2010. The applicant's CMP (AES, 2010d) outlines 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 Part 95.

Access to the site will be through the Entry Exit Control Point on the south end of the site.

1.2.4.3.4 Evaluation Findings

The NRC has reviewed the CMP for the EREF and considers it will satisfy the requirements of

10 CFR Part 95 when fully implemented. The CMP review generated a request for additional

information (RAI), to which the licensee responded in September 2009 (AES, 2009a).

The applicant stated that access to classified information will be controlled in accordance with

10 CFR Part 25 and 10 CFR Part 95 and has provided an acceptable CMP 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. The NRC will perform a readiness review of the site

against the EREF CMP when facilities are in place prior to classified material being allowed

onsite. By meeting these requirements, the applicant complies with the requirements of

10 CFR 70.22(m) to describe the security program to protect against unauthorized viewing of

classified enrichment equipment, and unauthorized disclosure of classified matter.

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However, NRC's authorization for the applicant to begin implementation of the CMP is

contingent upon an NRC inspection and finding prior to receipt of classified matter that AES's

classified matter program at EREF is being implemented in accordance with the CMP. The

NRC staff will impose the following license condition to ensure that classified matter is not

processed, stored, reproduced, transmitted, handled or accessed, except as permitted under

10 CFR Part 95:

The licensee shall not use, process, store, reproduce, transmit, handle, or allow

access to classified matter except as provided by applicable personnel and facility

clearances required under 10 CFR Part 95.

In Section 2.0 of the CMP, AES also commits to following Nuclear Energy Institute guidelines,

NEI 08-11, "Information Security Program Guidelines for Protection of Classified Material at

Uranium Enrichment Facilities" (NEI, 2009). The programs addressed in NEI-08-11 represent

new program elements 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).

Because AES has not yet designated the areas where the use and handling of classified

information will occur, NRC staff will impose the following license condition to ensure that areas

used for handling classified information are properly protected:

Prior to designating areas where the use and handling of classified information will

routinely occur, NRC will be notified to determine if additional security measures are

required. If NRC does determine the need for additional security measures, an

amendment request must be submitted, and approved, prior to establishment and use of

the area(s).

1.2.5 Evaluation Findings

The staff reviewed the institutional information for the proposed EREF, according to Section 1.2

of the Standard Review Plan (NRC, 2002). 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 will impose the

license condition provided in Section 1.2.3.3.3 of this SER:

The licensee shall provide proof of full liability insurance of $300 million as required

under 10 CFR 140.13(b), at least 30 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

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

information on one exemption request related to decommissioning funding, which meets the

requirements of 10 CFR 40.14 and 10 CFR 70.17.

The applicant has also adequately described information related to FOCI, 10 CFR 40.38 and

10 CFR 70.40, and its plans to secure classified matter for a facility clearance under 10 CFR

Part 95. The staff reviewed the applicant's Standard Practice Procedure Plan and found it to

satisfy the requirements of 10 CFR Part 95. Because a specific facility for use and storage of

classified matter has not been identified, staff will impose the following license condition,

provided in Section 1.2.4.3.4 of this SER:

The licensee shall not use, process, store, reproduce, transmit, handle, or allow

access to classified matter except as provided by applicable personnel and facility

clearances required under 10 CFR Part 95.

Because the areas where the use and handling of classified information will occur have not

been identified, staff will impose a license condition provided in Section 1.2.4.3.4 of this SER:

Prior to designating areas where the use and handling of classified information will

routinely occur, NRC will be notified to determine if additional security measures are

required. If NRC does determine the need for additional security measures, an

amendment request must be submitted, and approved, prior to establishment and use of

the area(s).

The staff also reviewed and approved the applicant's special authorization request to make

certain license application changes without prior NRC approval, as described in

Section 1.2.4.2.2 of this SER, and will impose the license condition stated in Section 1.2.4.2.2.

The staff concludes that the applicant has met the requirements and acceptance criteria in

Section 1.2.4.3 of NUREG-1520 (NRC, 2002).

1.3 Site Description

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

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

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

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

Environmental Report (ER), the Emergency Plan (EP), and the ISA Summary.

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). In addition,

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10 CFR 70.61(f) requires each licensee to establish a controlled area, as defined in

10 CFR 20.1003,2 and to retain authority to exclude or remove personnel and property from the

area.

1.3.2 Regulatory Guidance and Acceptance Criteria

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

SAR (AES, 2010a) are contained in 10 CFR 30.33, 10 CFR 40.32, 10 CFR 70.22,

10 CFR 70.65(b)(1), and Section 1.3.4.3 of NUREG-1 520 (NRC, 2002). Chapter 1 of

NUREG-1 520 is applicable to the EREF in its entirety.

1.3.3 Staff Review and Analysis

1.3.3.1 Site Geography

Section 1.3 of the SAR (AES, 2010a) describes the EREF's location and nearby roadways and

bodies of water, and significant geographical features.

The proposed site is located in southeastern Idaho, in Bonneville County, near the border with

Bingham County, and about 113 km (70 mi) west of the Idaho/Wyoming State line. The city of

Idaho Falls is located about 32 km (20 mi) east southeast from the site. The towns of Rigby and

Rexburg are located approximately 23 km (14 mi) and 42 km (26 mi) north of Idaho Falls,

respectively. Atomic City is about 32 km (20 mi) west of the site. The towns of Blackfoot and

Pocatello are located approximately 40 km (25 mi) and 76 km (47) miles south of the site,

respectively. The Fort Hall Indian Reservation comprises about 220,150 ha (544,000 ac) and

also lies to the south of the site. The nearest boundary of the reservation to the site is about

44 km (27 mi). The town of Fort Hall is about 60 km (37 mi) from the site.

The site is described as native rangeland, non-irrigated seeded pasture, and irrigated cropland.

The lands north, east and south of the site are a mixture of private-, State-, and Federal-owned

lands (AES, 201 Oa).

The proposed site is located in the east central part of the East Snake River Plain. The

proposed site is relatively flat with a gentle sloping surface with small ridges and areas of rock

outcrop. Most of the site is semi-arid steppe. Elevations at the site range from 1,556 m

(5,106 feet [ft]) to 1,600 m (5,250 ft).

The proposed EREF site contains no surface water bodies (AES, 2010b). The closest surface

water bodies are the Snake River and the Market Lake Wildlife Management Area. These two

surface water bodies are located about 32 km (20 mi) to the east and northeast of the site,

respectively (AES, 2010b).

The EREF will be located on the north side of U.S. Highway 20, a two-lane highway. A dirt road

currently provides site access from U.S. Highway 20, while other dirt roads provide access

2 A controlled area is defined in 10 CFR 20.1003 as an area, outside of a restricted area, but inside the

site boundary, access to which can be limited by the licensee for any reason.

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throughout the proposed site. To the east, U.S. Highway 20 intersects with Interstate 15 on the

west side of Idaho Falls, Idaho. To the west, U.S. Highway 20 intersects with U.S. Highway 26

northwest of Atomic City and ultimately intersects with Interstate 84 outside the town of

Mountain Home, Idaho, southeast of Boise.

There are no gas pipelines (industrial gases, natural gas, etc.) located on or nearby the

proposed facility site (AES, 2010b).

AES's SAR describes the main land uses nearby the EREF. Grazing and cropping are the main

land uses within 8 km (5mi) of the site (AES, 201 Oa). State land immediately west of the

proposed site and U.S. Bureau of Land Management land immediately east of the site are

grazed. Section 1.3.2.5 of the SAR provides additional information about nearby land uses

(AES, 2010a).

The SAR also includes information on nearby industrial facilities (AES, 2010a). The nearest is

the DOE's Idaho National Laboratory (INL). Its eastern boundary is 1.6 km (1 mi) west of the

proposed site. The INL property near the site is undeveloped rangeland. The closest facility on

the INL property is the Materials and Fuels Complex, located about 16 km (10 mi) west of the

EREF. In addition, there are landfills in Jefferson, Bonneville, and Bingham counties and two

waste transfer stations in Bonneville County. The nearest commercial carrier airport is Fanning

Field (Idaho Falls Regional Airport) in Idaho Falls about 32 km (20 mi) from the site. Pocatello

Regional Airport is located in Pocatello, about 113 km (70 mi) south of the site.

Information on public facilities, such as schools and hospitals, is also provided in the SAR

(AES, 201 Oa). Most of the public facilities are located in Idaho Falls, including hospitals, nursing

homes, schools and churches.

There are four fire departments within about a 48-km (30-mi) radius of the site; the Idaho Falls

Fire Department, the Ucon Volunteer Fire Department, the Shelley Firth Rural Fire Department,

and the Central Fire District which operates in Jefferson County. Fire support service for Idaho

Falls is provided by the Idaho Falls Fire Department, located approximately 32 km (20 mi) from

the EREF.

Information on rail lines is provided in Section 1.3.1.2 of the SAR (AES, 2010a). The nearest

rail lines include:

" The Union Pacific Railroad Aberdeen Branch-about 40 km (25 mi) south of the EREF

* The Union Pacific Railroad Scoville Branch-leading onto the INL

* The Eastern Idaho Rail Road operates short line tracks connecting towns north and east of

Idaho Falls to the Union Pacific Line

The site property boundary and the controlled area boundary for the EREF are shown on Figure

1.1-3, "Site Plan with Property and Control Area Boundary" of the SAR (AES, 2010a).3

3 Drawings of the site plan and facility layout have been withheld from public release as security-related

information under 10 CFR 2.390.

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1.3.3.1.1 Site Geography Conclusions

AES provided a summary describing the site geography, including its location relative to

prominent natural and manmade features (such as rivers, airports, population centers, schools,

and commercial and manufacturing facilities). The summary also described the site boundary

and controlled area boundary. The NRC staff reviewed information provided in the SAR

(AES, 2010a) and ISA Summary (AES, 201 Oa) onsite geography and finds the data used to be

accurate. The applicant's descriptions are consistent with the more detailed information in the

ISA Summary (AES, 2010b), the ER (AES, 2010c), and the EP (AES, 2010f). The information

the applicant provided is consistent with the guidance in Sections 1.3.4.3(1) and 1.3.4.3(5) of

NUREG-1520 (NRC, 2002) and is, therefore, acceptable.

1.3.3.2 Demographics

The proposed site is located in Bonneville County, Idaho. Portions of Bonneville, Jefferson, and

Bingham counties are within 8 km (5 mi) of the site. The combined population of Bonneville,

Bingham, and Jefferson counties in the EREF vicinity is 143,412, based on the 2000 U.S.

Census (AES, 2010a). This population represents an average annual increase of 1.4% from the

1990 population, less than that for the state of Idaho during the same period. It is expected that

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

rate for the state of Idaho (AES, 2010a).

Major population centers near the proposed site include:

* Idaho Falls, Idaho, about 32 km (20 mi) east southeast of the site

* Shelley, Idaho, about 45 km (28 mi) southeast of the site

* Blackfoot, Idaho, about 77 km (48 mi) southeast of the site

• Pocatello, Idaho, about 113 km (70 mi) south of the site

* Rexburg, Idaho, about 82 km (51 mi) northeast of the site

* St. Anthony, about 101 km (63 mi) northeast of the site

Aside from these communities, the population density is generally low. There are no

residences, schools, stores or other population centers within a 1.6 km (1 mi) radius of the site.

The nearest residence is 7.7 km (4.8 mi) east of the proposed site boundary (AES, 2010b).

The three hospitals in Bonneville County are located in Idaho Falls. The Eastern Idaho

Regional Medical Center is the largest of these. It is a short-term, acute care hospital with

242 beds. The other two hospitals are the Idaho Falls Recovery Center, a 7-bed, acute care

facility; and the Mountain View Hospital, a 20-bed, acute care facility. The closest nursing

homes or retirement facilities, schools, and churches are located in Idaho Falls.

Public use areas include a hiking trail south of the proposed site in Hell's Half Acre Wilderness

Study Area and a small lava tube cave located approximately 8 km (5 mi) east and south.

The applicant provided a summary of demographic information based on the most recent

census data that showed the population distribution as a function of distance from the proposed

facility. The applicant's descriptions are consistent with the more detailed information in the ISA

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Summary (AES, 2010b), the ER (AES, 2010c), and the EP (AES, 2010f). The information the

applicant provided is consistent with the guidance in Sections 1.3.4.3(2) and 1.3.4.3(5) of

NUREG-1520 (NRC, 2002) and is, therefore, acceptable.

1.3.3.3 Meteorology

Section 1.3.3 of the SAR (AES, 2010a) and Section 3.2.3 of the ISA Summary (AES, 2010b)

provide a meteorological description of the site and its surrounding area.

1.3.3.3.1 Tornado Hazard

Forty tornadoes were recorded from January 1, 1950, through April 30, 2008 (58 years of data),

in the four-county (Bonneville, Bingham, Butte, and Jefferson) region (AES, 2009a,b). Among

them, 1 is an F2 (Fujita Scale) tornado, 19 are F1 tornadoes, and the remaining are FE

tornadoes. Based on these data, the applicant calculated the annual probability for a tornado to

strike in any 2,589-km2 (1,000-mi 2) area in the vicinity of the EREF site to be 0.09 years

(AES, 201 Ob). The applicant concluded that the probability of a tornado to strike the EREF is

small. The NRC staff verified that the applicant's assessment of the probability of a tornado

strike at the site as small (approximately 1.0 x 10-5 to 1.0 x 10- 4/year) and, therefore, a tornado

strike is considered unlikely based on the applicant's likelihood definition. Thus, the applicant

assessed potential effects of tornado hazard to the EREF. NUREG/CR-4461 (NRC, 2006b)

provides recommended tornado design wind speed maps for the continental US. The applicant

used Figure 8-1 from NUREG/CR-4461 (NRC, 2006b) and determined that the recommended

tornado design wind speed, associated with a 1.0 x 10- 5

/year tornado at the EREF, to be 0

km/hour (0 mi/hour). The NRC staff verified that this value, presented in NUREG/CR-4461, is

obtained by smoothing out the spatial variation in tornado wind speeds for a given probability

level. Because tornadoes cover a finite discrete area for each event, a wind speed of zero

indicates that the probability of any part of a tornado wind field impacting the site is less than

1.0 X 10-5 yr 1. Staff note that the non-smoothed value of maximum tornado wind speed for a

probability level of 1.0 X 10-- yr1 is 82 mph, however, this is less than the design extreme wind

of 105 mph and does not need to be considered as a separate design bases. Consequently, no

special consideration of tornado hazard is necessary because of the estimated zero tornado

design wind speed. The NRC staff concludes that the applicant used appropriate

meteorological data to assess the probability of a tornado strike and has appropriately

concluded that no special consideration of tornado hazard is necessary in the design basis for

the EREF, thereby meeting the regulatory requirements in 10 CFR 70.65(b)(1) to provide a

general description of the site with emphasis on those factors that could affect safety.

1.3.3.3.2 High Winds

The applicant characterized the high-wind hazard at the proposed EREF site using the highwind hazard probability relationship reported in "Natural Phenomena Hazards Modeling Project:

Extreme Wind/Tornado Hazard Models for Department of Energy Sites," (Coats, 1985) for the

INL, whose eastern boundary is approximately 1.6 km (1 mi) west of the proposed site

(AES, 2010b). The NRC staff found that the applicant's assessment that the high-wind hazard

probability relationship for the INL is equally applicable to the proposed site, given the close

proximity of the proposed site to INL, is reasonable. The NRC staff also found that the high1-22
wind hazard recommended in "Natural Phenomena Hazards Modeling Project: Extreme

Wind/Tornado Hazard Models for Department of Energy Sites," (Coats, 1985) is acceptable

because this high-wind hazard was developed using an acceptable methodology and was

reviewed and commented on by experts from organizations, including the National Oceanic and

Atmospheric Administration, the NRC, and the U.S. Geologic Survey (USGS). According to

"Natural Phenomena Hazards Modeling Project: Extreme Wind/Tornado Hazard Models for

Department of Energy Sites" (Coats, 1985), the 10-m (33-ft) aboveground fastest mile wind

speed with an annual probability of 1.0 x 10-5 is 169 km/hr (105 mi/hr).

Because the proposed EREF is not located near a coastal area (approximately 925 km [575 mi]

from the coast), hurricanes affecting the coastal area will have no effect on the performance of

the facility. Consequently, consideration of hurricane wind hazards on the design of the

proposed EREF is not needed; and the staff conclude that the regulatory requirements in

10 CFR 70.65(b)(1 ) have been met with regard to this topic.

1.3.3.3.3 Extreme Precipitation

The applicant discusses extreme precipitation at the proposed EREF site in ISA Summary

Section 3.2.3.4.4 (AES, 2010b). The applicant estimated the extreme precipitations for 1-, 24-,

and 48-hour duration rainfalls corresponding to annual probability of 1.0 x 10-5 using the rainfall

data extracted from the "Precipitation Frequency Atlas of the Western United States"

(NOAA, 1973); the "Two-To-Ten-Day Precipitation for Return Periods of 2 to 100 Years in the

Contiguous United States (Commerce, 1964); and the National Weather Service's least square

regression procedure. The estimated 1-, 24-, and 48-hour, all-season, extreme local

precipitations for the annual probability of 1.0 x 10-5 are 60 mm (2.37 in), 112 mm (4.39 in), and

135 mm (5.31 in), respectively (AES, 2010b).

The NRC staff reviewed the analysis supporting the applicant's estimates of extreme

precipitations during an onsite review and finds that this analysis is acceptable because it used

rainfall data from reliable sources; and the least square regression procedure the applicant used

to extrapolate the extreme precipitations is an industry-accepted, appropriate method for rainfall

hazard assessment. The NRC staff concludes that the applicant used appropriate

meteorological data and an industry-accepted methodology to estimate the extreme

precipitations, thereby meeting the regulatory requirements in 10 CFR 70.65(b)(1).

1.3.3.3.4 Flood

ISA Summary Section 3.2.4.3 discusses potential for flooding at the EREF (AES, 2010b). The

applicant indicated that the proposed EREF is not located near any large body of water that

could cause a flood at the facility site. The nearest large surface waters are the Snake River,

which is approximately 32 km (20 mi) east, and Lake Wolcott, which is approximately 120 km

(75 mi) southwest of the site. Therefore, the applicant concluded that no credible flooding can

occur at the site resulting from either existing river sources or failure of upstream dams. The

NRC staff finds the applicant's conclusion, that flooding caused by overflow of surface rivers

and failure of upstream dams is not credible, is acceptable. The staff agrees that this type of

flooding hazard is highly unlikely.

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The applicant indicated that local precipitation-induced roof-ponding would be limited through

roof design. Consequently, roof-ponding hazard resulting from local intense precipitation is

highly unlikely (AES, 2010c). The NRC staff finds the applicant's conclusion acceptable

because, with appropriate roof design, the applicant will be able to limit accumulation of extreme

local precipitation on the roof to ensure that the resulting roof-ponding load does not exceed the

roof design load.

Regarding the floodwater intrusion potential because of extreme local precipitation, the

applicant indicated that the site of the proposed facility is located at a localized topographical

high ground (AES, 2010c). In addition, the ground floor levels of the proposed facility safetyrelated structures will be 0.15 m (6 in) above the adjacent finished grade. Based on these two

reasons, the applicant concluded that intrusion of floodwater into safety-related structures is

highly unlikely because the ground floor level is higher than the flood depths corresponding to

the 1- and 24-hour storm deluges corresponding to an annual probability of 1.0 x 10-5. The

NRC staff reviewed the information and concludes it is acceptable.

Additionally, the applicant's intent to make the finished grade slope away from buildings will

further prevent precipitation accumulations against the structures (AES, 201 Oc) and, therefore,

reduce floodwater intrusion potential.

In summary, the NRC staff concludes that flood hazards resulting from surface river overflows,

upstream dam failures, and extreme local precipitation are highly unlikely because no large

bodies of water are located near the proposed site. The NRC staff also concludes that because

the ground floor levels of the safety-related structures are above the potential flood levels for

the storm deluges corresponding to an annual probability of 1.0 x 10-5, flood hazard resulting

from extreme local precipitation is not a safety concern to the proposed facility. Consequently,

the NRC staff concludes that the regulatory requirements in 10 CFR 70.65(b)(1) have been met

with regard to this topic.

1.3.3.3.5 Snow

In Section 1.3.3.2 of the SAR, the applicant accounts for the extreme environmental snow

hazard through design considerations so that the potential events associated with this hazard

are highly unlikely (annual probability smaller than or equal to 1.0 x 10-5) (AES, 2010a).

The design basis extreme environmental snow load for the proposed EREF was the sum of the

50-year snowpack load and the load corresponding to the 48-hour winter season extreme

precipitation with an annual probability of 1.0 x 10-5 for the area (AES, 2009a, b).

The applicant used the data that the National Resources Conservation Service collected from

two locations near the proposed site to estimate the 50-year snowpack load (216.0 kg/M2

[44.2 lb/ft2]) (AES, 2009a). The applicant adopted the generalized extreme value method to

determine frequency distributions and calculate the snowfall depth corresponding to the 50-year

return period.

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The applicant estimated the 48-hour winter season extreme precipitation with an annual

probability of 1.0 x 10-5 for the EREF site to be 94.0 mm (3.70 in), corresponding to a load of

92.8 kg/M 2 (19.0 lb/ft2). As a result, the design basis extreme environmental snow load is

309.0 kg/M2 (63.2 lb/ft2).

The NRC staff reviewed the analysis the applicant performed for determining design basis

extreme environmental snow load and finds that the analysis is acceptable because: (i) the

snowfall data used are from a reliable source and (ii) the methods used to estimate the 50-year

return snowpack load and the 48-hour winter season extreme precipitation with annual

probability of 1.0 x 10-5 are the industry-accepted methods. Thus, the NRC staff concludes that

the regulatory requirements in 10 CFR 70.65(b)(1) have been met with regard to this topic.

1.3.3.3.6 Lightning

According to the applicant's ISA Summary (AES, 201 Ob), nine lightning strikes from January 1,

1950, to May 31, 2008, were recorded in the National Climate Data Center Storm Event

Database for the four-county region where the EREF site will be located. The applicant

estimated the lightning strike frequency for the proposed EREF site to be 0.75 flashes per year

using the methodology proposed by considering the attractive area of structures (Marshall,

1973). Based on the lightning strike frequency estimate, lightning strike hazard is likely at the

proposed site based on the applicant's likelihood definition. The applicant accounts for the

lightning hazard by providing lightning protection to the EREF as described in Section 7.3.7 of

the SAR (AES, 2010a). In Section 7.3.7, the applicant states that lightning protection will be

provided in accordance with National Fire Protection Association (NFPA) 780 "Standard for the

Installation of Lightning Protection Systems." In Section 7.3 of this SER, the staff finds the use

of NFPA standards to be in accordance with the guidance of Section 7.4.3 of NUREG-1520

(NRC, 2002) in regard to nationally recognized codes and standards that may be used. The

NRC staff finds that the applicant adequately considered lightning strike hazard through

lightning protection design. Consequently, the NRC staff concludes that the regulatory

requirements in 10 CFR 70.65(b)(1) have been met with regard to this topic. Meteorology

Conclusions

As discussed in Sections 1.3.3.3.1 through 1.3.3.3.6 of this SER, the applicant provided

appropriate meteorological data, including a summary of design-basis values for accident

analysis of maximum snow loads and probable maximum precipitation, as presented in the ISA

Summary (AES, 201 Ob). The applicant also provided appropriate design-basis information for

lightning, high winds, tornadoes, hurricanes, extreme precipitation, and temperature extremes.

The applicant's descriptions are consistent with the more detailed information in the ISA

Summary (AES, 201 Ob), the ER (AES, 201.0c), and the EP (AES, 2010f). The information the

applicant provided is consistent with the guidance in Sections 1.3.4.3(3) and 1.3.4.3(5) of

NUREG-1520 (NRC, 2002) and is, therefore, acceptable.

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1.3.3.4 Geology

1.3.3.4.1 Seismic Hazards

The applicant provided information on seismic hazards in Section 3.3.7 of the ER (AES, 2010c),

Section 1.3.5 of the SAR (AES, 2010a), and Section 3.2.6 of the ISA Summary (AES, 201 Ob).

In addition, the applicant provided additional details about the results of its seismic hazard

assessment in response to staff's request for additional information (AES, 2009a).

Specific areas of the seismic hazard analysis, that are applicable to the seismic design of the

proposed EREF, were reviewed:

Seismic source characterization

* ground motion attenuation models

* probabilistic seismic hazard assessment (PSHA)

* development of site-specific ground motion spectra

1.3.3.4.1.1 Seismic Source Characterization

Geological and Tectonic Settings

As the applicant discussed in the SAR and supporting documents (AES, 2009a, b, h, and g), the

proposed EREF site is located near the center of the Eastern Snake River Plain in Southeastern

Idaho. The Snake River Plain is a topographically subdued physiographic province that has

been structurally and volcanically active over the last 17 million years. The geology of the

Snake River Plain and resulting volcanic landscape was produced by the westward and

southwestward drift of the North American tectonic plate over the Yellowstone mantle hotspot.

The hotspot is now located beneath Yellowstone National Park.

The Snake River physiographic province is bordered on the northwest, west, south, and

southeast by the Basin and Range; on the northeast by the Yellowstone Plateau; and on the

north by Idaho Batholith provinces. These four physiographic provinces also correspond to

unique seismotectonic settings, each with the potential to generate earthquakes that could

impact the facilities at the proposed site. Tectonic activity in the surrounding provinces has

been ongoing for several hundred million years, reflecting the active tectonic evolution of

western North America. Over the last 60 million years, extensional tectonics of western

North America developed the classic Basin and Range topography, manifest as north-south or

northwest-southeast elongate and internally drained basins situated between fault-bound

mountain ranges of exhumed basement rock. Most of the faulting takes place along normal

faults at the interface between the basins and the ranges. Normal faulting along basin-bounding

faults in the Basin and Range is the main initiator of large-magnitude earthquakes in the region.

The applicant used the USGS database to develop a list of Quaternary fault sources. The

USGS database includes all known faults that have produced a magnitude 6.0 or larger

earthquake in the last 1.6 million years (Petersen, 2008). These include the Lost River fault,

which produced the 1983 Borah Peak earthquake; the Lemhi fault, which bounds the

southwestern face of the Lemhi Range; and the Beaverhead fault, which bounds the

southwestern side of the Beaverhead Mountains. The full list of seismic sources the applicant

used is documented in ISA Summary Appendix F, Table 4 (AES, 2010b).

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Historic Seismicity

Figure 6 in the ISA Summary, Appendix F (AES, 2010b), shows earthquake epicenters of the

Snake River Plain and surrounding areas based on earthquake records from the collection of

regional and national earthquake catalogs (USGS cited Peterson, 2008). As shown in the

figure, few earthquakes occurred within the Snake River Plain in contrast to the relatively large

number of earthquakes in the surrounding regions. Many geologists have attributed the lack of

seismicity in the Snake River Plain, compared to the surrounding provinces to continued high

heat flux of the Snake River Plain crust, such that a large fraction of the regional tensional

tectonic stress is accommodated by intrusion of magmatic dikes and crustal ductility (e.g.,

Parsons, 1998; Parsons, 1991). In contrast, tectonic stresses in the crust of the surrounding

provinces accumulate until the rocks are overstressed and ruptured, thus producing an

earthquake. The greater ductility of the Snake River Plain crust, combined with dike intrusions,

moderates the tectonic stress before it can reach levels sufficient to produce earthquakes.

For the PSHA, the applicant developed a catalog of historic earthquakes from regional and

national databases, including the Advanced National Seismic System catalog and the USGS

regional catalog. In the applicant's composite catalog, all earthquake magnitudes were

converted to moment magnitudes based on the equation of Johnston (Johnston, 1996). The

applicant's composite catalog identifies 66 earthquakes within approximately 322 km (200 mi) of

the site with a moment magnitude of 5.0 or larger. The applicant identified the 1959 Hebgen

Lake earthquake (moment magnitude [Mw[ = 7.3, surface wave magnitude [Ms[ = 7.5) and the

1983 Borah Peak earthquake (Mw = 6.8, Ms = 7.3) as the most significant historic earthquakes

in the region. The 1959 Hebgen Lake earthquake was the largest historical earthquake in the

intermountain region. The 1983 Borah Peak earthquake was the largest earthquake ever

recorded in Idaho. Also included in the applicant's list is the 1905 Shoshone earthquake, which

is important because it is one of the few earthquakes that may have occurred within the Snake

River Plain. Although, because it took place before seismic instrumentation, the location of its

epicenter is highly uncertain.

Seismic Sources

The applicant developed a list of fault sources as inputs to the PSHA based on the location of

known Quaternary faults in the region, as defined in the USGS database. The applicant derived

regional seismic source zones from interpretations of the historic seismicity. Seismicity

parameters, including activity rates, recurrence, and maximum magnitude for each of the fault

sources were adopted from the USGS database. Seismicity parameters for the regional source

zones were derived from the earthquake catalog information. To account for uncertainty in the

regional seismic source zones, especially the distributions of spatial densities of past

earthquakes, the applicant developed four alternative source models as inputs to the PSHA.

These alternative source models, which include both regional seismic source zones and the

fault sources, were used to develop four initial branches of the PSHA logic tree.

Staff Evaluation of Seismic Source Characterization

The NRC staff reviewed the information in ISA Summary, Appendix F, and finds that the

applicant properly identified and characterized the seismic sources that could impact the site.

The applicant's seismic source characterization is also consistent with other recent NRC

licensing actions, including the Idaho.Spent Fuel Facility at the nearby INL reservation

1-27
(NRC, 2004). Based on an evaluation of the applicant's database of historical seismicity, the

NRC staff also finds that the applicant developed an adequate set of historical seismic data

used to develop seismic source parameters. Based on this review, the NRC staff concludes,

that the information concerning seismic source characterization presented in the SAR (AES,

2010a), ISA Summary (AES, 2010b), and Appendix F to the ISA Summary is acceptable. The

information adequately summarized seismicity and potential fault and tectonic sources to meet

the regulatory requirements in 10 CFR 70.65(b)(1).

1.3.3.4.1.2 Ground Motion Attenuation

The applicant used the SEA99 (Spudich, 1999) and the B&A08 (Boore, 2008) ground motion

attenuation models in its PSHA. The applicant selected these models because they included

specific predictions for normal faults. Additionally, the models were based on a large set of

strong motion recordings and were applicable to bedrock conditions with shear wave velocities

greater than 620 m/s (2,034 feet per second [ft/s]) in the upper 30 m (98 ft) of the ground

surface (VS30). The applicant stated that these high shear wave velocities are consistent with

conditions present at the site. The site has basalt as bedrock with a thin veneer of alluvial

sediments no more than 4.3 m (14 ft) thick. Geophysical measurements from Payne (Payne,

2002) made at the INL indicate that the basaltic bedrock has shear wave velocities in the range

of 1,200-1,500 m/s (4,000-5,000 ft/s). In the PSHA, the applicant developed seismic hazard

results for SEA99- assuming VS30 = 620 m/s (2,034 ft/s), B&A08 assuming VS30 = 760 m/s

(2,493 ft/s), and B&A08 assuming VS30 = 1,300 m/s (4,265 ft/s). The applicant will conduct

additional geophysical investigation to verify the VS30 values (AES, 2009a).

Staff Evaluation of Ground Motion Attenuation Models

The NRC staff reviewed the information in the ISA Summary, Appendix F (AES, 201 Ob), and

finds that the ground motion modeling approach the applicant used reasonably predicts the

earthquake-induced ground motions at the proposed EREF site. Both the SEA99 and B&A08

models are well-established attenuation models that were derived from a large dataset of

western United States strong motion records. Use of these models for bedrock conditions with

VS30 values greater than 620 m/s (2,034 ft/s) is also appropriate to the geology of the site

because of similarity of the bedrock conditions to those at the INL. Based on this review and

the applicant's commitment to verify the VS30 values, the NRC staff concludes that the

information concerning ground motion attenuation models presented in the SAR (AES, 201 Oa);

the ISA Summary (AES,2010b); and Appendix F to the ISA Summary is acceptable. The

information provides an adequate approach to model ground motion attenuation to meet the

regulatory requirements in 10 CFR 70.65(b)(1).

1.3.3.4.1.3 Probabilistic Seismic Hazard Assessment Calculation (PSHA) Results

The applicant developed a PSHA based on a logic tree formulation using the four seismic

source models and two ground motion attenuation models described in the previous sections of

this review. This results in eight logic tree nodes. The applicant gave each of these nodes

equal weight of 0.125. The B&A08 attenuation model nodes were then subdivided based on the

two possible VS30 values, and these subnodes were given weights of 0.0625. For the source

parameters, the applicant developed single activity rates for each source and single estimates

of maximum magnitude without additional variations in these parameters.

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Results of the PSHA were the probabilities of exceeding the peak ground acceleration and a set

of spectral accelerations at 5-percent damping for ground motion periods ranging from 0.01 to

10.0 seconds. Results of the hazard analysis indicate that the peak horizontal accelerations at

an annual exceedance frequency of 1 x 10-3 would be 0.063 g (61.8 cm/s 2), increasing to

0.30 g (294 cm/s 2) at an annual exceedance frequency of 1 x 10-5. Spectral acceleration at

5 Hz (with 5 percent damping) increases from 0.16 g (161.15 cm/s 2) at an annual exceedance

frequency of 1 x 10-3 to 0.76 g (743.50 cm/s 2) at an annual exceedance frequency of 1 x 10-5.

The applicant also showed that seismic activity close to the site and within the Eastern Snake

River Plain or from the source zone at Yellowstone contributes the most to the total hazard.

The applicant noted that the major Basin and Range faults are too far away from the site to

produce significant ground motion, given the high rate of ground motion attenuation predicted in

both the SEA99 and B&A08 attenuation models for normal fault earthquakes.

Comparison to Other PSHA Results

The results of the applicant's PSHA are up to 40 percent lower than hazard results indicated by

the 2008 USGS National Seismic Maps indicated (Peterson, 2008). In ISA Summary

Appendix F (AES, 2010b) and supplemented by the supporting calculations in its response to

NRC staff's RAI (AES, 2009a), the applicant cited three reasons why it concludes that USGS

results are too high.

The USGS hazard maps are based on an assumption of equal contributions to the

hazard from both strike-slip and normal faults, whereas the applicant's model only

considered normal faults. The attenuation models the USGS National Seismic Hazard

Maps used predict higher amplitudes of ground motions for a strike-slip earthquake

compared to a similar-sized earthquake on a normal fault.

The USGS models are based on an assumed site condition with VS30 = 760 m/s

(2,493 ft/s), whereas the applicant's PSHA includes 75 percent of the cases with a

VS30 = 760 m/s (2,493 ft/s) and the remaining 25 percent of the cases with

VS30 = 1,300 m/s (4,265 ft/s). Given the same earthquake, sites with higher

VS30 values (e.g., harder rock conditions) generally experience smaller ground motion

amplitudes than sites with smaller VS30 values.

The method the USGS used to develop seismicity parameters for the seismic Eastern

Snake River Plain seismic source region leads to a conservative estimation of seismic

activity in this zone. For example, the USGS approach estimates 11 to 14 earthquakes

with moment magnitudes greater than or equal to 4.0 should have occurred in the region

in the last 45 years. The applicant's site-specific model predicts only about six

earthquakes with moment magnitudes greater than or equal to 4.0 over the same period.

In comparison, the historic record shows that the region has only experienced two

earthquakes with moment magnitudes greater than or equal to 4.0 since 1963.

The results of the applicant's PSHA also differ from those developed for facilities at the INL

(e.g., the Idaho Nuclear Technology and Engineering Center (INTEC) PSHA given in Payne,

2002). Although the INL hazards also tend toward greater ground motion hazards than the

applicant's PSHA, the main difference is in the shape of the response spectra. The INTEC

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uniform hazard spectra are much flatter than those the applicant developed. The applicant

considered these differences from the relative proximity of the INL sites to the Basin and Range

normal faults.

Staff Evaluation of Probabilistic Seismic Hazard Assessment Calculation

The NRC staff reviewed the information presented in the ISA Summary Appendix F and the

response to NRC staff's request for additional information and find that applicant's PSHA

adequately predicts the earthquake-induced ground motions at the proposed EREF site. The

NRC staff finds that the applicant developed its PSHA based on available seismic and geologic

information, supported by reasonable assumptions about the future seismic activity in the

region. The rationale explaining differences from other existing PSHA results, especially those

from the USGS National Seismic Hazard Map, are well founded and clearly reasoned. The staff

agrees that normal faulting is the most likely style of faulting, given the present tectonic setting

of the region. The staff also agrees that the USGS models are applicable to softer site

conditions, and they overestimate seismicity in the source zones. Based on this review, the

NRC staff concludes, that the PSHA presented in the SAR, the ISA Summary, and the ISA

Summary Appendix F is acceptable. The information provided an adequate approach to

develop seismic inputs for design and performance consideration in the application and thereby

meets the regulatory requirements in 10 CFR 70.65(b)(1).

1.3.3.4.1.4 Development of Site-Specific Ground Motion Spectra

Based on the PSHA results, the applicant developed the horizontal motion Uniform Hazard

Response Spectra (UHRS) for selected annual exceedance frequencies between 2.1 x 10-3 and

1.0 X 10-5, which correspond to return periods between approximately 475 and 100,000 years.

Based on an in-office review of the applicant's calculations, for seismic design inputs, the

applicant developed UHRS for 4 x 10- 4, 1 x 10-4, and 1 x 10-5 for damping values of 5, 7, and

10 percent. These UHRSs were used to adjust the design response spectra to meet their target

performance objectives per the guidance in American Society of Civil Engineers (ASCE) 43-05

(ASCE, 2005).

The applicant also developed vertical response spectra as two-thirds that of the horizontal

spectra at all periods based on Bozorgnia and Campbell (Bozorgnia, 2004). Bozorgnia and

Campbell showed that a vertical-to-horizontal ratio of two-thirds is appropriate for site conditions

that include hard bedrock and a site-to-source distance of 60 km (37.3 mi) or more. The

applicant noted that the proposed EREF site meets these conditions because it is situated

above bedrock with high VS30 values and seismic sources that are outside the Eastern Snake

River Plain province and thus more than 60 km (37.3 mi) distant. The applicant will develop

acceleration time histories necessary for some engineering analyses to support design of the

facilities (AES, 2009a). The applicant will use stochastic methods to calculate acceleration time

histories with response spectra matching the design response spectrum at all frequencies. In

addition, the applicant will further verify that the resulting response spectra from these

acceleration time histories envelope the design basis ground response spectra developed using

the ASCE 43-05 methodology.

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Staff Evaluation of Site-Specific Ground Motion Spectra

The NRC staff reviewed the information in the ISA Summary, Appendix F (AES, 2010b), and the

response to NRC staff's request for additional information and find that applicant's site-specific

ground motion spectra adequately represent potential future earthquake activity at the proposed

EREF site. The NRC staff finds that the applicant developed its site-specific ground motion

spectra based on standard seismological practice with sufficient detail to support seismic

design. Vertical ground motions that are two-thirds the horizontal motions are appropriate and

consistent with information in the seismological literature. Based on this review and the

applicant's plan to develop acceleration time histories to support design analyses, the NRC

staff concludes that the site-specific ground motion spectra presented in the SAR; the ISA

Summary; and the ISA Summary, Appendix F is acceptable. The information provided an

adequate approach to develop seismic inputs for design and performance consideration in the

application and thereby meets the regulatory requirements in 10 CFR 70.65(b)(1).

1.3.3.4.2 Volcanic Hazard

The proposed EREF site is located within the eastern half of the Snake River Plain

Physiographic Province. The Snake River Plain is characterized by extensive volcanic activity;

and much of the region has been volcanically active since approximately 17 million years ago,

when this portion of the North American tectonic plate overrode the Yellowstone hotspot. Based

on historical patterns of volcanic activity in the region, three potential impacts of volcanic events

were considered:

* Ash fallout from Cascade Range volcanoes

* Near-field silicic volcanism

* Inundation from basaltic lava flows

Each of these three types of volcanic activity could have potentially adverse effects on

structures, systems, and components relied on for safety if such volcanic activity occurred

during operation of the proposed EREF.

Because of the potential for volcanic activity at the site, the applicant developed a volcanic

hazard assessment, which is described in Appendix D of the ISA Summary and is summarized

in the ISA Summary Section 3.2.8 (AES, 2010b). The NRC staff reviewed information the

applicant provided in the ISA Summary as well as responses to staff requests for additional

information (AES, 2009a) regarding volcanic features of the site. The NRC staff also reviewed

relevant literature cited in the ISA Summary, previous studies of volcanic hazards potentially

affecting the INL (Kuntz, 1992; Lawrence Livermore, 1990; Hackett, 1994), and other literature

cited therein to evaluate the potential volcanic hazards at the proposed EREF site.

Ash Fallout from Cascade Range Volcanoes

The applicant estimated that the maximum ash thickness that could be deposited on roofs of the

proposed EREF is less than 8 cm (3.1 in) according to the results from the Volcanism Working

Group Assessment for the nearby INL site (Lawrence Livermore, 1990). Based on dry and wet

ash densities from the 1980 Mount Saint Helens eruption (Sarna-Wojcicki, 1981), the applicant

estimated that an 8-cm (3.1 -in)-thick ash blanket would result in roof loads of 4-10 g/cm 2

1-31
(8.2-20.5 lb/ft2). The applicant noted that these potential ash loads are much smaller than the

design basis roof loads for the extreme environmental snow load, which is 30.9 g/cm 2

(63.2 Ib/ift2). Thus, the roof loads from volcanic ash are bounded by the design basis snow load.

The NRC staff reviewed the information provided in the ISA Summary (AES, 2010b) and

responses to staff requests for additional information (AES, 2009a) and find it acceptable

because the potential ash loads at the EREF site from a Cascade Range volcanic eruption rely

on logical analyses that are supported by well-established volcanic data. Moreover, the

analysis and results the applicant provided are consistent with a similar analysis the NRC staff

relied on in support of the SED for the Idaho Spent Fuel Facility at the nearby INL site

(NRC, 2004). In that SER, NRC staff found that a 10-cm (3.94-in) ash load was bounding. The

NRC staff therefore concludes that the regulatory requirements in 10 CFR 10.65(b)(1) have

been met with regard to this topic.

Near-Field Silicic Volcanism

Within the past 1.2 million years, five silicic dome volcanoes have formed in a region of the

Eastern Snake River Plain named the Axial Volcanic Zone (Lawrence Livermore, 1990). As

shown in Figure D-1 of Appendix D in the ISA Summary, the closest of these domes is

East Butte, which is located more than 10 km (6.1 mi) from the proposed EREF (AES, 2010b).

The applicant estimated a recurrence interval for silicic volcanism of 220,000 years (5 events in

1.2 million years) or a rate of less than 4 x 10-6/year, which is an order of magnitude smaller

than recurrence rates for basaltic volcanism.

Small-volume flows and tephra falls are common features of silicic dome eruptions

(Heiken, 1987). Active silicic dome volcanoes comparable in size to those in the Eastern Snake

River Plain produce only local flows and ash-fall deposits approximately 10 cm (4 in) thick near

the volcano (Scott, 1987), with deposit thicknesses decreasing to 1-2 cm (0.5-1 in) thick at

distances of about 25 km (15.5 mi) from the volcano (Heiken, 1987). Thus, a 10-cm (4-in)-thick

ash-fall deposit is determined to be a credible upper limit for potential hazards from a new silicic

dome volcano forming at least 10 km (6.2 mi) from the proposed EREF.

Because of these characteristics of near-field silicic volcanism-small recurrence rate and

limited effects at the site-the applicant considered these types of volcanoes to pose no

significant hazard to the proposed EREF. The applicant also noted that the Volcanism Working

Group (Lawrence Livermore, 1990) at INL reached a similar conclusion. This group examined

the potential hazards from silicic volcanism with respect to critical nuclear facilities within the

INL.

The NRC staff reviewed the information provided in the ISA Summary (AES, 2010b) and finds it

acceptable because the potential hazards at the EREF site from near-field silicic dome

volcanoes are derived from well-known and appropriate volcanic information. In addition, the

analysis and results the applicant provided are consistent with a similar analysis the NRC staff

relied on in support of the Safety Evaluation Report for the Idaho Spent Fuel Facility at the

nearby INL site (NRC, 2004). The NRC staff therefore concludes that the regulatory

requirements in 10 CFR 70.65(b)(1) have been met with regard to this topic.

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Inundation from Basaltic Lavas

The applicant identified that inundation by basaltic lava flows is the most significant of the three

potential volcanic hazard types at the site. Inundation could be from a new basaltic volcano at

the site or, more likely, from slow-moving, effusive lava moving downgradient from volcanic

vents or cones within the known Eastern Snake River Plain volcanic zones. Based on their

similar characteristics to basaltic lavas from Hawaii, the applicant describes these flows as hot,

low viscosity, and slowly moving pahoehoe lavas with advancement rates of less than 1 km

(0.6 mi) per day. The main hazards to the proposed EREF would be inundation and burning

with associated release of corrosive gases.

The applicant developed a probabilistic volcanic hazard analysis for the proposed site based on

extensive volcanic and geologic data from the Eastern Snake River Plain (Appendix D of

AES, 2009b). The probabilistic analysis considered two approaches to develop estimates for

the likelihood of basaltic lava inundation at the proposed site. The first approach used an event

tree to define possible outcomes based on the probability of a random eruption in the axial

volcanic zone. Subsequent branches of the event tree considered the likelihood that lava will

reach the proposed site based on topographic effects and run-out distance from the volcano to

the site. The second approach also assumes that a volcano forms randomly within the volcanic

source zone and then conditions the probability that a volcano will form by the ratio of the

average area of a lava flow divided by the area of the volcanic source zone. Both methods yield

similar estimates of lava inundation at the proposed site on the order of 5 x 10-6/year. Based on

this probability, the applicant defines volcanism at the site as highly unlikely.

The NRC staff reviewed the probabilistic volcanic hazard assessment provided in the ISA

Summary (AES, 201 Ob), including the supporting geologic and volcanic information, and finds it

acceptable. The applicant's probabilistic volcanic hazard assessment is clearly documented

and supported by well-established volcanic information. The results of the applicant's analysis

are consistent with independent estimates of basaltic lava inundation for critical facilities within

the INL reservation, including the analysis for the Central Facilities Area (Hackett, 2002) and the

study done for the New Production Reactor (Lawrence Livermore, 1990). The probability

estimate the applicant provided is also consistent with the results NRC staff relied on in support

of the Safety Evaluation Report for the Idaho Spent Fuel Facility at the nearby INL site

(NRC, 2004). The NRC staff therefore finds that the regulatory requirements in

10 CFR 70.65(b)(1 ) have been met with regard to this topic.

1.3.3.4.3 Slope Stability

The applicant indicated that the proposed facility site is relatively flat with gently sloping

surfaces and small ridges (AES, 2009a,b). The NRC staff examined the topography maps

where the proposed site is located [e.g., Figures 3.2-5 and 3.2-9 of the ISA Summary

(AES, 2010b) and concur with the applicant that the proposed site is gently sloping; and slope

instability is not a safety concern to the EREF operations. Consequently, the NRC staff

concludes that the regulatory requirements in 10 CFR 70.65(b)(1) have been met with regard to

this topic.

1-33
1.3.3.4.4 Liquefaction

The applicant indicated that the soil thickness at the proposed site varies from 0 to 4.3 m

(0 to 14.0 ft) (AES, 2010b). The site soils are of eolian (wind-blown) origin and are classified as

CL: low plasticity clays according to the Unified Soil Classification System. The soil natural

moisture content ranges from 9.6 to 19.0 percent, and the groundwater table is more than

150 m (500 ft) below ground surface. The applicant concluded that liquefaction potential for the

site appears highly unlikely because the groundwater table is deep, and the soils at the site are

primarily clays. The NRC staff reviewed the information the applicant provided and find that the

applicant's conclusion is technically supportable and therefore acceptable.

In addition, the applicant stated in its ISA Summary Section 3.2.7 (AES, 2010b) that it intends to

conduct additional site subsurface investigations to support the final design of the proposed

EREF and verify through the investigation results the conclusion on soil liquefaction potential.

The applicant stated that, if the investigation results show that soil liquefaction at the proposed

site is possible, it will assess the site soil liquefaction potential using the applicable guidance of

Regulatory Guide 1.198 (NRC, 2003).

The NRC staff reviewed the information presented concerning soil liquefaction potential and

find that the because the applicant t will perform liquefaction assessment using

Regulatory Guide 1.198 for final facility design, that staff find this is acceptable. Thus, the

NRC staff concludes that the regulatory requirements in 10 CFR 70.65(b)(1) have been met with

regard to this topic.

1.3.3.4.5 Settlement

The soil settlement, bearing capacity, and static and dynamic soil and rock properties for the

proposed EREF site are not currently available. The applicant states in its ISA Summary

Section 3.2.7 (AES, 2010b) that it will conduct additional geotechnical investigations of the

proposed site to determine static and dynamic soil and rock properties, evaluate foundation

bearing capacity, estimate settlement and differential settlement, and provide geotechnical input

for soil-rock structure interaction analysis to support the final facility design.

For settlement evaluation, the applicant states in its ISA Summary Section 3.2.7

(AES, 201 Ob) that it will use the applicable methods provided in one or more of the following

documents: Naval Facilities Engineering Command (NAVFAC) Design Manual (DM)-7.01, Soil

Mechanics (Department of Navy, 1986a); Foundation Engineering Handbook

(Winterkorn, 1975); Foundation Analysis and Design (Bowles, 1996); and Foundation

Engineering (Peck, 1974). To determine allowable bearing pressure, the applicant states in its

ISA Summary Section 3.2.7 (AES, 201 Ob) that it will follow the applicable guidance in one or

more of the following documents: NAVFAC DM-7.02, Foundations and Earth Structures

(Department of Navy, 1986b); Foundation Engineering Handbook (Winterkorn, 1975);

Foundation Analysis and Design (Bowles, 1996); Foundation Engineering (Peck, 1974); and

Rock Foundations (ASCE, 1996). The NRC staff finds the applicant's plan to conduct an

additional geotechnical investigation to develop design information to support final facility design

acceptable. In addition, the NRC staff finds the guidance the applicant plans to use to obtain

information concerning differential settlements, soil-bearing capacity, and dynamic soil and rock

1-34
properties for the final facility design acceptable. Based on the applicant's plans to conduct

additional geotechnical study and the described guidance, the NRC staff concludes, that the

regulatory requirements in 10 CFR 70.65(b)(1) have been met with regard to these topics.

1.3.4 Evaluation Findings

The staff has reviewed the site description for the proposed EREF according to Section 1.3 of

the Standard Review Plan (NRC, 2002). The applicant has adequately described and

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

relative to prominent natural and manmade features such as 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 staff has verified that the site description is consistent with the information used as a basis

for the ER (USEC, 2003a), the EP (USEC, 2006a), and the ISA Summary (USEC, 2006b); 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

(AES, 2010a) AREVA Enrichment Services LLC, "Eagle Rock Enrichment Facility Safety

Analysis Report," 2010.

(AES, 2010b) AREVA Enrichment Services LLC, "Integrated Safety Analysis Summary for the

Eagle Rock Enrichment Facility," 2010.

(AES, 2010c) AREVA Enrichment Services LLC, "Environmental Report for the Eagle Rock

Enrichment Facility," 2010.

(AES, 2010d) AREVA Enrichment Services LLC, "Eagle Rock Enrichment Facility Standard

Practice Procedure Plan," 2010.

(AES, 2010e) AREVA Enrichment Services LLC, "Security Program for the Eagle Rock

Enrichment Facility," 2010.

(AES, 2010f) AREVA Enrichment Services LLC, "Emergency Plan for the Eagle Rock

Enrichment Facility," 2010.

(AES, 2010g) AREVA Enrichment Services LLC, Email, from Jim Kay of AREVA Enrichment

Services to Breeda Reilly, USNRC, "AES Proposal for Authorization to Make Changes to

License Commitments," dated August 20, 2010.

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(AES, 2009a) AREVA Enrichment Services LLC, "AREVA Enrichment Facility Responses to

U.S. Nuclear Regulatory Commission's Requests for Additional Information." AES-O-NRC-09-

00140-0. Marlborough, Massachusetts: AREVA Enrichment Services LLC. 2009.

(ASCE, 2005) American Society of Civil Engineers, "Seismic Design Criteria for Structures,

Systems, and Components in Nuclear Facilities." ASCE/SEI 43-05. New York City, New York:

American Society of Civil Engineers. 2005.

(ASCE, 1996) "Rock Foundations, Technical Engineering and Design Guides as Adapted from

the U.S. Army Corps of Engineers, Number 16." Reston, Virginia: American Society of Civil

Engineers. 1996.

(ASTM, 2003) American Society for Testing and Materials, ASTM C787-03, "Standard

Specification for Uranium Hexafluoride for Enrichment," 2003.

(Boore, 2008) Boore, D.M. and G.M. Atkinson, "Ground-Motion Prediction Equations for the

Average Horizontal Component of PGAJ PGV and 5%-Damped PSA at Spectral Periods

Between 0.01 s and 10 s." Earthquake Spectra. Vol. 24, No.1. pp. 99-138. 2008.

(Bowles, 1996) Bowles, J.A., Foundation Analysis and Design. 5 th Edition. Columbus, Ohio:

The McGraw-Hill Companies. 1996.

(Bozorgnia, 2004) Bozorgnia, Y. and K.W. Campbell, "The Vertical-to-Horizontal Response

Spectral Ratio and Tentative Procedures for Developing Simplified V/H and Vertical Design

Spectra." Journal of Earthquake Engineering. Vol. 18, No. 2. pp. 175-207. 2004.

(Coats, 1985) Coats, D.W. and R.C. Murray, "Natural Phenomena Hazards Modeling Project:

Extreme Wind/Tornado Hazard Models for Department of Energy Sites," UCRL-53526. Rev. 1.

Livermore, California: Lawrence Livermore National Laboratory. 1985.

(Commerce, 1964) U.S. Department of Commerce, "Two-To-Ten-Day Precipitation for Return

Periods of 2 to 100 Years in the Contiguous United States," 1964.

(Departments of Army, Navy, and Air Force, 1990) Departments of Army, Navy, and Air Force,

"Structures To Resist the Effects of Accidental Explosions." TM-1 300/NAVFAC P-397/

AFR 88-22. Rev. 1. 1990.

(Department of Navy, 1986a) Department of Navy, Naval Facilities Engineering Command,

"Design Manual 7.01 Soil Mechanics." 1986.

(Department of Navy, 1986a) Department of Navy, Naval Facilities Engineering Command,

"Design Manual 7.02 Foundations and Earth Structures." 1986.

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(DOE, 2006) U.S. Department of Energy, "Accident Analysis for Aircraft Crash into Hazardous

Facilities." DOE-STD-3014-2006. 2006.

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Concerning Foreign Ownership, Control, and Influence Determinations Associated with Certain

NRC Licensees. May 6, 2002.

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National Engineering and Environmental Laboratory, Southeast Idaho, in Bonnichsen." C.M.

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Volcanism of the Eastern Snake River Plain, Idaho." P.K. Link, M.A. Kuntz, and L.B. Platt, eds.

Geological Society of America Memoir. Vol. 179. pp. 227-267. 1992.

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Atlas of the Western United States," Atlas 2, Vol. 5, 1973.

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(NRC, 2009) U.S. Nuclear Regulatory Commission, Draft Regulatory Guide, DG-3037,

"Guidance for Fuel Cycle Facility Change Processes," 2009.

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the Review of Safety Analysis Reports for Nuclear Power Plants," 2007.

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for the American Centrifuge Plant in Piketon, Ohio," 2006.

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of the Contiguous United States," Rev. 2, 2006.

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for the National Enrichment Facility in Lea County, New Mexico," 2005.

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the Review of a License Application for a Fuel Cycle Facility," 2002.

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Standard Format and Content for the Protection of Classified Matter for NRC Licensee,

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Criteria for Assessing Seismic Soil Liquefaction at Nuclear Power Plant Sites," 2003.

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of the Contiguous United States." Rev. 2. Washington, DC: NRC. 2006.

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CHAPTER 2.0 ORGANIZATION AND ADMINISTRATION

The purpose of the U.S. Nuclear Regulatory Commission's (NRC's) review of the applicant's

organization and administration is to evaluate whether the license application describes

proposed management policies that provide reasonable assurance that the applicant will plan,

implement, and control site activities in a manner that ensures the safety of workers and the

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

and provided adequate qualification descriptions for key management positions.

2.1 Regulatory Requirements

Title 10 of the Code of Federal Regulations (10 CFR) 70.22(a)(6) requires that the applicant

provide the technical qualifications, including training and experience of the applicant and

members of the staff. In addition, 10 CFR 70.23(a)(2), 10 CFR30.33(a)(3), and

10 CFR 40.32(b) require that an applicant be qualified by reason of training and experience to

use the licensed material for the purpose requested. And, 10 CFR 70.23(a)(4),

10 CFR 30.33(a)(2), and 10 CFR 40.32(c) require that the applicant's proposed equipment and

facilities are adequate to protect health and minimize danger to life or property. Also, the

management measures required under 10 CFR 70.62(d) establish a management system and

administrative procedures that apply to items relied on for safety (IROFS) to ensure their

availability and reliability.

2.2 Regulatory Guidance and Acceptance Criteria

The guidance applicable to the NRC's review of the organization and administration section of

the Safety Analysis Report (SAR) (AES, 201 Oa) is contained in Chapter 2 of the "Standard

Review Plan for the Review of a License Application for a Fuel Cycle Facility," NUREG-1 520

(NRC, 2002). Section 2.4.3 of NUREG-1 520 lists the acceptance criteria for both new and

existing facilities. Applications for new facilities must address both sets of acceptance criteria.

2.3 Staff Review and Analysis

Chapter 2.0 of the SAR (AES, 201 Oa) describes the organizations that will be responsible for

managing the design, construction, operation, and decommissioning of the proposed Eagle

Rock Enrichment Facility (EREF). The key management and supervisory positions and

functions are described, including the personnel qualifications for each key position at the

facility. The shift crew composition is also listed.

2.3.1 Organization

The regulations in 10 CFR 70.23(a)(2), 10 CFR30.33(a)(3), and 10 CFR 40.32(b) require that an

applicant be qualified by reason of training and experience to use the licensed material for the

purpose requested. And, 10 CFR 70.23(a)(4), 10 CFR 30.33(a)(2), and 10 CFR 40.32(c)

require that the applicant's proposed equipment and facilities are adequate to protect health and

minimize danger to life or property. Thus, the applicant must implement an organization and

appropriate administrative elements to support these regulatory requirements. The acceptance

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criteria in Section 2.4.3 of NUREG 1520 (NRC, 2002) provide guidance that the applicant must

provide the following information: (1) identify and functionally describe the specific

organizational groups that are responsible for managing the design, construction, and operation

of the facility, (2) organizational charts, and (3) plans to commission the facility's startup and

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

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

In Section 2.1 of the SAR (AES, 201 Oa), AREVA Enrichment Services, LLC (AES) provided a

functional description of the specific organizational groups responsible for managing the design,

construction, operation, and decommissioning of the EREF. The following discussion

summarizes the description.

As described in Section 2.1 of the SAR (AES, 2010a), the top level of the applicant's

organization is the AES President. The AES President is responsible for the design, quality

assurance (QA), construction, operation, and decommissioning of the EREF (AES, 2010a).

During the engineering, procurement, and construction phase, the Project Director, Plant

Operations Manager, and QA Manager report to the AES President (AES, 2010a). During the

operating phase, the Plant Manager, the QA Manager, and the Safety Review Committee (SRC)

report to the President (AES, 2010a). Figure 1-2 of the Quality Assurance Program Description

(QAPD) (AES, 2010b), "Engineering, Procurement, and Construction Organizational Chart,"

illustrates the authority and lines of communication for the engineering, procurement, and

construction phase. Figure 1-1 of the QAPD (AES, 201 Ob), "Eagle Rock Enrichment Facility

Operations Organizational Chart," shows the authority and lines of communication for the

operating phase. In Sections 2.1.3 and 2.1.4 of the SAR (AES, 201 Oa), AES stated that the

position descriptions of key management personnel for the design, construction organization,

and the operating organization will be accessible to all affected personnel and to the NRC.

AES has contracted with Enrichment Technology Company Limited (ETC) to design the core

process technology (AES, 201 Oa). ETC will design, manufacture, and deliver the centrifuges

necessary for facility operation (AES, 201 Oa). In addition, ETC will supply technical assistance

and consultation services during installation and operation (AES, 2010a). AES will contract with

an architect/engineering firm to further specify, design, and build the supporting structures and

systems of the EREF (AES, 2010).

Section 2.1.4 of the SAR (AES, 2010a) describes AES's approach for the transition from the

design and construction phase to the operations phase. Towards the end of the construction of

the EREF, the focus of the AES organization will shift from design and construction to initial

startup and operation (AES, 2010a). As the facility nears completion, AES will staff the EREF

operating organization to ensure a smooth transition from construction to operations activities

(AES, 2010a). Design and construction personnel will be integrated into the operations

organization to provide technical support during initial startup of the facility and the transition

into the operations phase (AES, 2010a). Also, ETC will have personnel integrated into the AES

organization to provide technical support during startup of the facility and the transition to the

operations phase (AES, 2010a).

As the construction of systems is completed, the systems will undergo required acceptance

testing, followed by turnover from the construction organization to the operating organization by

means of a detailed transition plan (AES, 2010a). This turnover will include the physical

systems and corresponding design information and records (AES, 2010a). After turnover, the

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operating organization will be responsible for system maintenance and configuration

management (AES, 2010a). 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 Measures," of the SAR (AES, 201 Oa). Chapter 11 of this Safety

Evaluation Report (SER) describes the NRC staffs evaluation of the configuration management

system.

As stated above, the applicant has identified and described the proposed organization that

would be responsible for managing the design, construction, and operation of the proposed

facility. The applicant has also provided organization charts. In addition, AES will have written

position descriptions that will be available to all affected personnel and to the NRC. The

proposed organization provides an acceptable management system for ensuring that the

design, construction, and operation of the facility will meet the NRC's regulatory requirements.

The information provided by the applicant is consistent with the guidance in Section 2.4.3 of

NUREG-1520 (NRC, 2002) that the applicant should identify and describe the specific

organizational groups that are responsible for managing the design, construction, and operation

of the facility and provide organizational charts. In addition, the information provided by the

applicant is consistent with the guidance in Section 2.4.3 of NUREG 1520 (NRC, 2002) to

describe specific plans to commission the facility's startup and operation, including the transition

from the startup phase to operations under the direct supervision of the applicant's personnel

responsible for safe operations. The description of the proposed organization is, therefore,

acceptable.

2.3.2 Organizational Responsibilities and Qualifications

The regulations in 10 CFR 70.22(a)(6) require that the applicant provide the technical

qualifications, including training and experience of the applicant and members of the staff. In

addition, the regulations in 10 CFR 70.23(a)(2), 10 CFR30.33(a)(3), and 10 CFR 40.32(b)

require that an applicant be qualified by reason of training and experience to use the licensed

material for the purpose requested. Following the guidance in NUREG-1520 (NRC, 2002), the

staff reviewed the applicant's SAR (AES, 201 Oa) to ensure that the personnel responsible for

managing the design, construction, and operation of the facility have substantive breadth and

level of experience and are appropriately available; that clear, unambiguous management

controls and communications exist among the organizational units that are responsible for

managing the design and construction of the facility; that the qualifications, responsibilities, and

authorities for key supervisory and management positions with health, safety, and environment

(HS&E) responsibilities are clearly defined in position descriptions that are available to all

affected personnel and to the NRC, upon request; and that the individual delegated overall

responsibility for the HS&E functions will have the authority to shut down operations if they

appear to be unsafe and in that case, must approve restart of shutdown operations.

In Section 2.2 of the SAR (AES, 2010a), the applicant provides information concerning the

minimum qualifications, responsibilities, authorities, and lines of communication for key

management personnel. Responsible managers have the authority to delegate tasks to other

individuals; however, the responsible manager retains the ultimate responsibility and

accountability for implementing the applicable requirements (AES, 2010a). According to

Section 2.2.4 of the SAR (AES, 2010a), the AES President will evaluate the nuclear experience

of each individual. "Responsible nuclear experience" for these positions includes:

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(a) responsibility for and contributions toward support of facility(ies) in the nuclear fuel cycle and

(b) experience with chemical materials and/or processes (AES, 201 Oa). The AES President will

approve the assignment of individuals to management positions which report directly to the

President, and to positions on the SRC (AES, 201 Ga). Assignments for all other staff positions

will be made within the normal administrative practices of the facility (AES, 201 Oa).

The qualifications of the individuals assigned to the key facility positions described in

Section 2.2.1 of the SAR will be maintained in employee personnel files or other appropriate

files at the facility (AES, 2010a). Position descriptions of key management personnel will be

accessible to AES personnel and to the NRC (AES, 2010a). Development and maintenance of

qualification records and training programs are the responsibility of the Human Resources

Manager (AES, 2010a).

The AES President has overall responsibility for the design, QA, construction, operation, and

decommissioning of the EREF (AES, 2010a). He or she is also responsible for the QA Program

and for determining the status, adequacy, and effectiveness of its implementation (AES, 201 Oa).

The Plant Manager; the QA Manager; the Human Resources Manager; and the

Communications, Community Affairs Manager will be appointed by, and will report to, the AES

President (AES, 2010a). Their responsibilities are described in Section 2.2.1 of the SAR (AES,

2010a). The minimum qualifications for these positions are provided in Section 2.2.4 of the

SAR (AES, 201 Oa). The following summarizes their responsibilities and minimum qualifications.

* Plant Manager: The Plant Manager will have direct responsibility for operation of the

facility in a safe, reliable, and efficient manner (AES, 2010a). The Plant Manager will be

responsible for the protection of the facility staff and the general public from radiation

and chemical exposure and/or any other consequences of an accident at the facility and

also bears the responsibility for compliance with the facility license (AES, 2010a). The

Plant Manager or designee(s) will have the authority to approve and issue procedures

(AES, 201 Ga). The Plant Manager will be knowledgeable of the enrichment process,

enrichment process controls and ancillary processes, criticality safety control, chemical

safety, industrial safety, and radiation protection program concepts as they apply to the

overall safety of a nuclear facility (AES, 201 Ga). The Plant Manager will have, as a

minimum, a bachelor's degree (or equivalent) in an engineering or scientific field and ten

years of responsible nuclear experience (AES, 2010a).

* QA Manager: The QA Manager will have overall responsibility for development,

management, implementation, and independent oversight of the EREF QA Program

(AES, 201 Ga). The facility line managers and their staff who are responsible for

performing quality-affecting work will be responsible for ensuring implementation of and

compliance with the QA Program (AES, 2010a). The QA Manager position will be

independent from other management positions at the facility to ensure the QA Manager

has direct access to the AES President for matters affecting quality (AES, 2010a). The

QA Manager will have, as a minimum, a bachelor's degree (or equivalent) in an

engineering or scientific field and at least six years of responsible nuclear experience in

the implementation of a QA program (AES, 201 Ga). The QA Manager will have at least

four years experience in a QA organization at a nuclear facility (AES, 2010a).

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* Human Resources Manager: The Human Resource Manager will have responsibility for

community relations, ensuring adequate staffing, and providing administrative support

services to the facility including document control (AES, 2010a). The Human Resources

Manager will have as a minimum, a bachelor's degree in personnel management,

business administration or related field, and three years of appropriate, responsible

experience in implementing and supervising human resource responsibilities at an

industrial facility (AES, 2010a).

* Communications, Community Affairs Manager: The Communications, Community

Affairs Manager will have responsibility for providing information about the facility and

AES to the public and media (AES, 201 Oa). During an abnormal event at the facility, the

Communications, Community Affairs Manager will ensure that the public and media

receive accurate and up-to-date information (AES, 2010a). The Communications,

Community Affairs Manager will have as a minimum, a bachelor's degree in public

relations, political science or business administration and three years of appropriate,

responsible experience in implementing and supervising a community relations program

(AES, 2010a).

The Environmental, Health, Safety and Licensing (EHS&L) Manager; the Project Manager; the

Training Manager; the Uranium Management Manager; and the Operations Manager will report

to the Plant Manager. AES describes the responsibilities for each of these positions in

Section 2.2.1 of the SAR (AES, 201 Oa). The minimum qualifications for these positions are

provided in Section 2.2.4 of the SAR (AES, 201 Oa). The following summarizes the

responsibilities and minimum qualifications for these managers.

EHS&L Manager: The EHS&L Manager will have the overall responsibility for the

development and implementation of programs addressing worker health and safety;

environmental protection; and licensing and permitting (AES, 2010a). The EHS&L

Manager will be responsible for maintaining compliance with safeguards; appropriate

rules, regulations, and codes; and implementation and control of the Fundamental

Nuclear Material Control (FNMC) Plan, the Physical Security Plan (PSP), and the

Standard Practices Procedure Plan for the Protection of Classified Matter (SPPP) (AES,

2010a). EHS&L activities cover nuclear criticality safety, radiation protection, chemical

safety, environmental protection, emergency preparedness, industrial safety, and

development and implementation of security programs (AES, 201 Oa). The EHS&L

Manager will work with the other facility managers to ensure consistent interpretations of

EHS&L requirements, perform independent reviews, and support facility and operations

change control reviews (AES, 2010a). This position will be independent from other

operations management positions at the facility to ensure objective EHS&L audit, review,

and control activities (AES, 20101). The EHS&L Manager will have the authority to

order the shutdown of operations if they appear to be unsafe or non-compliant with

applicable regulatory requirements and must consult with the Plant Manager prior to

restart of shutdown operations after the deficiency, or unsatisfactory condition, has been

resolved (AES, 2010a). The EHS&L Manager will have, as a minimum, a bachelor's

degree (or equivalent) in an engineering or scientific field and at least five years of

responsible nuclear experience in EHS&L or related disciplines (AES, 2010a). The

EHS&L Manager will also have at least one year of direct experience in the

administration of nuclear criticality safety evaluations and analyses (AES, 201 Oa).

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* Project Manager: The Project Manager will have overall responsibility for managing the

engineering, procurement, construction, and startup of facility modifications and

expansion (AES, 2010a). This will include managing the work and contracts with the

technologysupplier (i.e., ETC) (AES, 2010a). The Project Manager will have, as a

minimum, a bachelor's degree (or equivalent) in an engineering or scientific field and

four years of responsible nuclear experience (AES, 201 Oa).

* Training Manager: The Training Manager will have responsibility for the development,

implementation, and administration of the plant training programs, including

maintenance of the plant training database (AES, 201 Oa). The training programs will

address qualifications of workers to perform work as well as required safety training

(AES, 201 Oa). The Training Manager will have a minimum of five years of appropriate,

responsible experience in implementing and supervising a training program (AES,

201 Oa).

* Uranium Management Manager: The Uranium Management Manager will have

responsibility for uranium hexafluoride (UF6) cylinder management (including compliance

with transportation requirements) and directing the scheduling of enrichment operations

(AES, 2010a). During an absence of the Plant Manager, the Uranium Management

Manager may assume the responsibilities and authorities of the Plant Manager (AES,

2010a). The Uranium Management Manager will have, as a minimum, a bachelor's

degree (or equivalent) in an engineering or scientific field and four years of responsible

nuclear experience (AES, 201 Oa).

* Operations Manager: The Operations Manager will have the responsibility of directing

the day-to-day operation of the facility and ensuring that that the operations are

conducted safely and in compliance with any license conditions (AES, 2010a). The

Operations Manager will also be responsible for the plant maintenance function, which

includes activities to assure that Items Relied on for Safety (IROFS) are reliable and

available when needed (AES, 2010a). During the absence of the Plant Manager, the

Operations Manager may assume the responsibilities and authorities of the Plant

Manager (AES, 201 Oa). The Operations Manager will have, as a minimum, a bachelor's

degree (or equivalent) in an engineering or scientific field and four years of responsible

nuclear experience (AES, 2010a).

AES stated that during an absence of the Plant Manager, the Uranium Management Manager or

the Operations Manager may assume the responsibilities and authorities of the Plant Manager

(AES, 2010a). The EHS&L Manager has the authority to order the shutdown of operations if

they appear to be unsafe or non-compliant with applicable regulatory requirements and must

consult with the Plant Manager before restart of shutdown operations after the deficiency, or

unsatisfactory condition, has been resolved (AES, 2010a).

The Radiation Protection/Chemistry Manager; the Nuclear Criticality Safety Manager; the

Licensing and Compliance Manager; the Safeguards Manager; the Safety, Security, and

Emergency Preparedness Manager; and the Industrial Safety Manager report to the EHS&L

Manager. AES described the responsibilities for each of these positions in Section 2.2.1 of the

2-6
SAR (AES, 2010a). The minimum qualifications for these positions are provided in Section

2.2.4 of the SAR (AES, 201 Oa). The following summarizes the responsibilities and minimum

qualifications for these managers.

" Radiation Protection/Chemistry Manager: The Radiation Protection/Chemistry Manager

will have responsibility for developing and implementing programs to limit personnel

radiological exposures and environmental impacts associated with facility operations,

including the As Low as Reasonably Achievable (ALARA) program (AES, 2010a).

During emergency conditions, the Radiation Protection/Chemistry Manager's duties may

expand as described in Section 2.2.1 of the SAR (AES, 2010a). The Radiation

Protection/Chemistry Manager will also be responsible for the implementation of

chemistry analysis programs and procedures for the facility, including, 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 (AES,

2010a). The Radiation Protection/Chemistry Manager will have, as a minimum, a

bachelor's degree (or equivalent) in an engineering or scientific field and four years of

responsible nuclear experience (AES, 201 Ga).

" Nuclear Criticality Safety Manager: The Nuclear Criticality Safety Manager will be

responsible for the development and implementation of the nuclear criticality safety

program (AES, 2010a). Key responsibilities will include the performance of nuclear

criticality safety analyses and evaluations of applicable operations involving special

nuclear material and changes to those operations; establishing limits and controls based

on those analyses and evaluations; assuring the proper incorporation of limits and

controls into applicable procedures and instructions; and monitoring plant compliance

with nuclear criticality safety requirements (AES, 2010a). The Nuclear Criticality Safety

Manager will have, as a minimum, a bachelor's degree (or equivalent) in an engineering

or scientific field and four years of responsible nuclear experience (AES, 201 Ga). The

Nuclear Criticality Safety Manager will also have at least one year of direct experience in

the administration of nuclear criticality safety evaluations and analyses (AES, 201 Oa).

Safeguards Manager: The Safeguards Manager will have responsibility for ensuring the

proper implementation of the FNMC Plan (AES, 2010a). This position will be separate

from and independent of other departments to ensure separation between the

safeguards group and the other departments. In matters involving safeguards, the

Safeguards Manager has direct access to the Plant Manager (AES, 2010a). The

Safeguards Manager will have, as a minimum, a bachelor's degree in an engineering or

scientific field, and five years of experience in the management of a safeguards program

for special nuclear material, including responsibilities for material control and accounting

(AES, 201 Ga). No credit for academic training may be taken toward fulfilling this

experience requirement (AES, 201 Ga).

Safety, Security and Emergency Preparedness Manager: The Safety, Security, and

Emergency Preparedness Manager will be responsible for implementation and

maintenance of the integrated safety analysis, industrial hygiene and safety, chemical

safety, fire protection, security, and emergency preparedness including the responsibility

for ensuring the facility remains prepared to react and respond to any emergency

situation that may arise (AES, 2010a). This will include emergency preparedness

2-7
training of facility personnel, facility support personnel, the training of, and coordination

with, offsite emergency response organizations (EROs), and conducting periodic drills to

ensure the training of facility personnel and offsite response organization personnel is

maintained up-to-date (AES, 2010a). The Safety, Security and Emergency

Preparedness Manager will also be 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 Safety,

Security and Emergency Preparedness Manager has direct access to the Plant Manager

(AES, 201 Ga). The Safety, Security and Emergency Preparedness Manager will have,

as a minimum, a bachelor's degree in an engineering or scientific field, and five years of

experience in the responsible management of physical security at a facility requiring

security capability similar to that required for the facility (AES, 2010a). No credit for

academic training may be taken toward fulfilling this experience requirement (AES,

201 Oa).

Industrial Safety Manager: The Industrial Safety Manager will have responsibility for the

implementation of industrial safety programs and procedures, including programs and

procedures for training individuals in safety and maintaining the performance of the

facility fire protection systems (AES, 2010a). The Industrial Safety Manager will have,

as a minimum, a bachelor's degree (or equivalent) (AES, 2010a).

The Engineering Manager; the Procurement Manager; the Construction Manager; the Startup

Manager; and the Information Technology Manager will report to the Project Manager. AES

describes the responsibilities for each of these positions in Section 2.2.1 of the SAR (AES,

201 Ga). The minimum qualifications for these positions are provided in Section 2.2.4 of the

SAR (AES, 2010). The following summarizes the responsibilities and minimum qualifications for

these managers.

" Engineering Manager: The Engineering Manager will be responsible for site

characterization; facility design and the design control process; configuration

management; engineering; and acceptance test coordination, including test control of

facility modifications and expansion (AES, 2010a). The Engineering Manager will be

also responsible for records management and document control, and approving

disposition of nonconforming items when dispositioned as "repair" or "use-as-is" during

operations (AES, 2010a). 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 five years of appropriate, responsible experience in implementing and

supervising a nuclear engineering program (AES, 2010a).

* Procurement Manager: The Procurement Manager will be responsible for procurement;

providing procurement material control services (including supplier qualification

coordination, purchasing, contracting, receiving and control of nonconforming items);

and material control (including handling, storage and shipping) (AES, 2010a). The

Procurement Manager will be responsible for supply strategy and development of

qualified long-lead-time and complex-system suppliers (AES, 2010a). The Procurement

Manager will have, as a minimum, a bachelor's degree (or equivalent) in an engineering

or business field and have a minimum of five years of responsible experience in

purchasing and supply chain management (AES, 201 Oa).

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* Construction Manager: The Construction Manager will be responsible for managing the

construction of facility modifications and expansion to the EREF (AES, 2010a). This

responsibility will include managing the activities of qualified contractors in the

preparation of construction documents and the construction of facility modifications and

expansion (AES, 201 Oa). The Construction Manager will have, as a minimum, a

bachelor's degree (or equivalent) in an engineering or scientific field and have a

minimum of five years of appropriate, responsible experience in implementing and

supervising a nuclear construction program (AES, 2010a).

" Startup Manager: The Startup Manager will be responsible for the overall preoperational

and startup test program of facility modifications and expansion (AES, 2010a). This

individual will be responsible for the development of preoperational and startup test

procedures, providing technical advice to personnel conducting the tests, briefing

personnel responsible for operation of the plant during the tests, ensuring that the tests

are performed in accordance with the applicable procedures, and generating test reports

(AES, 2010a). The Startup Manager will have, as a minimum, a bachelor's degree (or

equivalent) in an engineering or scientific field and have a minimum of five years of

responsible experience in nuclear plant operations and maintenance (AES, 2010a).

* Information Technology (IT) Manager: The IT Manager will be responsible for

maintaining all computer software programs related to the nuclear material accounting at

EREF (AES, 2010a). This individual will also be responsible for EREF computer

database for generation of nuclear material control charts (AES, 2010a). The IT

Manager will have, as a minimum, a bachelor's degree (or equivalent) in computer

science, and five years of experience in the computer related field (AES, 2010a).

The Cylinder Management Manager; Production Scheduling Manager; and the Warehouse and

Materials Manager will report to the Uranium Management Manager. The following summarizes

the responsibilities and minimum qualifications for these managers.

* Cylinder Management Manager: The Cylinder Management Manager will have

responsibility for ensuring that cylinders of UF6 are received and routed correctly at the

facility, and will be responsible for all transportation licensing (AES, 201 Oa). The

Cylinder Management Manager will have a minimum of three years of appropriate,

responsible experience in implementing and supervising a continuous production

scheduling program (AES, 2010a).

" Production Scheduling Manager: The Production Scheduling Manager will have the

responsibility for developing and maintaining production schedules for enrichment

services (AES, 2010a). The Production Scheduling Manager will have a minimum of

three years of appropriate, responsible experience in implementing and supervising a

continuous production scheduling program (AES, 2010a).

* Warehouse and Materials Manager: The Warehouse and Materials Manager will have

the responsibility for ensuring spare parts and other materials needed for operation of

the facility are ordered, received, inspected, and stored properly (AES, 2010a). The

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Warehouse and Materials Manager will have a minimum of three years of appropriate,

responsible experience in implementing and supervising a purchasing and inventory

program (AES, 2010a).

The Production Managers and the Maintenance Manager will report to the Operations Manager.

The following summarizes the responsibilities and minimum qualifications for these managers.

" Production Managers: The Production Managers will be responsible for enrichment

operations, feed and withdrawal operations, utilities, shift operations, packaging, and

transportation (AES, 2010a). Production Managers will have a minimum of five years of

appropriate, responsible experience in implementing and supervising a nuclear

operations program (AES, 2010a).

" Maintenance Manager: The Maintenance Manager will have the responsibility of

directing and scheduling maintenance activities to ensure proper operation of the facility,

including preparation and implementation of maintenance procedures (AES, 2010a).

The Maintenance Manager will also have responsible for safe and reliable performance

of preventive and corrective maintenance and support services on buildings/facilities and

equipment (including IROFS), and for integrated planning and scheduling (AES, 201 Oa).

In addition, the Maintenance Manager will coordinate and maintain testing programs for

the facility (AES, 201 Oa). This will include testing of systems and components to ensure

the systems and components are functioning as specified in design documents (AES,

201 Oa). The Maintenance Manager will have, as a minimum, a bachelor's degree (or

equivalent) in an engineering or scientific field and four years of responsible nuclear

experience (AES, 201 Oa).

Other key personnel will include the Measurement Control Program Manager; the Production

Supervisors; the Criticality Safety Engineer; the Chemical Safety Engineer; the QA Inspectors;

the QA Auditors; the QA Technical Support, the Administration Manager; and the Document

Control Manager. AES describes the responsibilities of each of these positions in Section 2.2.1

of the SAR (AES, 201 Oa). The minimum qualifications for these positions are provided in

Section 2.2.4 of the SAR (AES, 201 Oa). (AES, 201 Oa) The following summarizes the

responsibilities and minimum qualifications for these managers and staff.

* Measurement Control Program Manager: The Measurement Control Program Manager

will report to the Safeguards Manager and will have responsibility for the EREF

Measurement Control Program (AES, 2010a). The EREF Measurement Control

Program will be provided to ensure adequate calibration frequencies, sufficient control of

biases, and sufficient measurement precisions for nuclear material control and

accounting (AES, 2010a). The Measurement Control Program Manager will have, as a

minimum, a bachelor's degree in an engineering or scientific field, and five years of

experience in the management control program (AES, 2010a).

" Production Supervisors: The Production Supervisors will report to their respective

Production Managers (AES, 2010a). The Production Supervisors will be responsible for

control of materials, personnel, equipment, and activities in specific areas (AES, 2010a).

These responsibilities will include assuring that formal approved procedures are

available and adhered to by operators and other applicable personnel (AES, 2010a).

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Production Supervisors will have a minimum of three years of appropriate, responsible

experience in implementing and supervising a nuclear operations program (AES,

201 Oa).

Criticality Safety Engineer: Criticality Safety Engineers will report to the Nuclear

Criticality Safety Manager (via a designated supervisory position, if applicable) (AES,

2010a). Criticality Safety Engineers will have responsibility for the preparation and/or

review of nuclear criticality safety evaluations and analyses and for conducting and

reporting periodic nuclear criticality safety assessments (AES. 2010a). Nuclear criticality

safety evaluations and analyses require independent reviews by a Criticality Safety

Engineer. Criticality Safety Engineers will hold a Bachelor of Science or Bachelor of Arts

degree in an engineering or scientific field and will have successfully completed a

training program, applicable to the scope of operations, in the physics of criticality and in

associated safety practices (AES, 201 Oa).

Should a change to the facility require a nuclear criticality safety evaluation or analysis,

an individual who, as a minimum, possesses the equivalent qualifications of the

Criticality Safety Engineer will perform the evaluation or analysis (AES, 2010a). In

addition, this individual will have at least two years of experience performing criticality

safety analyses and implementing criticality safety programs (AES, 201 Oa). An

independent review of the evaluation or analysis will be performed by a qualified

Criticality Safety Engineer (AES, 201 Oa).

* Chemical Safety Engineer: The Chemical Safety Engineer will report to the Radiation

Protection/Chemistry Manager (via a designated supervisory position, if applicable) and

will be responsible for the preparation and/or review of chemical safety programs and

procedures for the facility (AES, 201 Oa). The Chemical Safety Engineer will have a

minimum of two years experience in the preparation and/or review of chemical safety

programs and procedures (AES, 2010a). This individual will hold a bachelor's degree

(or equivalent) in an engineering or scientific field and will have successfully completed a

training program, applicable to the scope of operations, in chemistry and in associated

safety practices (AES, 201 Oa).

" QA Inspectors: The QA Inspectors will report to the QA Manager (via a designated

supervisory position, if applicable) and will have the responsibility for performing

inspections related to the implementation of the AES QA Program (AES, 2010a).

* QA Auditors: The QA Auditors will report to the Quality Assurance Manager (via a

designated supervisory position, if applicable) and will have the responsibility for

performing audits related to the implementation of the AES QA Program (AES, 201 Oa).

* QA Technical Support: The QA Technical Support personnel report to the Quality

Assurance Manager (via a designated supervisory position, if applicable) and will have

the responsibility for providing technical support related to the implementation of the

AES QA Program (AES, 201 Oa).

" Administration Manager: The Administration Manager will report to the Human

Resources Manager and will have the responsibility for ensuring that support functions

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such as accounting, word processing, and general office management are provided for

the EREF (AES, 2010a). The Administration Manager will have a minimum of three

years of appropriate, responsible experience in implementing and supervising

administrative responsibilities at an industrial facility (AES, 201 0a).

Document Control Manager: The Document Control Manager will report to the Human

Resources Manager (AES, 2010a). The Document Control Manager will have a

minimum of three years of appropriate, responsible experience in implementing and

supervising a document control program (AES, 2010a).

In addition, AES briefly discussed the shift crew composition at the EREF in Section 2.2.2 of the

SAR (AES, 2010a). The minimum operating shift crew will consist of a Production Supervisor

(or Deputy Production Supervisor in the absence of the Production Supervisor), one Control

Room operator, one radiation protection technician, one operator for each Cascade Hall and

associated UF6 handling systems, and security personnel (AES, 201 Oa). When only one

Cascade Hall is in operation, a minimum of two operators will be required (AES, 2010a). At

least one criticality safety engineer will be available, with appropriate ability to be contacted by

the Production Supervisor, to respond to any routine request or emergency condition (AES,

2010a). This availability may be offsite if adequate communication ability is provided to allow

response as needed (AES, 201 Oa).

The staff concludes that the applicant has identified the responsibilities, qualifications, and

authorities of the key personnel responsible for managing the design, construction, and

operations of the proposed facility and for health, safety, and engineering responsibilities.

These responsibilities, qualifications, and authorities, as described above, are clearly defined

and include sufficient breadth and level of experience to ensure that competent managers will

be in place. The information the applicant provided is consistent with the guidance in Section

2.4.3(3) of NUREG-1 520 (NRC, 2002) that the personnel responsible for managing the design,

construction, and operation of the facility have substantive breadth and level of experience and

will be available; that the qualifications, responsibilities, and authorities for key supervisory and

management position descriptions will be accessible to affected personnel and to the NRC,

upon request, that clear management controls and communications will exist among

organizational units, and that the individual with overall health, safety, and environment

responsibilities is authorized to shut down operations if they appear to be unsafe. The staff

therefore finds the applicant's information acceptable.

2.3.3 Management Control

The management measures required under 10 CFR 70.62(d) establish the management system

and administrative procedures that apply to IROFS to ensure their availability and reliability.

Following the guidance in NUREG 1520, Section 2.4.3 (NRC, 2002), the staff reviewed the

applicant's SAR (AES, 201 Oa) to ensure that the health, safety, and environment organizations

are independent of the operations organizations, allowing objective audits, reviews, or control

activities; that activities essential for effective implementation of the health, safety, and

environment functions are documented in formally approved, written procedures and prepared

in compliance with a formal document control program; that a simple mechanism will be

available for use by any person in the plant, for reporting potentially unsafe conditions or

activities to the health, safety, and environment organization; that formal management

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measures will be established to ensure the availability and reliability of IROFS; and that written

agreements will exist with offsite emergency resources such as fire, police, ambulance/rescue

units, and medical services. Section 2.3 of the SAR (AES, 201 Oa) summarizes how the

activities that are essential for implementation of the management measures and other EHS&L

functions are documented in formally approved, written procedures and prepared in compliance

with a formal document control program.

The QA Manager position will be independent from other management positions at the facility to

ensure that the QA Manager has direct access to the AES President for matters affecting quality

(AES, 2010a). The EHS&L Manager will be independent from other management positions at

the facility to ensure objective EHS&L audit, review, and control activities (AES, 2010a). In

matters involving physical protection of the facility or classified matter, the Safety, Security &

Emergency Preparedness Manager will have direct access to the Plant Manager. The

Safeguards Manager position is separate from and independent of other departments to ensure

separation between the safeguards group and the other departments. Thus, the QA, EHS&L,

Safety, Security & Emergency Preparedness, and Safeguards Managers are independent from

the Operations Managers. In addition, as described in Sections 2.3.1 and 2.3.2 of this SER, the

organizations having QA, engineering, safety and health, environmental, security, safeguards,

and operations responsibilities have clear and well-defined lines of communication and

authority.

In addition, AES will establish management measures to ensure compliance with the

performance requirements of 10 CFR 70.61. Management measures, defined in 10 CFR 70.4,

include configuration management, maintenance, training, procedures, and other QA measures.

A configuration management program is provided for IROFS throughout facility design,

construction, testing, and operation; this is a means of establishing and maintaining a technical

baseline for the facility based on clearly defined requirements. Section 11.3.1 of this SER

evaluates the configuration management program.

In Section 2.3 of the SAR, AES described plans to implement a maintenance program for the

operations phase of the facility that will include planned and scheduled preventive maintenance,

surveillance, and performance trending to ensure that IROFS will be available and reliable to

perform their intended safety functions.

AES will establish a formal training program that will include indoctrination training for all

employees and will address criticality, radiological, chemical, and industrial safety; ALARA

practices; and emergency procedures. In-depth training programs shall be provided to

individuals, depending on specific job requirements, addressing radiological safety (for all

personnel with access to the Restricted Area) and criticality safety control. Refresher training

on radiological and criticality safety will be provided at least annually. Changes to the training

program will be implemented to address incidents potentially compromising safety or changes to

facilities or processes. Training records will be maintained by the Human Resources Manager.

Evaluation of the applicant's training program is provided in Section 11.3.3 of this SER.

AES will conduct all activities involving licensed materials in accordance with approved, written

procedures. These plant procedures will generally include operating, administrative,

maintenance, and emergency procedures.

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In Section 2.3.5.3 of the SAR (AES, 2010a), AES stated that the facility operating organization

will provide, as part of the routine supervisory function, timely and continuing monitoring of

operating activities to keep the Plant Manager current on general facility conditions and to verify

that the day-to-day operating activities are conducted safely and in accordance with applicable

administrative controls.

AES described the SRC for the EREF in Section 2.2.3 of the SAR (AES, 201 Oa). The SRC will

report to the AES President and provide technical and administrative review and audit of

operations that could impact the safety of workers, members of the public, or the environment.

The SAR also describes the scope of activities to be reviewed and audited by the committee

and the frequency of these audits. The SRC will be composed of at least 5 members, to include

experts in operations, criticality safety, radiological safety, chemical safety, and industrial safety.

The members, including the chairman and alternate members, will be formally appointed by the

AES President. Members of the SRC will have an academic degree in an engineering or

physical science field; and a minimum of 5 years of technical experience, of which a minimum of

3 years shall relate directly to one or more of the safety disciplines (criticality, radiological,

chemical, or 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 the NRC. After a

reportable incident, the SRC will review the incident's causes, the responses, and both specific

and generic corrective actions to ensure that the problem is resolved. A written report of each

SRC meeting and audit will be forwarded to the AES President, the Plant Manager, and other

appropriate managers within 30 days, and be retained in accordance with the EREF records

management system.

The AES will implement a QA Program that requires periodic audits of activities affecting quality,

by the AES QA Department, to ensure that these activities are being conducted in accordance

with QA Program requirements and established procedures. Section 11.3.5 of this SER

evaluates audits and assessments.

In Section 2.3.5.4 of the SAR (AES, 201 Oa), AES stated that audited organizations will correct

identified deficiencies in a timely manner. Audited organizations will respond to each audit

report within the time period specified in the audit; and, for each identified deficiency, identify the

corrective action taken or to be taken-whether or not the deficiency is considered to be

indicative of other problems-and the corrective action taken or not to be taken for any such

problems determined. Copies of audit reports and responses will be maintained in accordance

with AES's records management system.

A Corrective Action Program (CAP) will be implemented in order to identify, report, evaluate,

and investigate abnormal events that have the potential to threaten or lessen the effectiveness

of AES's health, safety, or environmental protection programs. Written procedures, to be

followed in the case of an abnormal event, will address incident identification, investigation, root

cause analysis, environmental protection analysis, recording, reporting, and follow-up.

Section 11.3.6 of this SER evaluates incident investigations and the CAP.

In Section 2.3.7 of the SAR (AES, 2010a), AES stated that employees who feel that safety or

quality is being compromised have the right and responsibility to initiate a "stop work" process.

The process is implemented through project or facility procedures; line management or other

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facility management; the safety organization (i.e., any of the safety engineers or managers); the

NRC's requirements under 10 CFR Part 19, "Notices, Instructions and Reports to Workers:

Inspection and Investigations;" and the CAP.

In Section 2.3.8 of the SAR (AES, 2010a), AES discussed the implementation of a records

management program to control the preparation and issuance of documents, including any

changes to these documents. Document control procedures will be established to formalize the

process of reviewing, approving, releasing, transmitting, and distributing new and revised

documents and for the destruction or retention of superseded documents. Additional details on

the records management program are discussed and evaluated in Section 11.3.7 of this SER.

The Emergency Plan for the EREF includes a description of the ERO and of coordination with

offsite EROs through written agreements. Evaluation of emergency management is provided in

Chapter 8 of this SER.

Based on the above discussion, the staff concludes that an EHS&L Manager position will be

established independent of other management positions to ensure objective EHS&L audits,

reviews, and control activities; that activities involving licensed material will be documented in

formally approved, written procedures; that document control procedures will be established for

the review, approval, release, and distribution of new and revised documents; that a "stop work"

process will be available for use by employees for reporting situations where safety is being

compromised; that management measures will be established to ensure the availability and

reliability of IROFS; and that written agreements with offsite emergency resources will be

implemented. (The management measures are evaluated in Chapter 11 of this SER.

Emergency management is evaluated in Chapter 8 of this SER.) The information that the

applicant provided is consistent with the guidance in Section 2.4.3 of NUREG 1520 (NRC, 2002)

and is, therefore, acceptable.

2.4 Evaluation Findings

The staff has reviewed the organization and administration for the EREF in accordance with the

acceptance criteria in Chapter 2 of NUREG-1 520 (NRC, 2002). The staff reviewed the

applicant's organization, key management position summaries and qualifications, and

management controls for providing adequate safety management and management measures

for the safe operation of the facility. These organizational and administrative elements describe:

(1) clear responsibilities and associated resources for the design, construction, and operation of

the facility and (2) the applicant's plans for managing and operating the project. The staff has

reviewed these plans as described in the SAR (AES, 201 Oa) and concludes that they provide

reasonable assurance that an acceptable organization, administrative policies, and sufficient

competent resources have been or will be established in a manner that will allow for the design,

construction, and safe operation of the facility.

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2.5 References

(AES, 2010a) AREVA Enrichment Services LLC. "Eagle Rock Enrichment Facility Safety

Analysis Report," 2010.

(AES, 2010b) AREVA Enrichment Services LLC. "Eagle Rock Enrichment Facility Quality

Assurance Program, Revision 2," 2009.

(NRC, 2002) U.S. Nuclear Regulatory Commission. NUREG-1520, "Standard Review Plan for

the Review of a License Application for a Fuel Cycle Facility," 2002.

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CHAPTER 3.0 INTEGRATED SAFETY ANALYIS (ISA) AND ISA

SUMMARY

The purpose of the U.S. Nuclear Regulatory Commission's (NRC's) review of the applicant's

Integrated Safety Analysis (ISA) and ISA Summary is to evaluate whether the application meets

the regulatory requirements specified in Title 10 of the Code of Federal Regulations (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 through independent analysis or qualitative

evaluations of accepted engineering practices. The review also determines 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.

In particular, the review as described in this chapter considered information provided by the

applicant (that is not security-related nor export controlled information) that is related to:

* Commitments regarding the applicant's safety program, including the ISA, pursuant to the

requirements of 10 CFR 70.62; and

" The ISA Summary submitted in accordance with 10 CFR 70.62(c)(3)(ii) and 70.65.

The staffs evaluation of the use of baseline design criteria (BDC) for the design of the facility in

accordance with 10 CFR 70.64(a), the applicants defense-in-depth practices in accordance with

10 CFR 70.64(b), and those sections of the ISA Summary which contain security-related or

export controlled information is discussed in Appendix A of this SER.

3.1 Regulatory Requirements

The following regulatory requirements are applicable to the ISA and ISA Summary content:

* 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;

* 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;

* 10 CFR 70.64 specifies requirements for baseline design criteria and facility and system

design and facility layout; and

* 10 CFR 70.65(b) describes 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 10 CFR 70.61.

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The safety program is required to contain: (1) process safety information, (2) an ISA, and

(3) management measures. The ISA must be conducted and maintained by the applicant and

must identify the following, in accordance with 10 CFR 70.62(c):

* Radiological hazards related to possessing or processing licensed material at the facility;

* Chemical hazards of licensed material and hazardous chemicals produced from licensed

material;

" Facility hazards that could affect the safety of licensed materials and thus present an

increased radiological risk;

* Potential accident sequences caused by process deviations or other events internal to the

facility and credible external events, including natural phenomena;

* Consequences and likelihood of occurrence of each potential accident sequence identified

and the methods used to determine the consequences and likelihoods; and

" Each IROFS identified pursuant to 10 CFR 70.61(e), the characteristics of its preventative,

mitigative, or other safety function, and the assumptions and conditions under which the

item is relied upon to support compliance with the performance requirements of

10 CFR 70.61.

The regulations, in 10 CFR 70.61, provide that the ISA must evaluate compliance with

performance requirements. The requirements in 10 CFR 70.61(b) specify that the risk of each

credible, high-consequence event must be limited such that the likelihood of occurrence is

highly unlikely; and the requirements in 10 CFR 70.61(c) specify that the risk of each credible,

intermediate-consequence event must be limited such that the likelihood of occurrence is

unlikely.

The 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.

As outlined in 10 CFR 70.65(b), the ISA 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,

including an identification of the controlled area boundaries;

* A description of each process analyzed in the ISA in sufficient detail to understand the

theory of operation; and, for each process, the hazards that were identified in the ISA

pursuant to § 70.62(c)(1 )(i)-(iii) and a general description of the types of accident

sequences;

* Information that demonstrates the licensee's compliance with the performance requirements

of 10 CFR 70.61, including a description of the management measures; the requirements for

criticality monitoring and alarms in 10 CFR 70.24; and, if applicable, the requirements of

10 CFR 70.64;

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* A description of the team, qualifications, and the methods used to perform the ISA;

* A list briefly describing each IROFS identified pursuant to § 70.61(e) in sufficient detail to

understand their functions in relation to the performance requirements of 10 CFR 70.61;

* A description of the proposed quantitative standards used to assess the consequences to

an individual from acute chemical exposure to licensed material or chemicals produced from

licensed materials which are on-site, or expected to be on-site as described in § 70.61(b)(4)

and (c)(4);

* A descriptive list that identifies all IROFS that are the sole item preventing or mitigating an

accident sequence that exceeds the performance requirements of 10 CFR 70.61; and

• A description of the definitions of unlikely, highly unlikely, and credible, as used in the

evaluations in the ISA.

3.2 Regulatory Guidance And Acceptance Criteria

The guidance applicable to the NRC's review of the applicant's ISA and ISA Summary

(AES, 2010b) is contained in Chapter 3 of "Standard Review Plan for the Review of a License

Application for a Fuel Cycle Facility," NUREG-1 520 (NRC, 2002). These sections are

applicable in their entirety with three exceptions. The first exception is 3.4.3.2(4)(c) which

addresses criticality monitoring because criticality monitoring is addressed in Chapter 5 of this

SER. The second exception is Section 3.4.3.2(5)(b)(i-ix) regarding process hazard analysis

methods. This section provides conditions which should be met for hazard analysis methods

used by the applicant if the methods are not described in NUREG-1 513 (NRC, 2001). Because

the method used by the applicant, the hazard and operability method (HAZOP), is described in

NUREG-1 513 (NRC, 2001), these conditions in Section 3.4.3.2(5)b(i-ix) do not have to be

addressed. The third exception is various subparts in Section 3.4.3.2(9) regarding qualitative

methods of defining and evaluating likelihood because the applicant uses a quantitative method.

The acceptance criteria applicable to this review are contained in Sections 3.4.3.1 and 3.4.3.2 of

NUREG-1520 (NRC, 2002).

3.3 Staff Review and Analysis

This section contains the staffs programmatic review of the applicant's proposed safety

program, the proposed ISA commitments, proposed ISA methodology, proposed BDC, and

proposed defense-in-depth. The staff's review of sensitive information, including export

controlled information provided in the ISA and ISA Summary (AES, 2010b) is found in

Appendix A to this SER. The staff's review of other information in the ISA, as determined from

onsite reviews, is provided in Appendix A, Section A.3.2, of this SER.

3.3.1 Safety Program and ISA Commitments

The staff reviewed the applicant's proposed safety program commitments identified in

Section 3.1 and Chapter 11 of the Safety Analysis Report (SAR) (AES, 2010a) to determine

whether the three elements of process safety information, the ISA, and management measures

demonstrate compliance with the requirements of 10 CFR 70.62; and that records will be

established and maintained for documenting each discovery that an IROFS or management

measure has failed or degraded such that it cannot perform its intended safety function.

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3.3.1.1 Process Safety information

According to Section 3.0.1 of the SAR (AES, 2010a), the applicant has compiled and maintains

process information addressing:

* The hazards of materials used or produced in the process - including information on

chemical and physical properties (e.g., toxicity, acute exposure limits, reactivity, and

chemical and thermal stability);

* The description of the technology of the process - including block flow diagrams or

simplified process flow diagrams, a brief outline of process chemistry, safe upper and lower

limits for controlled parameters (e.g., temperature, pressure, flow, and concentration), and

evaluation of the health and safety consequences of process deviations; and

* Equipment used in the process, which includes general information on topics such as the

materials of construction, piping and instrumentation diagrams, ventilation, design codes

and standards employed, material and energy balances, IROFS, electrical classification,

and relief system design.

The process-safety information described above will be maintained up-to-date by the

configuration management program described in Section 11.1 of the SAR, "Configuration

Management." As discussed in Section 11.3.1 of this SER, the applicant uses its configuration

management system to control documentation and review design changes.

AES has developed procedures and criteria for changing the ISA. These include

implementation of a facility change mechanism that meets the requirements of 10 CFR 70.72.

The development and implementation of procedures is described in Section 11.4 of the SAR,

"Procedures Development and Implementation" and is evaluated in Section 11.3.4 of this SER.

Therefore, the staff concludes that the above-mentioned program elements provide reasonable

assurance of the following:

1. Consistent with Section 3.4.3.1(1)(a) of NUREG-1 520 (NRC, 2002), the SAR (AES,

2010a) contains commitments to compile and maintain an up-to-date database of

process safety information and is, therefore, acceptable, and

2. Consistent with Section 3.4.3.1(1)(b) of NUREG-1 520 (NRC, 2002), the process safety

element of the applicant's safety program includes procedures and criteria for changing

the ISA, along with a commitment to design and implement a facility change mechanism,

and is, therefore, acceptable.

As described in Section 3.2 of the SAR, AES uses personnel with expertise in engineering,

safety analysis, and enrichment process operations and experience (individually or collectively)

in nuclear criticality safety, radiological safety, fire safety, chemical safety, operations and

maintenance, and ISA methods to maintain the ISA. 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. The qualifications of the ISA team are

evaluated in Section A.3.1.5 of this SER. Training and qualifications of individuals responsible

for maintaining the ISA are described in the following sections of the SAR: Section 11.3,

"Training and Qualifications" (and evaluated in Section 11.3.3 of this SER), Section 2.2, "Key

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Management Positions," and Section 3.2, "Integrated Safety Analysis Team." Therefore, the

staff concludes that consistent with Section 3.4.3.1(1)(c) of NUREG-1 520 (NRC, 2002), the

process safety element of the applicant's safety program contains a commitment to engage

personnel with appropriate experience and expertise in engineering and process operations,

and is, therefore, acceptable.

3.3.1.2 ISA Commitments

In Section 3.0.2 of the SAR (AES, 201 Oa), the applicant identifies ISA program elements that

were used to establish the ISA process. Those elements include the performance of an ISA for

each process that identifies the radiological hazards, chemical hazards that could increase

radiological risk, chemical hazards from materials involved in processing licensed material,

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. The staffs evaluation of the applicant's methods and criteria for implementing

the ISA methodology is contained in Section 3.3.2 of this SER.

AES has implemented programs to maintain the ISA and supporting documentation so that it is

accurate and up-to-date. Changes to the ISA Summary will be submitted to the NRC, in

accordance with 10 CFR 70.72(d)(1) and (3). The ISA update process accounts for any

changes made to the facility or its processes.

Personnel used to update and maintain the ISA and ISA Summary (AES, 201 Ob) are trained in

the ISA methods(s). The general training and qualification of personnel used to update or

maintain the ISA are described in Section 11.3, "Training and Qualification," of this SER. The

ISA Summary (a non-public document) contains a list of the ISA team members, their areas of

expertise, qualifications, and experience.

Proposed changes to the facility or its operations are evaluated by the ISA method(s) described

in the ISA Summary. New or additional IROFS and appropriate management measures are

designated as required. The adequacy of existing IROFS and associated management

measures are promptly evaluated to determine if they are impacted by changes to the facility

and/or its processes.

Unacceptable performance deficiencies associated with IROFS will be addressed and identified

through updates to the ISA.

Written procedures will be maintained onsite. Section 11.4 of the SAR discusses the applicant's

procedures program.

All IROFS will be maintained so that they are available and reliable when needed.

Items A & B of Section 3.0.2 of the SAR (AES, 201 Oa) contain commitments to conduct an ISA

of appropriate complexity for each process. These commitments are consistent with

Section 3.4.3.1(2)(a) of NUREG-1 520 (NRC, 2002).

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Item B of Section 3.0.2 of the SAR (AES, 201 Oa) contains a commitment to maintain the ISA

and its supporting documentation. This commitment is consistent with Section 3.4.3.1(2)(b) of

NUREG-1520 (NRC, 2002).

Item C of Section 3.0.2 of the SAR (AES, 201 Ga) contains a commitment to train personnel in

the facility's ISA methods and/or use suitably qualified personnel to update and maintain the ISA

and ISA Summary (AES, 2010b). This commitment is consistent with Section 3.4.3.1(2)(c) of

NUREG-1520 (NRC, 2002).

Item D of Section 3.0.2 of the SAR (AES, 2010a) contains a commitment to evaluate proposed

changes to the facility or its operations by means of the ISA method(s) and to designate new or

additional IROFS and appropriate management measures. This commitment is consistent with

Section 3.4.3.1(2)(d) of NUREG-1 520 (NRC, 2002).

Item E of Section 3.0.2 of the SAR (AES, 2010a) contains a commitment to address any IROFS'

unacceptable performance deficiencies that are identified through updates to the ISA. This

commitment is consistent with Section 3.4.3.1(2)(e) of NUREG-1 520 (NRC, 2002).

Item F of Section 3.0.2 of the SAR (AES, 2010a) contains a commitment to maintain written

procedures onsite. This commitment is consistent with Section 3.4.3.1(2)(f) of NUREG-1 520

(NRC, 2002).

Item G of Section 3.0.2 of the SAR (AES, 2010a) contains a commitment to establish all IROFS

(if not already established) and to maintain them so that they are available and reliable when

needed. This commitment is consistent with Section 3.4.3.1(2)(g) of NUREG-1 520

(NRC, 2002).

3.3.1.3 Management Measures

In Section 3.0.3 of the SAR (AES, 2010a), 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 (AES, 2010b). 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, 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 to

provide identifiable and retrievable documentation. The management measures are further

detailed in Chapter 11 of the SAR (AES, 201 Ga), and evaluated in Chapter 11 of this SER.

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3.3.1.4 Safety Program and ISA Commitments Conclusion

Based on the evaluations in Sections 3.3.1 through 3.3.3, above, 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,

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.

3.3.2 ISA Methodology

The ISA methodology is described in Section 3.1 of the SAR (AES, 201 Oa) and Section 3.1.1 of

the ISA Summary (AES, 2010b). The applicant described the approach used for performing the

ISA as a semi-quantitative risk index method to categorize accident sequences in terms of their

likelihood of occurrence and their consequences of concern. Further, the applicant stated that

the risk index method identifies which accident sequences have consequences that could

exceed the performance requirements of 10 CFR 70.61 and, therefore, require designation of

IROFS and supporting management measures.

3.3.2.1 Hazard Identification and Evaluation

The applicant used the HAZOP to identify potential hazards related to uranium hexafluoride

(UF6) process systems and Technical Support Building (TSB) systems. 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 which do not meet the 10 CFR 70.61

performance requirements.

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-1 513, "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.

As stated in Section 3.1 of the SAR (AES, 2010a), 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 the process systems (e.g., UF6 feed system). However, hazards

related to fires and external events were assessed on a facility-wide basis (Section 3.1 of the

SAR). The Fire Hazard Analysis (discussed in Section 7.2 of the SAR [AES, 2010a]) 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,

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highway, railroad, and internal flooding from above-ground storage tanks. The facility wide

assessment resulted in natural phenomena events being assessed against all structures without

regard to location or design differences and fires were assessed by plant area (or fire area) and

included all possible fire hazards within the area. These assessments by the applicant are

evaluated by the staff in the discipline related chapters of the SER (primarily, Chapters 5, 6, and

7 and Appendices A, B, D, E, F, and G).

As described in Section 3.1 of the SAR, the applicant gave special consideration to 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 commoncause situations that could lead to interactions between processes, systems, or buildings.

Chapter 6, Tables 6.1-1 through 6.1-6, of the SAR (AES, 201 Oa) identifies the hazardous

properties of all chemicals used onsite and the inventories and locations of chemicals of

concern (including UF6). Potential interactions involving UF 6 and any reaction products are

identified in Section 6.1.2 of the SAR (AES, 2010a). These chemicals include hydrogen

fluoride (HF) and uranyl fluoride (U0 2F2). This section also 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 Section 6.2.1 of the SAR (AES, 2010a). Section 7.2 of the

SAR (AES, 2010a) describes the Fire Hazards Analysis for those facility areas containing

licensed material. Chapter 3 of the ISA Summary (AES, 2010b) identifies the external hazards.

The applicant identified either a loss of confinement (of UF 6) or a criticality as the hazard of

concern in Section 3.1.4 of the ISA Summary (AES, 2010b). Potential accident sequences that

could result in an UF6 release or criticality of high or intermediate consequence are discussed in

detail in Section 3.7 of the ISA Summary (AES, 2010b).

Table 3.7-1, Accident Sequence and Risk Index, of the ISA Summary (AES, 201 Ob) lists the

potential accident sequences that were identified that could have consequences that exceed the

performance requirements of 10 CFR 70.61. Such sequences could be caused by external

events, facility events external to the process being analyzed, deviations from normal

operations, and failures of IROFS. In this list, the applicant demonstrates with the application of

IROFS that high-consequence accident sequences are highly unlikely and intermediateconsequence accident sequences are unlikely.

In Section 3.1.3 of the SAR (AES, 2010a), the applicant describes the consequence verses

likelihood risk matrix. The risk matrix and computed index values are shown in Table 3.1-6 of

the SAR (AES, 2010a) with the likelihood categories across the top and the consequence

categories along the left side. The risk matrix shows the combinations of likelihood and risk that

are unacceptable. Sequences that fall into these combinations will be mitigated or prevented

with IROFS.

The risk index evaluation process as described in Section 3.1.4 of the SAR 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. For sequences that place the system in a vulnerable state, the duration

of the vulnerable state is considered, and a duration index assigned (Section 3.1.4 of the SAR

[AES, 2010a]). The values of all index numbers in a sequence are added to obtain a total

likelihood index of T (Section 3.1.4 of the SAR [AES, 2010a]). Accident sequences are then

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assigned to one of three likelihood categories, depending on the value of the index in

accordance with Table 3.1-8 of the SAR (AREVA, 2010a). The criteria of Tables 3.1-9 through

3.1-11 of the SAR (Section 3.1.4 of the SAR [AES, 2010a]) 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 and consistent

with the methodology described in Appendix A of Chapter 3 of NUREG 1520 (NRC, 2002). 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.

Consequence Analysis Method

Consequence analysis methods for determining the chemotoxic exposure to HF and U0 2F2 are

discussed in Section 6.3.2 of the SAR (AES, 2010a), The radiological and chemical

consequence severity levels are provided in SAR Table 3.1-3 (AES, 2010a). Information on the

chemical dose limits specific to the EREF is found in Table 3.1-4 of the SAR (AES, 2010a). 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-641 0, "Nuclear Fuel Cycle Facility Accident Analysis Handbook" (NRC, 1998).

The consequences of an inadvertent criticality were conservatively assumed to be high for both

the public and the workers by the applicant because of the potential high radiation exposure to

an individual.

The staff finds that the applicant's method for consequence determination is consistent with the

guidance described in NUREG-1513 (NRC, 2001) and NUREG-1 520 (NRC, 2002a), and is,

therefore, acceptable. Further evaluation of chemical and radiological consequences is found in

Chapter 6 and Appendix B of this SER.

Consequence Categqories

Accident sequences identified by the applicant as a result of the process hazards analysis are

categorized as either as a high-consequence event or an intermediate-consequence event, or

are considered a low consequence event in accordance with the performance criteria of

10 CFR 70.61. Table 3.1-3 of the SAR (AES, 2010a) identifies the consequence severity

categories for workers, offsite public and the environment for radiation and chemical doses as

defined by the applicant. The values proposed by the applicant for both radiological dose and

chemical dose are consistent with the guidance in NUREG-1 520, Appendix A, Table A-5

(NRC, 2002a). All of the identified high-consequence and intermediate-consequence events

are listed in the ISA Summary (AES, 201 Ob) and are comprised of sequences caused by

external events, facility events external to the process being analyzed, deviations from normal

operations, and failures of IROFS.

Safe-by-Design IROFS

The SAR defines a special class of nuclear criticality safety (NCS) controls as safe-by-design

IROFS. As defined in SER Section 5.3.8.1:

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"A safe-by-design IROFS is a passive engineered control which - by its geometry and

configuration alone - will prevent a criticality accident from occurring. This means that all

NCS parameters, except geometry and interaction, are assumed to be in the optimum or

worst case credible condition. By definition, it is considered to be highly unlikely for a safeby-design IROFS to fail in a manner that would cause a criticality accident. The most

significant requirements to qualify as a safe-by-design IROFS are that:

1. The only credible failure mechanism that could result in a criticality would be to

implement an improper design change.

2. Credible process deviations or events do not adversely impact performance of the

safety function.

3. Quality Assurance Level 1 is applied to the feature.

4. No human actions are required for the component to perform its safety function."

The failure of a safe-by-design IROFS due to a loss of configuration control is considered to be

an initiating event with an index of -5 (SER Section 5.3.9). The staff determined that a safe-bydesign IROFS would be considered failed if it were installed, approved for use, and an adequate

margin of subcriticality (i.e., keff < 0.95) would not be maintained for all credible process

deviations or events. It would require multiple human errors in the configuration management

process for such a failure to occur, which is sufficient to accept that this event is highly unlikely.

Safe-by-design IROFS must also have a "significant margin," as defined in the SAR. The staff

determined that EREF's application of significant margin does not provide any additional benefit

to safety, and it is not relied upon to make any conclusions for this review.

The staff concludes that components that meet the safe-by-design criteria will be capable of:

(1) preventing a criticality accident under normal and credible abnormal conditions, (2)

maintaining an adequate margin of subcriticality for safety, and (3) ensuring that a criticality

accident is highly unlikely to occur (SER Section 5.3.9).

3.3.2.2 Definition of Receptors for Consequence Evaluations

The receptors considered for the consequence severity categories as listed in Table 3.1-3 of the

SAR are:

* Workers - individuals inside or outside of facility buildings within the controlled area

boundary

* Offsite Public - individuals at the controlled area boundary nearest to the release point

" Environment - a point at the restricted area boundary nearest to the point of release for

which a 24-hour average concentration is calculated

The boundaries of the controlled and restricted areas are provided in Figure 1.1-3 of the SAR.

For workers and the offsite public, both high and intermediate consequence categories may be

applicable. For the environment, only the intermediate-level consequence category is

applicable.

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3.3.2.3 Likelihood Evaluation Method

Section 3.1.3.2 of the SAR (AES, 2010a) discusses the definitions of "Not Unlikely," "Unlikely,"

and "Highly Unlikely," and SAR Table 3.1-5 (AES, 201 Oa), cross references the three likelihood

categories with a probability of occurrence based on approximate order of magnitude ranges.

The proposed values, ranging from less than 10-5 events per year for "highly unlikely" to more

than 10 -4 events per year for "not unlikely",are consistent with NUREG-1 520, Chapter 3,

Appendix A, Table A-6 (NRC, 2002a). The ISA Summary Tables 3.7-1 through 3.7-4

(AES, 2010b), show how each designated IROFS acts to prevent or mitigate the consequences

of an accident sequence. When multiple IROFS are designated, redundant systems will be

separate and independent from each other, as stated in Section 3.8.1 of the ISA Summary

(AES, 2010b).

The likelihood of failure was qualitatively evaluated for each IROFS, often based on the

operational history of similar facilities. Each sequence was evaluated as "Not Unlikely,"

"Unlikely," or "Highly Unlikely."

3.3.2.4 Chemical Consequences

This section evaluates the proposed chemical quantitative risk levels to protect workers and the

public. Radiological consequence limits are specified in 10 CFR 70.61(b) and (c). However,

chemical consequence limits are not specified in the regulation but are required to be proposed

by the applicant and described in the ISA Summary in accordance with 10 CFR 70.65(b)(7).

The applicant's proposed chemical consequence levels for HF and soluble uranium are

proposed in the SAR (AES, 201 Oa) and the ISA Summary (AES, 2010b), and were evaluated by

the staff, as discussed below.

The proposed chemical consequences were classified as high or intermediate based on the

performance requirements contained in 10 CFR 70.61. Low consequences are not defined in

10 CFR 70.61. However, the applicant defined low consequence accident scenarios as events

that do not exceed the intermediate threshold values, that is, those sequences that do not

require controls or IROFS to meet 10 CFR 70.61. This definition is acceptable to the staff. The

applicant proposed the Acute Exposure Guideline Levels (AEGLs) values, established by the

National Advisory Committee for Acute Guideline Levels for Hazardous Substances, as the

quantitative standard used to assess the consequences to an individual from acute chemical

exposure to licensed material or chemicals produced from licensed materials which are onsite

or expected to be onsite. The staff finds that the use of this standard is consistent with page 3-

23 of NUREG-1 520 (NRC, 2002) which allows the use of values from national and international

accepted standards.

The applicant defined the AEGLs as follows:

AEGL-1 (non-disabling) "The airborne concentration of a substance above which it is

predicted that the general population, including susceptible

individuals, could experience notable discomfort, irritation or

certain asymptomatic effects. However, the effects are not

disabling and are transient and reversible upon cessation of

exposure."

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AEGL-2 (disabling) "The airborne concentration of a substance above which it is

predicted that the general population, including susceptible

individuals, could experience irreversible or other serious, longlasting adverse health effects, or an impaired ability to escape."

AEGL-3 (lethality) "The airborne concentration of a substance above which it is

predicted that the general population, including susceptible

individuals, could experience life-threatening health effects or

death."

The comparison of the AEGLs with the calculated airborne concentrations allows the applicant

to classify a high, intermediate, or low chemical consequence level. In the case of exposures to

soluble forms of uranium (e.g., UF6 and U0 2F2), the applicant uses intakes of uranium

expressed in milligrams (mg) of material. As part of the chemical consequence methodology as

presented in Tables 3.1.3 and 3.1.4 of the SAR (AES, 201 Oa), the applicant states that 21 mg of

soluble uranium (U) intake 4 would be the high consequence limit for an individual at the

controlled area boundary instead of 30 mg of soluble uranium intake as required in

10 CFR 70.61(b)(3). The high-consequence limit taken by the applicant is more conservative

than the value in the 10 CFR 70.61(b)(3); therefore, it is acceptable. The intermediateconsequence limit for an individual at the controlled area boundary is 4.06 mg of soluble U

intake. The applicant indicates that the soluble U intake limits do not apply to the worker

because the worker is more conservatively protected by the UF6 AEGL limits. The NRC staff

reviewed the calculation to support this statement and finds it acceptable. The soluble U intake

limits are taken from NUREG-1391 (NRC, 1991). The AEGL and the soluble U intake limits are

shown in Table 3.3.-l.

Table 3.3 - 1 Consequence Level as Related to AEGL Values and Soluble Uranium Intake

Consequence Worker Individual at the

Level Controlled Area Boundary

High [HF] > AEGL-3 [HF] > AEGL-2

[UF 6] > AEGL-3 Soluble U intake > 21 mg

Intermediate AEGL-2 < [HF] < AEGL-3 AEGL-1 < [HF] < AEGL-2

AEGL-2 < [UF 6] < AEGL-3 4.06 mg < Soluble U intake < 21 mg

Low [HF]• AEGL-2 [HF]5 <AEGL-1

[UF 6] < AEGL-2 Soluble U intake < 4.06 mg

Note: [HF] and [UF 6] are the concentrations of HF and UF 6 as given by the AEGL value

(i.e., mg/m 3

) (not molar concentration (i.e., mol/L)), respectively.

The use of the AEGL standard allows the applicant to vary the exposure duration. The chemical

consequence levels proposed by the applicant are shown in Table 3-2, for HF, and Table 3-3,

for UF6.

4The applicant actually states a 21-mg body burden (i.e., the amount of uranium that stays in the body),

but this limit does not take credit for the loss of uranium through exhalation. Therefore, it is appropriate to

equate the 21-mg body burden to 21 mg of soluble U intake.

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Table 3.3-2 Chemical Consequence Levels Proposed by the Applicant for Hydrogen Fluoride

Exposure AEGL-1 AEGL-2 / AEGL-31

Duration (mg/m3) (mg/m3) (mg/m3)

10 min 0.8 78 139

30 min 0.8 28 51

1 hour 0.8 20 36

4 hours 0.8 9.8 18

8 hours 0.8 7 12

Table 3.3-3 Chemical Consequence Levels Proposed by the Applicant for Uranium Hexafluoride

Exposure AEGL-1 AEGL-2 AEGL-3

Duration (mg/m3) (mg/m3) (mg/m3)

10 min 3.6 28 216

30 min 3.6 19 72

1 hour 3.6 9.6 36

4 hours - 2.4 9

8 hours - 1.2 4.5

Consistent with Section 3.4.3.2(7) of NUREG-1 520 (NRC, 2002), the applicant's quantitative

standards for chemical consequences meet the following criteria and, therefore, are acceptable:

* There are unambiguous, quantitative standards for each of the applicable hazardous

chemicals that meet the criteria of 10 CFR 70.65(b)(7) onsite;

* The quantitative standard of 10 CFR 70.61(b)(4)(i) addresses exposures that could

endanger the life of a worker;

" The quantitative standards for 10 CFR 70.61(b)(4)(ii) and 10 CFR 70.61(c)(4)(i) will correctly

categorize all exposures that could lead to health effects on individuals that could be long

lasting; and

* The quantitative standard for 10 CFR 70.61(c)(4)(ii) will correctly categorize all exposures

that could cause individual mild transient health effects.

3.3.2.5 Radiological Consequences

Radiological consequences limits are specified in 10 CFR 70.61(b) and (c), and are presented

along with chemical consequences in Table 3.1-3 of the SAR (AES, 2010a). As presented

below, the applicants radiological limits are the same as those specified in 10 CFR 70.61 and

are acceptable.

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Table 3.3-4 Radiological Consequence Levels Proposed by the Applicant

Consequence level Worker Individual at the

controlled area

boundary

High > 100 rem (1 Sv) > 25 rem (0.25 Sv)

Intermediate > 25 rem (.25 Sv) > 5 rem (.05 Sv)

Low < 25 rem (.25 Sv) < 5 rem (.05 Sv)

3.3.2.6 Environmental Consequences

The performance requirements of 10 CFR 70.61 (c)(3) specify that the environmental

consequences of each credible, intermediate-consequence event must be limited so that the

24-hour averaged release concentration of U outside the restricted area is less than 5,000 times

the value specified in 10 CFR 20, Appendix B, Table 2. Appendix E of the ISA Summary

(AES, 2010b) presents the methodology that was used to evaluate the consequences of

hypothesized accidental releases of UF6, which is based on NUREG/CR-6410 (Nuclear Fuel

Cycle Facility Accident Analysis Handbook) and NUREG/CR-5659 (Control Room Habitability

System Review Models). The NRC staff reviewed the applicant's methodology for determining

compliance with the environmental consequences requirement and finds it to be acceptable. As

discussed in Section 9.3.3 of this SER, the NRC staff determined that environmental

consequences may occur only if uncontrolled, intermediate or high consequences to workers

are also present. The NRC staff did not identify any accident sequence that would fail to meet

the environmental performance requirements of 10 CFR 70.61(c)(3). The NRC staff also made

confirmatory calculations for a subset of potential accident scenarios as discussed in the nonpublic Appendix B of this SER.

3.3.2.7 ISA Methodology Conclusion

Based on the above information, the staff concludes that the applicant has used a methodology

consistent with NUREG-1 513 (NRC, 2001) to identify hazards related to this type of facility and

credible events that could exceed the performance requirements of 10 CFR 70.61. The

applicant has also established definitions of likelihood consistent with Section 3.4.3.2 of

NUREG-1 520 and applied those definitions to demonstrate that intermediate-consequence

events are unlikely and high-consequence events are highly unlikely.

3.4 Evaluation

The staff finds that the applicant's maintenance of process safety information is in accordance

with the guidance of NUREG-1520 (NRC, 2002) (described in Section 3.3.1.1 of this SER).

The staff finds that the applicant's commitment to conduct and maintain an ISA is in accordance

with the requirements of 10 CFR 70.62(c)(1) and the guidance in NUREG-1520 (NRC, 2002).

(described in Sections 3.3.1.1 and 3.3.1.2 of this SER).

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The staff considers the ISA methodology to be complete by its use of the appropriate accident

identification methodology from NUREG-1 513 (NRC, 2001). The staff considers the

consequence determinations to be acceptable and in accordance with the guidance in

NUREG/CR-6410 (NRC, 1998). The staff has also evaluated the consequence determination

methodology, and the staff's evaluation is presented in Appendix B of this SER. The staff

considers the likelihoods to have been derived using acceptable methods and to comply with

acceptable definitions of "not unlikely," "unlikely," and "highly unlikely" as evaluated in

Appendix A of this SER. The staff concludes that these descriptions (described in Section 3.3.2

of this SER) conform with the guidance provided in NUREG-1 520 (NRC, 2002), and meet the

requirements of 10 CFR 70.65(b)(4).

Area boundaries, including the controlled area boundary and the locations of restricted areas,

are adequately described for the purpose of determining consequences (described Section

3.3.2.2 of this SER) and meet the requirements of 10 CFR 70.65(b)(2).

The likelihood evaluation, including the definition of "not credible," conforms to the guidance

provided in NUREG-1 520 (NRC, 2002) and is acceptable (described in Section 3.3.2.3 of this

SER).

The determination of chemical consequences (described in Section 3.3.2.4 of this SER)

conforms to the guidance provided in NUREG-1520 (NRC, 2002) and is acceptable.

3.5 References

(AES, 2010a) AREVA Enrichment Services LLC, "Eagle Rock Enrichment Facility, Safety

Analysis Report," 2010.

(AES, 2010b) AREVA Enrichment Services LLC, "Integrated Safety Analysis Summary for the

Eagle Rock Enrichment Facility," 2010.

(ANSI, 1974) American National Standards Institute/Institute of Electrical and Electronics

Engineers, "IEEE Standard for Type Test of Class 1 E Electrical Cables, Field Splices, and

Connections for Nuclear Power Generating Stations," ANSI/IEEE-383-1974, 1974.

(NRC, 2002) U.S. Nuclear Regulatory Commission, NUREG-1 520, "Standard Review Plan for

the Review of a License Application for a Fuel Cycle Facility," 2002.

(NRC, 2001) U.S. Nuclear Regulatory Commission, NUREG-1513, "Integrated Safety Analysis

Guidance Document," 2001.

(NRC, 1998) U.S. Nuclear Regulatory Commission, NUREG/CR-6410, "Nuclear Fuel Cycle

Accident Analysis Handbook," 1998.

(NRC, 1991) U.S. Nuclear Regulatory Commission, NUREG-1 391, "Chemical Toxicity of

Uranium Hexafluoride Compared to Acute Effects of Radiation," February 1991.

(NRC, 1990) U.S. Nuclear Regulatory Commission and Science Applications International

Corporation, "Control Room Habitability System Review Models," December 1990.

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CHAPTER 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 Title 10 of the Code of Federal Regulations (10 CFR)

Parts 19, 20, 30, 40, and 70. Public and environmental protection is discussed in Chapter 9 of

this Safety Evaluation Report (SER).

4.1 Regulatory Requirements

* Regulations applicable to the establishment of an RP program are presented in Part 20,

Subpart B, "Radiation Protection Programs."

" Regulations applicable to the As Low As Is Reasonably Achievable (ALARA) program

are presented in 10 CFR 20.1101, "Radiation Protection Programs."

* Regulations applicable to the organization and qualifications of the radiological

protection staff are presented in 10 CFR 30.33, "General Requirements for Issuance of

Specific Licenses;" 10 CFR 40.32, "General Requirements for Issuance of Specific

Licenses;" and 10 CFR 70.22, "Contents of Applications."

* 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."

* 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"

* Regulations applicable to the ventilation and respiratory protection programs are

presented in 10 CFR Part 20, Subpart H, "Respiratory Protection and Controls to

Restrict Internal Exposure in Restricted Areas."

* Regulations in 10 CFR Part 20 are applicable to radiation surveys and monitoring

programs:

1. 10 CFR Part 20, Subpart F "Surveys and Monitoring"

2. 10 CFR Part 20, Subpart C "Occupational Dose Limits"

3. 10 CFR Part 20, Subpart L "Records"

4. 10 CFR Part 20, Subpart M "Reports"

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* The following NRC regulations are applicable to the additional program requirements:

1. 10 CFR Part 20, Subpart L "Records"

2. 10 CFR Part 20, Subpart M"Reports"

3. 10 CFR 70.61 "Performance Requirements"

4. 10 CFR 70.74 "Additional Reporting Requirements"

4.2 Regulatory Guidance and Acceptance Criteria

The acceptance criteria for the NRC staff'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 RP Program Implementation

The staff reviewed the applicant's RP program implementation against the acceptance criteria in

Section 4.4.1.3 of NUREG-1520 (NRC, 2002). The following sections identify each acceptance

criteria from NUREG-1 520 and discuss the staff's analysis as to whether the information

provided by the applicant in the Safety Analysis Report (SAR) (AES, 2009a) meets the criteria.

1. "[D]esign and implement a RP program that meets the regulatory requirements of 10 CFR

Part 20, Subpart B"

In Section 4.1 of the Eagle Rock Enrichment Facility (EREF) SAR (AES, 2009a), the

applicant ensures that the RP program would meet the 10 CFR 20, Subpart B requirements

by committing that the program would be consistent with the guidance provided in

Regulatory Guide (RG) 8.2, "Guide for Administrative Practice in Radiation Monitoring (NRC,

1973)." The facility would develop, document, and implement its RP program

commensurate with the risks posed by a uranium enrichment operation. The facility would

use, to the extent practicable, procedures and engineering controls based upon sound RP

principles to achieve occupational doses and doses to members of the public that are

ALARA. Constraints on atmospheric releases would be established such that no member of

the public would be expected to receive greater than 10 millirem per year (mrem/yr) total

effective dose equivalent (TEDE) from exposure to the release. Also, the RP program

content and implementation would be reviewed at least annually.

2. "[O]utline the RP program structure and define the responsibilities of key program

personnel"

The applicant outlined the RP program structure and defined the responsibilities of key

program personnel in Section 4.1.1 of the EREF SAR (AES, 2009a). The AREVA

Enrichment Services's (AES's) President would have overall responsibility for the operation

of the EREF, including RP. The Plant Manager would report to the AES President and

would be responsible for the protection of all persons against radiation exposure resulting

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from facility operations and materials and for compliance with applicable NRC regulations

and the facility license. The Environmental, Health, Safety, and Licensing (EHS&L)

Manager would report to the Plant Manager and would have overall responsibility for

development and implementation of the RP program. The Radiation Protection/Chemistry

Manager would report to the ESH&L Manager; would be responsible for implementing the

RP program; and, in matters involving RP, would have direct access to the Plant Manager;

and, would establish the RP program independent of operations.

The Operations Manager would report to the Plant Manager and would have the

responsibility for the safe day-to-day operation of the facility including operating in

accordance with procedures that incorporate RP practices. Facility personnel would also be

required to work safely and follow rules, regulations, and procedures established for their

protection and protection of the public.

3. "[S]taff the RP program with suitably trained people, provide sufficient resources, and

implement the program"

Staffing of the RP program is addressed in Sections 4.1.1.4 and 4.1.2 of the EREF SAR

(AES, 2009a). The Radiation Protection/Chemistry Manager would be responsible to

sufficiently staff the facility with suitably trained people to implement an effective RP

program. The staff would be trained and qualified consistent with the guidance provided in

the American National Standards Institute (ANSI)/American Nuclear Society (ANS) Standard

N3.1-1993, "Selection, Qualification and Training of Personnel for Nuclear Power Plants"

(ANSI/ANS, 1993).

4. "[C]ommit to the independence of the radiation protection function from the facility's

operations"

Section 4.1.3 of the EREF SAR (AES, 2009a) addresses the independence of the RP

program. The RP program will be independent of operations because the Radiation

Protection/Chemistry Manager has direct access to the Plant Manager, and the RP staff

reports to the Radiation Protection/Chemistry Manager.

5. "[R]eview, at least annually, the content and implementation of the radiation protection

program as required by 10 CFR 20.1101(c)"

As previously stated, the applicant, in Section 4.1 of the SAR (AES, 2009a), commits to

review, at least annually, the content and implementation of the RP program.

Based on the analysis as summarized above, the staff finds that the commitments in the EREF

SAR (AES, 2009a) satisfactorily address the application acceptance criteria in Section 4.4.1.3 of

NUREG-1520 (NRC, 2002).

4.3.2 ALARA Program

The staff reviewed the applicant's ALARA program commitments against the acceptance criteria

in NUREG-1 520 Section 4.4.2.3 (NRC, 2002). The following sections identify each acceptance

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criteria from NUREG-1520 and discuss the staff's analysis as to whether the information

provided by the applicant in the SAR (AES, 2009a) meets the criteria.

1. "[E]stablish a comprehensive, effective, and written ALARA program"

Section 4.2 of the EREF SAR addresses the proposed ALARA program. The applicant

would implement an ALARA program with the objective to make every reasonable effort to

maintain exposures to radiation as far below the regulatory dose limits as is practical. The

program would be written and policies documented to ensure the ALARA goal is met. The

design and implementation of the ALARA program would be consistent with RGs: 8.2,

"Guide for Administrative Practice in Radiation Monitoring" (NRC, 1973a); 8.13 "Instruction

Concerning Prenatal Radiation Exposure," (NRC, 1999); 8.29, "Instruction Concerning Risks

from Occupational Radiation Exposure," (NRC, 1996); and 8.37 "ALARA Levels for Effluents

from Materials Facilities," (NRC, 1993). The operation of the facility would be consistent

with Regulatory Guide 8.10, "Operating Philosophy for Maintaining Occupational Radiation

Exposures as Low as is Reasonably Achievable," (NRC, 1977). Facility procedures would

incorporate the ALARA philosophy into the routine facility operations. Also, the Radiation

Protection/Chemical Manager would prepare an annual ALARA program evaluation report.

2. "[P]repare policies and procedures to ensure occupational exposures are maintained

ALARA, and that such exposures are consistent with the requirements of 10 CFR 20.1101"

As described in Section 4.2 of the EREF SAR (AES, 2009a), the applicant would maintain

exposures to members of the public such that they are not expected to exceed the limits in

10 CFR 20.1101(d) which established a 10 mrem/yr dose constraint to the public from

atmospheric releases.

3. "[O]utline specific ALARA program goals, establish an ALARA program organization and

structure, and have written procedures for its implementation in the facility design and

operations"

Section 4.2 of the EREF SAR (AES, 2009a) commits to specific goals for the ALARA

program such as maintaining occupational exposures (both individual and collective

exposures), as well as environmental releases, as far below regulatory limits as is

reasonably achievable. The Radiation Protection/Chemistry Manager would be responsible

for implementing the ALARA program and ensuring adequate resources are available to

make the program effective. As previously noted, the applicant has committed that facility

procedures would incorporate the ALARA philosophy into the routine facility operations.

The size and number of areas at the facility with higher dose rates would be minimized,

consistent with accessibility for performing necessary services, while areas where facility

personnel spend significant amounts of time would be designed to maintain the lowest dose

rates reasonably achievable. Radiological zones, such as "Radiation Area," "Airborne

Radioactivity Area," and "Contaminated Area," will be established to minimize the spread of

contamination and reduce unnecessary radiation exposure to personnel.

4. "[E]stablish an ALARA Committee, or equivalent organization, with sufficient staff,

resources, and clear responsibilities to ensure that the occupational radiation exposure dose

limits specified in 10 CFR Part 20 are not exceeded under normal operations"

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Sections 2.2.3 and 4.2.1 of the EREF SAR (AES, 2009a) include commitments to establish

a Safety Review Committee (SRC) which would meet at least quarterly and within 30 days

of any NRC-reportable incident and would fulfill the duties of the ALARA Committee. The

SRC would be composed of at least five members including the Chairman. The five

members would include experts on operations and all safety disciplines (criticality,

radiological, chemical, and industrial). The Chairman, members, and alternate members, of

the SRC would be formally appointed by the AES President. The SRC would conduct at

least one facility audit per year that would review and audit the following subject areas as

pertain to the facility.

" Radiation protection

" Nuclear criticality safety

" Hazardous chemical safety

" Industrial safety including fire protection

" Environmental protection

" ALARA policy implementation

" Changes in facility design or operations.

As part of its duties, the SRC would review the effectiveness of the ALARA program and

determine if exposures, releases, and contamination levels are in accordance with the

ALARA concept. It also would evaluate the results of assessments made by the RP

organization, reports of facility radiation levels, and employee exposures for identified

categories of workers and types of operations. The committee is responsible for ensuring

that occupational radiation exposures do not exceed the dose limits in 10 CFR 20.1201

under normal operations.

5. "[U]se the ALARA program as a mechanism to facilitate interaction between RP and

operations personnel"

The applicant would assure interaction between RP and operations personnel because both

operations and radiation safety management will be participating on the SRC and making

joint recommendations for improvements. Recommendations would be tracked to

completion using the facility's Corrective Action Program.

6. "[R]egularly review and revise, when appropriate, the ALARA program goals and objectives

and to incorporate, when appropriate, new approaches, technologies, operating procedures

or changes that could reduce potential radiation exposures at a reasonable cost"

As discussed in Section 4.2.1 of the EREF SAR, the SRC would review the effectiveness of

the ALARA program and determine if exposures, releases and contamination levels are in

accordance with the ALARA concept. The SRC would also periodically review the goals

and objectives of the ALARA program. These goals and objectives would be revised to

incorporate, as appropriate, new technologies or approaches and operating procedures or

changes that could cost-effectively reduce potential radiation exposures.

Based on the analysis as summarized above, the staff finds that the commitments in the EREF

SAR (AES, 2009a) satisfactorily address the application acceptance criteria in Section 4.4.2.3 of

NUREG-1520 (NRC, 2002).

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4.3.3 Organization and Personnel Qualifications

The staff reviewed the applicant's organization and personnel qualifications against the

acceptance criteria in NUREG-1 520, Section 4.4.3.3 (NRC, 2002). The following sections

identify each acceptance criteria from NUREG-1 520 and discuss the staffs analysis as to

whether the information provided by the applicant in the SAR (AES, 2009a) meets the criteria.

1. "[A]ppoint suitably trained RP personnel and identify their authority and responsibilities"

The applicant described the qualifications, authority, and responsibilities of RP personnel in

Sections 2.2.1, 4.1.1, and 4.3 of the EREF SAR (AES, 2009a). The Radiation

Protection/Chemistry Manager would 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

Radiation Protection/Chemistry Manager would have direct access to the Plant Manager

regarding all matters involving RP, would be skilled in the interpretation of RP data and

regulations, would be familiar with the operation of the facility and RP concerns of the site,

and would be a resource in radiation safety management decisions.

The Radiation Protection/Chemistry Manager would have, as a minimum, a bachelor's

degree (or equivalent) in an engineering or scientific field and 4 years of responsible nuclear

experience associated with the implementation of an RP program. At least one member of

the RP staff would have a minimum of 2 years experience at a facility that processes

uranium, including uranium in soluble form. The RP staff would be trained and qualified

consistent with guidance provided in ANSI/ANS Standard 3.1 "Selection, Qualification, and

Training of Personnel for Nuclear Power Plants," (ANSI/ANS, 1993).

2. "[E]stablish clear organizational relationships among the individual positions responsible for

the radiation protection program and other line managers"

The organizational relationships between the various managers are described in Section

2.2.1 of the SAR (AES, 2009a). Of these relationships, the Radiation Protection/Chemistry

Manager reports up through the ESH&L Manager to the Plant Manager. The operations line

managers organizational relationships also lead to the Plant Manager through an unrelated

managerial path.

3. "[A]ppoint a suitably trained RP program director (typically referred to as the radiation safety

officer) who has direct access to the facility manager, who is skilled in the interpretation of

data and regulations pertinent to radiation protection, who is familiar with the operation of

the facility and RP concerns of the site, who is used as a resource in radiation safety

management decisions, and who will be responsible for establishing and implementing the

RP program"

In Section 4.3 of the EREF SAR (AES, 2009a), the applicant designated the Radiation

Protection/Chemistry Manager as the individual who would have responsibility for

establishing and implementing the RP Program. This individual would be skilled in the

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interpretation of RP data and regulations, familiar with the operation of the facility and

relevant RP concerns, and be a resource for radiation safety management decisions.

4. "[A]ssign responsibility to the RP program staff for implementation of the RP program

functions"

The applicant would assign responsibility for implementing the RP program functions to the

RP staff as stated in Section 4.3 of the EREF SAR (AES, 2009a).

5. "[D]escribe the minimum training requirements and qualifications for the RP staff"

As described in Sections 2.2.4, 4.1.2 and 4.3 of the EREF SAR (AES, 2009a), the Radiation

Protection/Chemistry Manager's qualifications would minimally include a bachelor's degree

and at least four years of nuclear experience. At least one member of the staff would have

at least two years of experience in a uranium processing facility. Other members of the staff

would be trained and qualified consistent with the guidance provided in ANSI standard 3.1,

"Selection, Qualification, and Training of Personnel for Nuclear Power Plants," (ANSI 1993).

Based on the analysis as summarized above, the staff finds that the information provided in the

EREF SAR (AES,2009a) satisfactorily address the application acceptance criteria in Section

4.4.3.3 of NUREG-1 520 (NRC, 2002).

4.3.4 Written Procedures

The staff reviewed the applicant's written procedure commitments against the acceptance

criteria in NUREG-1 520, Section 4.4.4.3 (NRC, 2002). The following sections identify each

acceptance criteria from NUREG-1 520 and discuss the staff's analysis as to whether the

information provided by the applicant in the SAR (AES, 2009a) meets the criteria.

1. "[P]repare written, approved RP procedures to carry out activities related to the RP program"

The applicant has committed to using written procedures to conduct operations involving

licensed materials in Section 4.4 of the EREF SAR (AES, 2009a). RP procedures would be

prepared, reviewed, and approved to carry out activities related to the RP program. They

would also be 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).

2. "[S]pecify how the RP procedures will be prepared, authorized, approved, and distributed"

As discussed in Section 4.3 of the EREF SAR (AES, 2009a), the applicant committed that

RP procedures would be prepared by qualified personnel, reviewed by members of the

facility staff (i.e., personnel with enrichment plant operating experience and other staff

members, as appropriate), and approved by the Radiation Protection/Chemistry Manager

(or designee) and the Plant Manager (or designee).

3. "[S]pecify written, approved RWPs for activities involving licensed material that are not

covered by written RP procedures"

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The applicant discussed RWPs in Section 4.4.1 of the EREF SAR (AES, 2009a). The

applicant commits to perform all work in Restricted Areas in accordance with a RWP. The

applicant would also issue RWPs for activities involving licensed materials not covered by

operating procedures, where radioactivity levels are likely to exceed airborne radioactivity

limits, or whenever deemed as necessary by the Radiation Protection/Chemistry Manager.

RWPs would provide a description of the work or authorized activities; summary results of

dose rate, contamination, and airborne radioactivity surveys; and precautions, which may

include personnel protective equipment, stay-times or dose limits, recordkeeping

requirements, and coverage by an RP technician. RWPs would require approval by the

Radiation Protection/Chemistry Manager or qualified designee, and would have a

predetermined period of validity with a specified expiration or termination time.

Based on the analysis as summarized above, the staff finds that the commitments in the EREF

SAR (AES, 2009a) satisfactorily address the application acceptance criteria in Section 4.4.4.3 of

NUREG-1520 (NRC, 2002).

4.3.5 Training

The staff reviewed the applicant's training commitments against the acceptance criteria in

NUREG-1520, Section 4.4.5.3 (NRC, 2002). The following sections identify each acceptance

criteria from NUREG-1520 and discuss the staff's analysis as to whether the information

provided by the applicant in the SAR (AES, 2009a) meets the criteria.

1. "[D]esign and implement an employee RP training program that complies with the

requirements of 10 CFR Parts 19 and 20"

The applicant has incorporated the provisions of 10 CFR 19.12 into the RP training program,

as discussed in Section 4.5, of the EREF SAR (AES, 2009a). 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 milliSievert (mSv) (100 mrem)

per year. The applicant's RP training program would ensure that workers likely to receive

such exposures are:

* Kept informed of the storage, transfer, or use of radioactive material

* Instructed in the health protection problems associated with exposure to radiation and

radioactive material, in precautions or procedures to minimize exposure, and in the

purposes and functions of protective devices employed

" Required to observe, to the extent within the worker's control, the applicable provisions

of the NRC regulations and licenses for the protection of personnel from exposure to

radiation and radioactive material

* Instructed of their responsibility to report promptly to the facility management, any

condition which may cause a violation of NRC regulations and licenses or unnecessary

exposure to radiation and radioactive material

* Instructed in the appropriate response to warnings made in the event of any unusual

occurrence or malfunction that may involve exposure to radiation and radioactive

material

* Advised of the various notifications and reports to individuals that a worker may request

in accordance with 10 CFR 19.13

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2. "[P]rovide training, to all personnel and visitors entering restricted areas, that is

commensurate with the health risk to which they may be exposed, or provide trained

escorts"

The applicant has committed in Section 4.5 of the EREF SAR (AES, 2009a) to an RP

training program that will be designed and implemented to provide training to all personnel

and visitors, unless provided with trained escorts, who enter Restricted Areas or Controlled

Areas, and which would be commensurate with the radiological hazard to which they may

be exposed. As per Section 11.3.3.1.1 of the SAR (AES, 2009a), "personnel access

procedures ensure the completion of formal nuclear safety training prior to permitting

unescorted access into the Controlled Access Area."

3. "[P]rovide a level of training based on the potential radiological health risks associated with

that employee's work responsibilities"

In Section 4.5 of the EREF SAR (AES, 2009a), the applicant committed to provide a level of

training based on the potential radiological health risks associated with the individual's work

responsibilities.

4. "[I]ncorporate, in the RP training program, the provisions of 10 CFR 19.12 and topics such

as: correct handling of radioactive materials; minimization of exposures to radiation and/or

radioactive materials; access and egress controls and escort procedures; radiation safety

principles, policies, and procedures; monitoring for internal and external exposures;

monitoring instruments; contamination control, including protective clothing and equipment;

ALARA and exposure limits; radiation hazards and health risks; and, emergency response"

As discussed in Section 11.3.3.1.1 of the EREF SAR (AES, 2009a), topics covered in the

applicant's training program would include:

* Notices, reports and instructions to workers

* Practices designed to keep radiation exposures ALARA

* Methods of controlling radiation exposures

" Contamination control methods (including decontamination)

" Use of monitoring equipment

* Emergency procedures and actions

* Nature and sources of radiation

" Safe use of chemicals

" Biological effects of radiation

* Use of personnel monitoring devices

* Principles of nuclear criticality safety

" Risk to pregnant females

* Radiation protection practices

* Protective clothing

* Respiratory protection

* Personnel surveys

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5. "[R]eview the RP training program at least every 3 years and to conduct refresher training at

least every 3 years to address changes in policies, procedures, requirements, and the

facility ISA"

As discussed in Section 4.5.1 of the EREF SAR (AES, 2009a), individuals requiring

unescorted access to a Restricted Area would receive annual retraining. Retraining for

individuals would be scheduled and reported by means of a computerized tracking system.

Contents of the formal training program would be reviewed and updated at least annually by

the EHS&L Manager or the Radiation Protection/Chemistry Manager to ensure the program

is current and up to date. In Sections 4.5 and 11.3.3.1.1 of the EREF SAR (AES, 2009a),

the applicant more conservatively commits to evaluate the RP sections of the training

program at least annually.

6. "[E]valuate the effectiveness and adequacy of the training program's curriculum and

instructors"

The applicant would evaluate the effectiveness and adequacy of the training program

through initial examination, audits, and assessments of operations and maintenance

personnel responsible for following the requirements of criticality safety and RP. This is

discussed in Section 11.3.3.1.1 of the EREF SAR (AES, 2009a).

Based on the analysis as summarized above, the staff finds that the commitments in the EREF

SAR (AES, 2009a) satisfactorily address the application acceptance criteria in Section 4.4.5.3 of

NUREG-1520 (NRC, 2002).

4.3.6 Ventilation and Respiratory Protection Programs

The staff reviewed the applicant's ventilation and respiratory protection program commitments

against the acceptance criteria in NUREG-1 520, Section 4.4.6.3 (NRC, 2002). The following

sections identify each acceptance criteria from NUREG-1520 and discuss the staffs analysis as

to whether the information provided by the applicant in the SAR (AES, 2009a) meets the criteria.

1. "[I]nstall appropriately sized ventilation and containment systems in areas of the facility

identified in the ISA Summary as having potential airborne concentrations of radionuclides

that could exceed the occupational, derived air concentration values specified in 10 CFR

Part 20, Appendix B, during normal operations"

The applicant discussed ventilation commitments in Section 4.6.1 of the EREF SAR

(AES, 2009a). The confinement of uranium and the attenuation of its associated radiation

are considered a design requirement for the facility. The internal radiation exposure of

workers would be controlled primarily by the containment of uranium hexafluoride (UF6)

within process equipment. The entire UF6 enrichment process, except for liquid sampling,

would be operated under a partial vacuum so that leaks are into the system and not into

work areas. This indicates that the entire process will be fully contained such that little

airborne radioactivity would be present.

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The applicant stated that the design of the ventilation would be consistent with the guidance

contained in the documents ANSI N510-1989, "Testing of Nuclear Air Cleaning Systems"

(ANSI, 1989) and "DOE Nuclear Air Cleaning Handbook" (DOE, 2003). Ventilation systems

serving potentially contaminated areas include design features that provide for confinement

of radiological contamination. These systems exhaust 100% of the air handled to the

environment through exhaust vents after being filtered to remove radioactive particulates.

2. "[D]escribe management measures, including preventive and corrective maintenance and

performance testing, to ensure that the ventilation and containment systems designated as

IROFS operate when required, and are within their design specifications"

The applicant's ISA Summary (AES, 2009b) Tables 3.8-1 and 3.8-2 list all items relied on for

safety (IROFS) identified by the facility. Staff observed no IROFS identified as resulting

from potential radiological exposure consequences. This is not unexpected as the licensed

material is natural and low enriched uranium compounds and, typically, the consequences

resulting from the toxicity of reaction products outweigh the radiological exposure

consequences for a release of materials. As such, this specific criterion is not applicable to

this evaluation. Discussion of chemical and criticality IROFS and their management

measures is presented in Chapters 3, 5, and 6 and in Appendices A, B, and G of this SER.

3. "[D]escribe the design criteria for the ventilation and containment systems, including

minimum flow velocity at openings in these systems, maximum differential pressure across

filters and types of filters to be used"

The applicant addressed the design of the facility ventilation in Sections 4.6, 4.6.1, and 4.8

of the EREF SAR (AES, 2009a). The applicant committed for ventilation design to be

consistent with the "DOE Nuclear Air Cleaning Handbook" (DOE, 2003) and ANSI 510-1989,

"Testing of Nuclear Air Cleaning Systems" (ANSI, 1989b).

All effluents released from potentially contaminated areas would be filtered to remove

radioactive particulates before it is released. The ventilation systems would be designed to

maintain the potentially contaminated areas at a slightly negative pressure relative to the

uncontaminated areas. This ensures that the airflow direction is from areas of little or no

contamination to areas of higher contamination. The systems would operate slightly below

atmospheric pressure to remove potentially hazardous vapors and particulates from

confined areas of the plant. The systems would contain high efficiency particulate air

(HEPA) and carbon adsorption filters to remove radioactive materials from the gas stream

prior to release from the plant. Effluent air streams would be continuously monitored for

alpha activity and hydrogen fluoride (HF) concentrations and automatically alarm if

concentrations above the set point are detected. Differential pressure across high efficiency

particulate air filters in potentially contaminated exhaust systems would be monitored

monthly or automatically. Automatic monitors would have alarm features. Filters would be

replaced when differential pressure exceeds the manufacturers' ratings or if the filters fail to

function properly.

Gloveboxes would be designed to maintain a negative differential pressure of about

0.623 millibar (mbar) (0.25 inch [in] water [H201). This differential pressure would be

maintained anytime that the glovebox is in use. If the differential pressure is lost, use of the

glovebox would be suspended until the required differential pressure is restored.

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Air flow rates at exhausted enclosures and close-capture points, when in use, would be

adequate to preclude escape of airborne uranium and minimize the potential for intake by

workers. Air flow rates would be checked monthly when in use and after modification of any

hood, exhausted enclosure, close-capture point equipment or ventilation system serving

these barriers.

4. "[D]escribe the frequency and types of tests to measure ventilation and containment

systems performance, the acceptance criteria, and actions to be taken when the acceptance

criteria are not satisfied"

As discussed in Section 4.6.1 of the EREF SAR (AES, 2009a), filter inspection, testing,

maintenance and change out criteria would be specified in written procedures approved by

the Operations Manager, or a designated alternate. Change out frequency would be based

on considerations of filter loading, operating experience, differential pressure data and any

UF6 releases indicated by HF alarms.

Gloveboxes would be designed to maintain a negative differential pressure of about

0.623 mbar (0.25 in H20). This differential pressure would be maintained anytime that the

glovebox is in use. If the differential pressure is lost, use of the glovebox would be

suspended until the required differential pressure is restored.

5. "[E]stablish a respiratory protection program that meets the requirements of 10 CFR Part 20,

Subpart H"

Section 4.6.2 of the EREF SAR (AES, 2009a) discusses the respiratory protection program.

In that section, the applicant stated that, if the decision is made to permit the use of

respiratory protection equipment to limit the intake of radioactive material, only National

Institute of Occupational Safety and Health (NIOSH) certified equipment would be used.

The respiratory protection program would meet the requirements of 10 CFR 20, Subpart H,

"Respiratory Protection and Controls to Restrict Internal Exposure in Restricted Areas."

In Section 4.6 of the SAR (AES, 2009a), the applicant states the design of the respiratory

protection program would be consistent with RG 8.15, "Acceptable Programs for Respiratory

Protection," (NRC 1999), and ANSI Z88.2-1992, "Practices for Respiratory Protection,"

(ANSI, 1992).

6. "[P]repare written procedures for the selection, fitting, issuance, maintenance, maintenance,

testing, training of personnel, monitoring, and recordkeeping for individual respiratory

protection equipment, and for specifying when such equipment is to be used"

In Section 4.6.2 of the EREF SAR (AES, 2009a), the applicant committed to develop written

procedures for the respiratory protection program to address the following subjects:

* Monitoring, including air sampling and bioassays;

" Supervision and training of respirator users;

* Fit testing;

" Respirator selection;

* Breathing air quality;

* Inventory and control;

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" Storage, issuance, maintenance, repair, testing, and quality assurance of respiratory

protection equipment;

* Record keeping; and,

* Limitations on periods of respirator use and relief from respirator use.

7. "[Rlevise the written procedures for use of individual respiratory protection equipment as

applicable, when processing, facility, or equipment changes are made"

Section 4.6.2 of the EREF SAR (AES, 2009a) states that respiratory protection procedures

would be revised as necessary whenever changes are made to the facility, processing, or

equipment.

8. "[M]aintain records of the respiratory protection program, including training for respirator use,

and maintenance"

In Section 4.6.2 of the EREF SAR (AES, 2009a), the applicant stated that records for the

respiratory protection program (including training for respirator use and maintenance) would

be maintained in accordance with the facility records management program as described in

Section 11.7 of the EREF SAR (AES, 2009a). Section 11.7 discusses control of records

and maintenance of the master file. It states that records related to health and safety would

be maintained in accordance with the requirements of Title 10, Code of Federal Regulations.

Based on the analysis as summarized above, the staff finds that the commitments in the EREF

SAR (AES, 2009a) satisfactorily address the application acceptance criteria in Section 4.4.6.3 of

NUREG-1520 (NRC, 2002).

4.3.7 Radiation Survey and Monitoring Programs

The staff reviewed the applicant's radiation survey and monitoring program commitments

against the acceptance criteria in NUREG-1 520, Section 4.4.7.3 (NRC, 2002). The following

sections identify each acceptance criteria from NUREG-1 520 and discuss the staffs analysis as

to whether the information provided by the applicant in the SAR (AES, 2009a) meets the criteria.

1. "[Hjave radiation surveys and monitoring programs consistent with the requirements of

10 CFR Part 20, Subpart F"

The applicant committed to written procedures that will assure compliance with 10 CFR Part

20, Subpart F, in Section 4.7 of the EREF SAR (AES, 2009a). 10 CFR Part 20, Subpart F,

has multiple requirements which are satisfied by the applicant's commitments meeting the

other acceptance criteria addressed in this section.

2. "[P]repare written procedures for the radiation survey and monitoring program that include

an outline of the program objectives, sampling procedures, data analysis methods, types of

equipment and instrumentation to be used, frequency of measurements, recordkeeping and

reporting requirements, and actions to be taken when measurements exceed 10 CFR Part

20 occupational dose limits or administrative levels established by the applicant"

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In Section 4.7 of the EREF SAR (AES, 2009a), the applicant committed that the radiation

survey 'and monitoring programs would be consistent with multiple guidance documents

including the ones below:

* RG 8.2, "Guide for Administrative Practice in Radiation Monitoring" (NRC, 1973a)

* RG 8.4, "Direct-Reading and Indirect-Reading Pocket Dosimeters" (NRC, 1973b)

" RG 8.7, "Instructions for Recording and Reporting Occupational Radiation Exposure

Data," Rev. 2 (NRC, 2005)

* RG 8.9, "Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay

Program" (NRC, 1993e)

* RG 8.24, "Health Physics Surveys During Enriched Uranium-235 Processing and Fuel

Fabrication," (NRC, 1979)

* RG 8.25, "Air Sampling in the Workplace" (NRC, 1992a)

" RG 8.34, "Monitoring Criteria and Methods To Calculate Occupational Radiation Doses"

(NRC, 1992b)

" NUREG-1400, "Air Sampling in the Workplace" (NRC, 1993f)

" ANSI/HPS N13.1-1999, "Guide to Sampling Airborne Radioactive Materials in Nuclear

Facilities," (ANSI/HPS, 1999)

* ANSI N323-1978, "Radiation Protection Instrumentation Test and Calibration" (ANSI,

1978)

* ANSI/HPS N13.11-2001, "Dosimetry-Personnel Dosimetry Performance-Criteria for

Testing" (ANSI/HPS, 2001)

* ANSI/HPS N13.22-1995, "Bioassay Program for Uranium" (ANSI/HPS, 1995)

* ANSI/HPS N13.27-1981, "Performance Requirements for Pocket-Sized Alarm

Dosimeters and Alarm Ratemeters," (ANSI/HPS, 1981)

* ANSI/HPS N13.30-1996, "Performance Criteria for Radiobioassay," (ANSI,1996)

* ANSI N13.6-1966 (R1 989), "Practice for Occupational Radiation Exposure Records

Systems" (ANSI, 1989)

The applicant stated that the procedures that would be developed to implement the

programs would include an outline of the program objectives, sampling procedures, data

analysis methods, types of equipment and instrumentation to be used, frequency of

measurements, recordkeeping and reporting requirements, and actions to be taken when

measurements exceed the occupational dose limits in 10 CFR Part 20 or the administrative

levels established by the applicant.

3. "[D]esign and implement a personnel monitoring program for external occupational radiation

exposures that outlines methods or procedures to

a. Identify the criteria for worker participation in the program

b. Identify the types of radiation to the monitored

c. Specify how exposures will be measured, assessed and recorded

d. Identify the type and sensitivity of personal dosimeters to be used, when they will

be used, and how the collected data will be processed and evaluated

e. Identify the facility's administrative exposure levels or action levels at which

actions are taken to investigate the cause of exposures exceeding these levels"

In Section 4.7.5 of the EREF SAR (AES, 2009a), the applicant committed to having all

personnel whose duties require them to enter Restricted Areas wear individual external

dosimetry devices, e.g., passive dosimeters such as thermoluminescent dosimeters (TLDs)

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that are sensitive to beta, gamma and neutron radiation. External dosimetry devices would

be evaluated at least quarterly to ascertain external exposures. The applicant proposes an

administrative limit of 1 rem/y in Table 4.1-1 of the EREF SAR (AES, 2009a). If 25 percent

of the annual administrative limit (i.e., 2.5 mSv or 0.250 rem) is exceeded in any quarter,

then an investigation would be performed and documented to determine what types of

activities may have contributed to the worker's external exposure. If the administrative limit

is exceeded, the Radiation Protection/Chemistry Manager would be informed. The

Radiation Protection/Chemistry Manager would be responsible for determining the need for

and recommending investigations or corrective actions to the responsible manager(s).

4. "[D]esign and implement a personnel monitoring program, for internal occupational radiation

exposures, based on the requirements of 10 CFR 20.1201, 20.1204, and 20.1502(b), that

outlines methods or procedures to:

a. Identify the criteria for worker participation in the program

b. Identify the type of sampling to be used, the frequency of collection and

measurement, and the minimum detection levels

c. Specify how worker intakes will be measured, assessed, and recorded

d. Specify how the data will be processed, evaluated, and interpreted

e. Identify the facility's administrative exposure levels or the levels at which actions

are taken to investigate the causes of exposures exceeding these levels"

The applicant made commitments in Section 4.7.6 of the EREF SAR (AES, 2009a) that

internal exposures for all personnel wearing external dosimetry devices would be evaluated

via direct bioassay (e.g., in vivo body counting), indirect bioassay (e.g., urinalysis), or an

equivalent technique. For soluble (Class D) uranium, regulations in 10 CFR 20.1201(e)

require worker intake be no more than 10 milligrams (mg) of soluble uranium in a week.

This limit is to protect workers from the toxic chemical effects of uranium. The facility annual

administrative limit for the TEDE would be 1.0 rem. Internal doses would be evaluated at

least annually. If the facility annual administrative limit is exceeded as determined from

bioassay results, then an investigation would be performed and documented to determine

what types of activities may have contributed to the worker's internal exposure.

5. "[C]omply with the requirements of 10 CFR 20.1202 for summation of external and internal

occupational radiation exposures through the use of procedures such as those outlined in

RGs 8.7 or 8.34"

In Section 4.7.7 of the EREF SAR (AES, 2009a), the applicant stated that the internal and

external exposure values would be summed in accordance with 10 CFR 20.1202.

Procedures for the evaluation and summation of doses would be based on the guidance

contained in RGs 8.7 and 8.34.

6. "[D]esign and implement an air sampling program in areas of the facility identified as

potential airborne radioactivity areas, to conduct air surveys, and to calibrate and maintain

the airborne sampling equipment in accordance with the manufacturer's recommendations"

Section 4.8.8.2 of the EREF SAR (AES, 2009a) addresses the facility air monitoring and

sampling program. In this section, the applicant states that active on-line monitors for

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airborne alpha emitters that would be used to measure representative airborne

concentrations of radionuclides that may be due to facility operation. On-line monitoring for

gross alpha activity would be performed assuming all the alpha activity is due to uranium.

When airborne activity data would be used for dose calculations, the assumption would be

that all the activity is due to uranium-234, class D material. The lower limit of detection

would be either 0.02 mg of uranium in the total sample or a concentration of 3.7

nanoBecquerel per milliliter (nBq/ml) (1E-13 microCurie per milliliter [pCi/ml]) gross alpha.

An action level would be established at 1 mg of total uranium likely to be inhaled by a worker

in seven days.

In addition, permanent monitors would be operated in the restricted areas to collect

continuous samples. The filters in these monitors would be changed and analyzed weekly

and following any indication of release that might lead to airborne concentrations of uranium

likely to exceed (1) 10 percent of the values listed in 10 CFR 20.1003 or (2) the total

uranium action level of one milligram of total uranium inhaled in one week. The filters would

also be changed each shift following changes in process equipment or process control or

following detection of any event (e.g., leakage, spillage or blockage of process equipment)

that would likely exceed (1) 10 percent of the values listed in 10 CFR 20.1003, or (2) the

total uranium action level of one milligram inhaled by a worker in one week. The

representativeness of workstation air samplers would be checked annually and when

significant process or equipment changes have been made.

In Section 4.7 of the EREF SAR (AES, 2009a), the applicant stated that calibration will be

performed in accordance with written established procedures and documented prior to the

initial use of each airflow measurement instrument (used to measure flow rates for air or

effluent sampling) and each radioactivity measurement instrument. Periodic operability

checks would be performed in accordance with written established procedures. Calibrations

would be performed and documented on each airflow measurement and radioactivity

measurement instrument at least annually (or according to manufacturers'

recommendations, whichever is more frequent) or after failing an operability check, or after

modifications or repairs to the instrument that could affect its proper response, or when it is

believed that the instrument has been damaged.

7. "[I]mplement additional procedures, as may be required by 10 CFR Part 20 and the ISA

Summary, to control the concentration of airborne radioactive material (e.g., control of

access, limitation of exposure times to licensed materials, and use of respiratory protection

equipment)"

The applicant addressed practices contained in this criteria citation in Section 4.6.2 of the

EREF SAR (AES, 2009a). The applicant commits that when it is not possible to control the

concentrations of radioactive material in the air to values below those that define an airborne

radioactivity area, other means would be implemented to maintain the total effective dose

equivalent ALARA. In these cases, the ALARA goal would be met by an increase in

monitoring and the limitation of intakes by one or more of the following means: control of

access; limitation of exposure times; use of respiratory protection equipment; or other

controls, as available and appropriate. Respiratory protection procedures are discussed in

Section 4.3.6(6) of this SER.

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8. "[C]onduct a contamination survey program in areas of the facility identified in the ISA

Summary most likely to be radiologically contaminated (the program must include the types

and frequencies of surveys for various areas of the facility and the action levels and actions

to be taken when contamination levels are exceeded)"

Section 4.8.5.1 of the EREF SAR (AES, 2009a) states that contamination survey monitoring

would be performed for all UF6 process areas and areas in which uranic materials are

handled or stored. Surveys would include routine checks of non-UF6 process areas,

including areas normally not contaminated. Monitoring would include direct radiation and

removable contamination measurements. Survey procedures would be based on the

potential for contamination of an area and operational experience. The Restricted Areas

would be surveyed at least weekly. The lunch room and change rooms would be surveyed

at least daily.

If surface contamination levels exceed the following action levels, clean-up of the

contamination would be initiated within 24 hours of the completion of the analysis:

" Removable contamination: 83.3 Becquerel per 100 square centimeters (Bq/100 cm 2)

(5000 disintegrations per minute per 100 square centimeters [dpm/100 cm2

]) alpha

or beta/gamma

" Fixed contamination: 4.2 kBq/100 cm 2 (250,000 dpm/100 cm 2) alpha or beta/gamma.

In addition, in Section 4.7 of the EREF SER (AES, 2009a), the applicant committed to

developing procedures consistent with guidance found in RG 8.24, "Health Physics Surveys

During Enriched Uranium-235 Processing and Fuel Fabrication," (NRC, 1979). This

guidance establishes frequencies for contamination surveys and administrative limits for

different areas of the facility.

9. "[I]mplement the facility's corrective action program when the results of personnel monitoring

or contamination surveys exceed the applicant's administrative personnel contamination

levels"

The applicant stated in Sections 4.1 and 4.7 of the EREF SAR (AES, 2009a) that the facility

corrective action process would be implemented if: (1) personnel dose monitoring results or

personnel contamination levels exceed the administrative personnel limits; or if an incident

results in airborne occupational exposures exceeding the administrative limits, or (2) the

dose limits in 10 CFR 20, Appendix B, or 10 CFR 70.61 are exceeded. In the event the

occupational dose limits given in 10 CFR 20, Subpart C, would be exceeded, notification of

the NRC would be in accordance with the requirements of 10 CFR 20, Subpart M, "Reports."

10. "[I]mplement the facility's corrective action program when any incident results in airborne

occupational exposures to radiation exceeding the facility's administrative limits, or the dose

limits in 10 CFR Part 20, Appendix B, or 10 CFR 70.61"

See the discussion under number 9 above.

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11. "[U]se equipment and instrumentation with sufficient sensitivity for the type or types of

radiation being measured and to calibrate and maintain equipment and instrumentation in

accordance with manufacturers' recommendations"

Section 4.7 of the EREF SAR (AES, 2009a) states that portal monitors, hand and foot

monitors, and friskers would have the required sensitivity to detect alpha contamination on

personnel to ensure that radioactive materials do not spread to the areas outside the

Restricted Areas. Instruments would be calibrated with sources that are within ±5 percent of

the reference value and are traceable to the National Institute of Standards and Technology

(NIST) or equivalent. The background and efficiency of laboratory counting instruments,

when used for radiation protection purposes, would be determined daily. This determination

would be less frequent only if necessary due to long counting intervals.

Discussion of air sampling and monitoring equipment and calibration/maintenance is

provided in Section 4.3.7(6) of this SER.

12. "[E]stablish policies to ensure equipment and materials removed from restricted areas to

unrestricted areas are not contaminated above the specified release levels in NRC Branch

Technical Position, "Guidelines for Decontamination of Facilities and Equipment Prior to

Release for Unrestricted Use or Termination of Licenses for Byproduct, Source, or Special

Nuclear Material," April 1993"

The applicant stated in Section 4.7.8 of the EREF SAR (AES, 2009a) that monitor stations

will be established at the entry and exit points for Restricted Areas. Monitors would be

provided to detect radioactive contamination on personnel and their personal items,

including hard hats. All personnel would be required to monitor themselves, any handcarried personal items, and hard hats prior to exiting a Restricted Area. In Section 4.7.2 of

the EREF SAR (AES, 2009a), the applicant states that action levels for skin and personal

clothing contamination at the point of egress from Restricted Areas and any additional

designated areas within the Restricted Area (e.g., a Contaminated Area which is provided

with a step-off pad and contamination monitor) would not exceed 2.5 Bq/100 cm2 (150

dpm/100 cm2

) alpha or beta/gamma contamination (corrected for background). Clothing

contaminated above egress limits shall not be released unless it can be laundered to within

these limits. If skin or other parts of the body are contaminated above egress limits,

reasonable steps that exclude abrasion or other damage will be undertaken to effect

decontamination.

The applicant further stated in Section 4.10 of the EREF SAR (AES, 2009a) that the facility

will follow NRC Branch Technical Position (BTP): "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). This guide would apply

to the abandonment or release for unrestricted use, of surfaces, premises and equipment.

13. "[L]eak-test all sealed sources in accordance with the following NRC Branch Technical

Positions: (1) "License Condition for Leak-Testing Sealed Byproduct Material Sources," April

1993, (2) "License Condition for Leak-Testing Sealed Plutonium Sources," April 1993, (3)

"License Condition for Plutonium Alpha Sources," April 1993, (4) "License Condition for

Leak-Testing Sealed Source Which Contains Alpha and/or Beta-Gamma Emitters," April

1993, and (5) "License Condition for Leak-Testing Sealed Uranium Sources," April 1993"

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Section 4.11.1 of the EREF SAR (AES, 2009a) states that leak-testing of sources would be

performed in accordance with the following NRC BTPs:

* License Condition for Leak-Testing Sealed Byproduct Material Sources (NRC,1993b)

* License Condition for Leak-Testing Sealed Source Which Contains Alpha and/or BetaGamma Emitters (NRC,1993c)

* License Condition for Leak-Testing Sealed Uranium Sources (NRC, 1993d).

14. "[E]stablish and implement an access control program that ensures that (a) signs, labels,

and other access controls are properly posted and operative, (b) restricted areas are

established to prevent the spread of contamination and are identified with appropriate signs,

and (c) step-off pads, change facilities, protective clothing facilities, and personnelmonitoring instruments are provided in sufficient quantities and locations"

The applicant stated in Section 4.7.2 of the EREF SAR (AES, 2009a) that the facility would

establish and implement an access control program that ensures that (a) signs, labels, and

other access controls are properly posted and operative, (b) restricted areas are established

to prevent the spread of contamination and are identified with appropriate signs, and (c)

step-off pads, change facilities, protective clothing facilities, and personnel monitoring

instruments are provided in sufficient quantities and locations.

Because there will be no High Radiation Areas in the facility, there would be no areas where

access is physically prevented due to radiation level. Access control would be by

administrative methods. Access to certain areas may be physically prevented for security

reasons. Personnel who have not been trained in radiation protection procedures would not

be allowed access to a Restricted Area without escort by other trained personnel.

Access to and egress from a Restricted Area would be through one of the monitor stations

at the particular Restricted Area boundary. Access to and egress from each Radiation Area,

Contaminated Area or Airborne Radioactivity Area within a Restricted Area may also be

individually controlled. A contamination monitor (e.g., frisker, hand and foot monitor or

portal monitor), step-off pad, and container for any discarded protective clothing may be

provided at the egress point from certain of these areas to prevent the spread of

contamination.

15. "[H]ave a radiation reporting program consistent with the requirements of 10 CFR Parts 19

and 20"

The applicant states in Section 4.7 of the EREF SAR (AES, 2009a) that the written

procedures for the radiation survey and monitoring programs would assure compliance with

the requirements of 10 CFR 20, Subparts L, "Records," and M, "Reports." Procedures

would also be consistent with guidance in RG 8.7, "Instructions for Recording and Reporting

Occupational Radiation Exposure Data," Rev. 2 (NRC, 2005). The applicant further

elaborated in Section 4.11.2 (AES, 2009a) that the facility would meet the regulations for the

additional program commitments applicable to records and reports: 10 CFR 20 Subpart L,

"Records;" Subpart M, "Reports;" Section 70.61, "Performance requirements," and Section

70.74, "Additional Reporting Requirements." The facility would maintain complete records of

the Radiation Protection Program for at least the life of the facility.

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By procedure, the facility would report to the NRC, within the time specified in 10 CFR

20.2202 and 10 CFR 70.74, any event that results in an occupational exposure to radiation

exceeding the dose limits in 10 CFR Part 20. The facility would prepare and submit to the

NRC an annual report of the results of individual monitoring, as required by 10 CFR

20.2206(b).

Section 11.7 of the EREF SAR (AES, 2009a) discusses record management for the facility.

This section states that records related to health and safety would be maintained in

accordance with the requirements of Title 10, Code of Federal Regulations.

The applicant stated in Section 2.3.7 of the EREF SAR (AES, 2009a) that employees would

have access to various resources to ensure their safety or quality concerns are addressed

including NRC's requirements under 10 CFR Part 19, "Notices, Instructions and Reports to

Workers: Inspection and Investigations."

Based on the analysis as summarized above, the staff finds that the commitments in the EREF

SAR (AES, 2009a) satisfactorily address the application acceptance criteria in Section 4.4.7.3 of

NUREG-1520 (NRC, 2002).

4.3.8 Additional Program Requirements

The staff reviewed the applicant's additional program commitments against the acceptance

criteria in NUREG-1 520, Section 4.4.8.3 (NRC, 2002). The following sections identify each

acceptance criteria from NUREG-1 520 and discuss the staff's analysis as to whether the

information provided by the applicant in the SAR (AES, 2009a) meets the criteria.

1. "[M]aintain records of the RP program (including program provisions, audits, and reviews of

the program content and implementation), radiation survey results (air sampling, bioassays,

external-exposure data from monitoring of individuals, internal intakes of radioactive

material), and results of its corrective action program referrals, RWPs, and planned special

exposures"

The applicant stated in Section 4.11.2 of the EREF SAR (AES, 2009a) that the facility would

maintain complete records of the RP Program for at least the life of the facility. This would

specifically include: RP program provisions, audits, and reviews of the program content

and implementation, radiation survey results (air sampling, bioassays, external-exposure

data from monitoring of individuals, internal intakes of radioactive material), and results of

corrective action program referrals, RWPs and planned special exposures.

2. "[E]stablish a program to report to the NRC, within the time specified in 10 CFR 20.2202 and

10 CFR 70.74, any event that results in an occupational exposure to radiation exceeding the

dose limits in 10 CFR Part 20"

As stated in Section 4.11.2 of the EREF SAR (AES, 2009a), the applicant would develop

procedures such that the facility would report to the NRC, within the time specified in

10 CFR 20.2202 and 10 CFR 70.74, any event that results in an occupational exposure to

radiation exceeding the dose limits in 10 CFR 20.

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3. "[P]repare and submit to the NRC an annual report required by 10 CFR 20.2206(b)"

Also in Section 4.11.2 of the EREF SAR (AES, 2009a), the applicant committed to prepare

and submit to the NRC an annual report of the results of individual monitoring, as required

by 10 CFR 20.2206(b).

4. "[R]efer to its corrective action program any radiation incident that results in an occupational

exposure that exceeds the dose limits in 10 CFR Part 20, Appendix B, or is required to be

reported per 10 CFR 70.74, and to report to the NRC both the corrective action taken (or

planned) to protect against a recurrence and the proposed schedule to achieve compliance

with the applicable license condition or conditions"

The applicant stated in Sections 4.1 and 4.11.2 of the EREF SAR (AES, 2009a) that the

facility corrective action process would be implemented if: (1) personnel dose monitoring

results or personnel contamination levels exceed the administrative personnel limits; or if an

incident results in airborne occupational exposures exceeding the administrative limits; or

(2) the dose limits in 10 CFR 20, Appendix B or 10 CFR 70.61 are exceeded.

If the dose limits in 10 CFR 20, Appendix B, or 10 CFR 70.61 would be exceeded, or an

event is required to be reported per 10 CFR 70.74, the NRC would be informed of the

corrective action taken or planned to prevent recurrence and the schedule established by

the facility to achieve full compliance.

Based on the analysis as summarized above, the staff finds that the commitments in the EREF

SAR (AES, 2009a) satisfactorily address the application acceptance criteria in Section 4.4.8.3 of

NUREG-1520 (NRC, 2002).

4.4 Evaluation Findings

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

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7. Other programs to maintain records, the applicant will report to the NRC in accordance with

Parts 20 and 70-and correct for upsets at the facility.

As discussed in its application and based on the staff's analysis, the applicant's RP program

would meet the applicable requirements of Parts 19, 20, 30, 40, and 70.

4.5 References

(AES, 2009a) AREVA Enrichment Services LLC, "Eagle Rock Enrichment Facility Safety

Analysis Report," Revision 1, 2009.

(AES, 2009b) AREVA Enrichment Services LLC, "Eagle Rock Enrichment Facility Integrated

Safety Analysis Summary," Revision 1, 2009.

(ANSI, 1992) American National Standards Institute, American National Standard for

Respiratory Protection, ANSI Z88.2, "Practices for Respiratory Protection," 1992.

(ANSI, 1989) American National Standards Institute, ANSI N13.6-1966 (R1989), "Practice for

Occupational Radiation Exposure Records Systems," 1989.

(ANSI, 1978) American National Standards Institute, ANSI N323-1978, "American National

Standard Radiation Protection Instrumentation Test and Calibration" 1978.

(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.

(ANSI/ASME, 1989) American National Standards Institute/American Society of Mechanical

Engineers,. ANSI/ASME N510-1989, "Testing of Nuclear Air Cleaning Systems," 1989.

(ANSI/HPS, 2001) American National Standards Institute/Health Physics Society, ANSI/HPS

N13.11-2001, "Dosimetry-Personnel Dosimetry Performance-Criteria for Testing," 2001.

(ANSI/HPS, 1999) American National Standards Institute/Health Physics Society, ANSI/HPS

N13.1-1999, "Sampling and Monitoring Releases of Airborne Radioactive Substances From the

Stacks and Ducts of Nuclear Facilities," 1999.

(ANSI/HPS, 1996) American National Standards Institute/Health Physics Society, ANSI/HPS

N13.30-1996, "Performance Criteria for Radiobioassay," 1996.

(ANSI/HPS, 1995) American National Standards Institute/Health Physics Society, ANSI/HPS

N13.22-1995, "Bioassay Program for Uranium," 1995.

(ANSI/HPS, 1981) American National Standards Institute/Health Physics Society, ANSI/HPS

N13.27-1981, "Performance Requirements for Pocket-Sized Alarm Dosimeters and Alarm

Ratemeters," 1981.

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(DOE, 2003) U.S. Department of Energy, DOE-HDBK-1 169-2003, "DOE Handbook Nuclear Air

Cleaning Handbook," 2003.

(NRC, 2005) U.S. Nuclear Regulatory Commission, Regulatory Guide 8.7, Rev. 2,

"Instructions for Recording and Reporting Occupational Radiation Exposure Data," 2005.

(NRC, 2002) U.S. Nuclear Regulatory Commission, NUREG-1 520, "Standard Review Plan for

the Review of a License Application for a Fuel Cycle Facility," 2002.

(NRC 1999) U.S. Nuclear Regulatory Commission, Regulatory Guide 8.15, "Acceptable

Programs for Respiratory Protection," 1999.

(NRC, 1992a) U.S. Nuclear Regulatory Commission, Regulatory Guide 8.25, "Air Sampling in

the Workplace," 1992.

(NRC, 1992b) U.S. Nuclear Regulatory Commission, Regulatory Guide 8.34, "Monitoring

Criteria and Methods to Calculate Occupational Radiation Doses," 1992.

(NRC, 1993a) U.S. Nuclear Regulatory Commission, Branch Technical Position, "Guidelines for

Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or

Termination of Licenses for Byproduct, Source, or Special Nuclear Material," 1993.

(NRC, 1993b) U.S. Nuclear Regulatory Commission, Branch Technical Position, "License

Condition for Leak-Testing Sealed Byproduct Material Sources," 1993.

(NRC, 1993c) U.S. Nuclear Regulatory Commission, 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, Branch Technical Position, "License

Condition for Leak-Testing Sealed Uranium Sources," 1993.

(NRC, 1993e) U.S. Nuclear Regulatory Commission, Regulatory Guide 8.9, "Acceptable

Concepts, Models, Equations, and Assumptions for a Bioassay Program, 1993.

(NRC, 1993f) U.S. Nuclear Regulatory Commission, NUREG-1400, "Air Sampling in the

Workplace," 1993.

(NRC, 1991) U.S. Nuclear Regulatory Commission, NUREG-1400, "Air Sampling in the

Workplace," 1991.

(NRC, 1979) U.S. Nuclear Regulatory Commission, Regulatory Guide 8.24, "Health Physics

Surveys During Enriched Uranium-235 Processing and Fuel Fabrication," 1979.

(NRC, 1977) U.S. Nuclear Regulatory Commission, Regulatory Guide 8.10, "Operating

Philosophy for Maintaining Occupational Radiation Exposures As Low As Is Reasonably

Achievable," 1977.

4-23
(NRC, 1973a) U.S. Nuclear Regulatory Commission, Regulatory Guide 8.2, "Guide for

Administrative Practice in Radiation Monitoring," 1973.

(NRC, 1973b) U.S. Nuclear Regulatory Commission, Regulatory Guide 8.4, "Direct-Reading

and Indirect-Reading Pocket Dosimeters," 1973b.

4-24
CHAPTER 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 operation of the facility, as required by Title 10 of

the Code of Federal Regulations (CFR) Part 70. The applicant's Safety Analysis Report (SAR),

Integrated Safety Analysis (ISA) Summary, and responses to requests for additional information

(RAI) provided the basis for this determination. The information and commitments which are

significant to this determination are described in Section 5.3 of this chapter.

The NCS programmatic review determines whether: (1) the applicant provided for the

appropriate management of the NCS program; (2) the applicant 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

10 CFR Part 70.

5.1 Regulatory Requirements

The review of AREVA Enrichment Service's (AES's) NCS program should verify that the

information the applicant provided meets the requirements of 10 CFR 70.22 and 70.65, which,

respectively, specify the general and additional content of an 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.

5.2 Regulatory Acceptance Criteria

The acceptance criteria for the U.S. Nuclear Regulatory Commission's (NRC's) review of the

applicant's NCS program are outlined in NUREG-1 520, Section 5.4 (NRC, 2002). This includes

the commitment to use Regulatory Guide (RG) 3.71, Revision 1 (NRC, 2005), which endorses

the use of the American National Standards Institute/American Nuclear Society (ANSI/ANS)

Series-8 NCS standards with certain modifications. In addition, the acceptance criteria for the

NRC's review of the applicant's ISA methodology and ISA Summary are outlined in

NUREG-1520, Section 3.4 (NRC, 2002).

5.3 Staff Review and Analysis

5.3.1 Industry Standards

The acceptance criteria for NRC's review of industry standards are contained in Sections 5.4

and 5.4.2 of NUREG-1520 (NRC, 2002).

5-1
AES commits to RG 3.71, Revision 1 (NRC, 2005).5 Thus, AES will follow the requirements

(i.e., "shall" statements) in the ANSI/ANS Series-8 standards, as endorsed by the NRC. A

commitment to these standards is an acceptable means of meeting many of the acceptance

criteria in Section 5.4 of NUREG-1 520.

In committing to RG 3.71, AES commits to comply with all the ANS standards as endorsed by

NRC. However, some of the standards are not applicable because of the nature of facility

operations. The staff determined that the following ANSI/ANS Series-8 standards do not apply

to any activities that the Eagle Rock Enrichment Facility (EREF) will be conducting: ANSI/ANS8.5-1996, -8.6-1983, -8.12-1987, -8.15-2004, -8.17-2004, and -8.21-1995. While AES has

committed to these standards, this has no impact on its operations as AES would have to

request a license amendment to conduct activities for which these standards apply. Although

ANSI/ANS-8.10-1983, Criteria for Nuclear Criticality Safety Controls in Operations with

Shielding and Confinement, (ANSIIANS, 1983), does not apply to any operations at EREF,

because it is not a shielded facility, the technical information in this standard will be used to

determine the consequences of a criticality accident.

5.3.2 Organization and Administration

The acceptance criteria for NRC's review of organization and administration are contained in

Sections 5.4.3.2 and 2.4 of NUREG-1 520 (NRC, 2002).

During all phases of facility operation (design, construction, operation, and decommissioning)

EREF will comply with ANSI/ANS-8.1-1983, Nuclear Criticality Safety in Operations with

Fissionable Material Outside Reactors (ANSI/ANS, 1998), and ANIS/ANS-8.19-2005,

Administrative Practices for Nuclear Criticality Safety (ANSI/ANS, 2005), as they relate to

organization and administration (AES, 2010b). The NCS function will be performed within the

engineering organization during the design of the facility. When the facility transitions from

design (including construction) to operations the NCS function will transition to an independent

function within the Environmental, Health, Safety and Licensing (EHS&L) function.

The EHS&L function at the proposed EREF will be created during the transition to operations

and will be independent of production responsibilities (AES, 201 Ob). During the operations

phase, the manager of the NCS function reports to the EHS&L manager. The NCS function has

the authority to shutdown potentially unsafe operations, and must approve restart of any

operation it has requested to be shutdown. The minimum qualifications for the NCS manager

are a bachelor's degree (or equivalent) in an engineering or scientific field and four years of

nuclear experience. At least one year of direct experience with the administration of NCS

analyses and evaluations is also required.

NCS engineers report to the engineering organization managers during the design phase and to

the NCS Manager during the operations phase of the facility (AES, 201 Ob). The minimum

qualifications for an NCS engineer are a bachelor's degree in an engineering or scientific field,

two years experience in the implementation of a criticality safety program, and successful

completion of an NCS training program. In addition to these requirements, individuals who

5This regulatory guide endorses, with some exceptions, specific nuclear criticality safety standards developed by the

American Nuclear Society's Standards Subcommittee 8 (ANS-8), "Operations with Fissionable Materials Outside

Reactors."

5-2
perform NCS analyses or evaluations for facility changes shall have at least two years

experience with NCS programs and performing NCS analyses.

The staff has reviewed AES's organizational structure and finds that it is acceptable because

the applicant has adequately addressed the acceptance criteria in NUREG-1 520, Sections 2.4

and 5.4.3.2 as they relate to NCS. The NCS function will be independent from the production

staff, NCS evaluations are performed by qualified reviewers, with independent review for quality

assurance, and AES's administrative practices are consistent with the requirements in

ANSI/ANS-8.19-2005.

5.3.3 Management of the NCS Program

The acceptance criteria for NRC's review of the management of the NCS program are

contained in Section 5.4.3.1 of NUREG-1520 (NRC, 2002).

As described in Section 5.1.1 of the SAR (AES, 2010b), the NCS program will evaluate changes

to operations, recommend changes to processes to maintain safe operations, and select

appropriate items relied on for safety (IROFS) and management measures. Section 5.1.5 of the

SAR (AES, 201 Ob) describes the responsibilities of the NCS manager for the development and

implementation of the NCS program. These responsibilities include:

* Performing NCS analyses and evaluations

" Establishing limits and controls

• Assuring limits and controls are properly implemented

" Monitoring plant compliance with NCS requirements.

As described in Section 5.2.1.4 of the SAR (AES, 201 Ob), during both the design and operating

phases, qualified NCS engineers will prepare and independently review NCS analyses. Once

operating, NCS engineers are also responsible for conducting and documenting NCS

evaluations and assessments. During the operating phase, the NCS Manager must approve

NCS analyses and evaluations (AES, 201 Ob).

At least one NCS engineer will be available when the facility is operating to respond to any

routine request or emergency condition. As stated in Section 4.2.2.1 of the Emergency Plan

(AES, 2010b), in the event of a fire, at least one responder will be assigned as the criticality

safety officer who is responsible for ensuring that NCS is not compromised during firefighting

activities.

As described in Section 5.1.1 of the SAR, the NCS program will be updated to reflect changes

in the ISA or NCS methodologies (AES, 2010b). NCS program records, which include NCS

analyses and evaluations, will be retained in accordance with the AES document control and

records management procedures (See Chapter 11 of this SER) (AES, 201 Ob).

The staff has reviewed the applicant's management of the NCS program and finds that it is

acceptable because the applicant has adequately addressed the acceptance criteria in

NUREG-1520, Section 5.4.3.1.

5-3
5.3.4 NCS Management Measures

The acceptance criteria for NRC's review of NCS management measures are contained in

Sections 5.4.3.3 and 5.4.3.4.7 of NUREG-1520 (NRC, 2002).

5.3.4.1 Training

The training program for personnel who handle nuclear material is based upon ANSI/ANS-8.20-

1991, Nuclear Criticality Safety Training (ANSI/ANS, 1991). The training will include the

required response to an NCS deviation and to the activation of the criticality accident alarm

(AES, 201 Ob). Successful completion of nuclear safety training will be required to gain

unescorted access into the Controlled Access Area (CAA) of the facility; retraining must be

completed annually to maintain access (AES, 2010b).

The staff has reviewed the applicant's commitments regarding personnel training as they relate

to NCS and finds them acceptable because the applicant has adequately addressed the

acceptance criteria in NUREG-1520, Section 5.4.3.3(1).

5.3.4.2 Procedures

AES commits to ANSI/ANS-8.19-2005 (ANSI/ANS, 2005) as it relates to procedures. Personnel

will only perform actions in accordance with approved written procedures. Procedures will be

written such that no single, inadvertent departure from a procedure could cause a nuclear

criticality accident (AES, 201 Ob). If a particular situation is not covered by procedure, then

personnel shall report the issue and take no further action until the situation has been evaluated

and recovery procedures are provided.

In addition to procedures, postings will be used to identify administrative controls in applicable

work areas (AES, 2010b). These postings will be maintained current (AES, 201 Ob).

The staff has reviewed the applicant's commitments regarding procedures as they relate to NCS

and finds them acceptable because the applicant has adequately addressed the acceptance

criteria in NUREG-1520, Section 5.4.3.3(2).

5.3.4.3 Audits and Assessments

The NCS program will use operational inspections, audits, and investigations to promptly detect

NCS deficiencies (AES 2010b). NCS deficiencies will be entered into the corrective action

program to prevent recurrence (AES, 2010b). Records of the corrective actions will be retained

(AES, 2010b).

AES commits to ANSI/ANS-8.19-2005 (ANSI/ANS, 2005) as it relates to NCS audits and

assessments (AES, 2010b). Audits will be conducted to verify compliance with regulations,

procedures, and the license (AES, 2010b). NCS audits will be conducted and documented

quarterly such that the entire NCS program will be audited at least every two years (Section

11.5.2, AES, 2010b). Weekly NCS walkthroughs of uranium hexafluoride (UF6) process areas

will also be conducted and documented (AES, 201 Ob).

5-4
NCS assessments are performed under the direction of the NCS staff by personnel who are not

directly involved in the function or area being assessed (AES, 2010b). Assessments are

focused on the effectiveness of activities and ensuring that Quality Assurance Level 1 and 2

items are available and reliable (AES, 2010b). The Operations Department will be periodically

assessed to ensure that NCS procedures are being followed and process conditions have not

adversely changed (AES, 2010b). Assessments will be conducted at least semi-annually, but

the exact frequency will be based on the identified NCS controls (AES, 201 Ob).

The staff has reviewed the applicant's commitments regarding audits and assessments as they

relate to NCS and finds them acceptable because the applicant has adequately addressed the

acceptance criteria in NUREG-1520, Section 5.4.3.3(3).

5.3.5 NCS Methodologies and Technical Practices

The acceptance criteria for NRC's review of NCS methodologies and technical practices are

contained in Sections 5.4.3.4.1, 5.4.3.4.4, and 5.4.3.4.5 of NUREG-1 520 (NRC, 2002).

AES commits to use the acceptance criteria in NUREG-1 520, Section 5.4.3.4, "Methodologies

and Technical Practices" (NRC, 2002), with the exception of Sections 5.4.3.4.2 (9), (13), and

(15) (RAI NCS-1 and AES, 201 Ob), when analyzing criticality accidents. In the response to RAI

NCS-1, AES stated that the acceptance criteria in Section 5.4.3.4.2(9) will not be used because

there is no planned use of density as a control parameter; that the criteria in

Section 5.4.3.4.2(13) will not be used because there is no planned use of concentration as a

control parameter; and the criteria in Section 5.4.3.4.2(15) will not be used as there is no

planned use of neutron absorption as a control parameter (AES, 2009a). In addition, the criteria

in NUREG-1 520, Section 5.4.3.4.1(4) does not apply to NCS analysis but does apply to the

criticality accident alarm system (CAAS) (AES, 2009a). Further information about how AES will

comply with these criteria is described in this section.

5.3.5.1 NCS Methodologies

Double Continqency Principle

AES commits to meet the double contingency principle for each process where a criticality

accident is possible (AES, 2010b). AES will incorporate into process designs sufficient factors

of safety to ensure that at least two unlikely, independent, and concurrent changes in process

conditions are required before a criticality accident is possible (AES, 2010b). AES's

commitment to NUREG-1520, Section 5.4.3.4, commits it to the policy that no single credible

event or failure can result in a criticality accident (AES, 2010b).

The NCS analyses and evaluations will show how-the double contingency principle is met for

each process, and it will be used to determine NCS controls and items relied on for safety

(IROFS) (AES, 2010b). AES commits to NUREG-1520, Section 5.4.3.4, thus double

contingency will be assured by independent controls on at least two parameters or by multiple

independent controls on a single parameter (AES, 2010b). The exception to the double

contingency principle described in NUREG-1 520, Section 5.4.3.4.4(7)(c) will not apply to EREF

since AES explicitly commits to meet the double contingency principle as defined in

ANSI/ANS-8.1-1998 for all processes (AES, 2010b).

5-5
NCS Determinations

AES will ensure that each process is adequately subcritical by demonstrating that the effective

neutron multiplication (keff) is less than 0.95 for normal and credible abnormal conditions

(AES, 2010b). keff will be calculated using the equation:

keff = k, + 3o,

where kc is the calculated neutron multiplication factor and a is the associated standard

deviation.

NCS analyses and the safe values listed SAR Table 5.1-1 (see Table 5.3-1 of this chapter) form

the nuclear criticality safety basis for the facility. NCS analyses describe the calculations

performed to demonstrate that each process will be adequately subcritical under normal and

credible abnormal conditions (AES, 2010b). Each NCS analysis will include a discussion of any

assumptions used and will identify required limits and controls (AES, 2010b).

An NCS evaluation is prepared and approved whenever there is a change that involves or could

affect uranium. The NCS evaluation determines if existing NCS analyses bound the change

being evaluated or if new or revised NCS analyses are required (AES, 2010b). The basis for

the evaluation is documented to allow the independent review by a second NCS engineer to

confirm the analyst's conclusions (AES, 2010b). Controlled parameters and limits upon which

NCS depends must be determined and identified in the NCS evaluation (AES, 2010b). NCS

analyses and evaluations will be performed in accordance with ANSI/ANS-8.1-1998 and

ANSI/ANS-8.19-2005 requirements (NRC, 2005). NCS analyses and evaluations will be used

to determine operating limits on controlled parameters that will be as or more conservative than

the safety limits derived from the NCS evaluations and analyses to ensure operations will be

subcritical. NCS operating limits will be derived from safety limits and provide an additional

margin of safety (AES, 2010b). This additional safety margin will account for the variability and

uncertainty in processes and take into consideration changes in process parameters.

The facility will be designed and operated in accordance with SAR Table 5.1-2, which specifies

safety criteria for many of the systems and components that will be used (AES, 2010b). These

criteria are primarily based on the EREF safe values for 6 weight percent (wt%) uranium-235

(

235U) (See Table 5.3-1).

5-6
Table 5.3-1: Comparison of the EREF Safe Values for 6 wt% 235U and the NRC Endorsed

Single Parameter Limits and Safety Factors

Review Comparison EREF Safe Value

Parameter Single Parameter Calculated Safety for 6 wt%

Limit SRP Safety Factor Limits [4]

Volume 30.6 L [1] 0.75 22.95 L 19.3 L

Cylinder Diameter 26.6 cm [1] 0.90 23.94 cm 22.4 cm

Slab Thickness 12.6 cm [1] 0.85 10.71 cm 10.1 cm

Uranium Mass N/A 32.8 kg U 20.1 kg U

-no double batching 1.64 kg 235U [1] 0.75 24.6 kg U 19.4 kg U

-double batching 0.45 14.76 kg U 12.2 k U

Areal Density 0.40 g "JOU/cmL [2] N/A 8.0 g U/cm 7.9 g U/cm

Water Mass [3] N/A N/A N/A 11.9 kg H20

[1] ANSI/ANS-8.1-1998, 5 wt% 235U;

[2] ANSI/ANS-8.1-1998, 100 wt% 2 35 U;

[3] ANSI/ANS-8.1-1998 does not have a single parameter limit for water;

[4] Calculated safety limits assume 5 wt% 235U; they are the single parameter limit multiplied by the SRPrecommended safety factor, and are expressed in the same units as EREF safe values.

The details for how the safe values for the parameters listed in SAR Table 5.1-1 were

determined are described in Section 5.1.2 of the SAR (AES, 2010b). AES calculated the safe

values for the parameters in the table above assuming optimally moderated U0 2 F2 solution with

full water reflection (i.e., 30 cm of tight-fitting water reflection around fissile units), for both 4 and

6 wt% 23U. The safe values correspond to a calculated keff of 0.95. Although AES will only

enrich uranium up to 5 wt% 235U, it will establish all its process safety limits, except for the

contingency dump system traps and tails cylinders, assuming 6 wt% 235U which provides added

safety margin. Thus, an independent review of only the safe values listed for 6 wt% 235U was

performed by NRC staff. Except for water mass, Table 5.3-1 of this SER shows that the values

calculated by AES at 6 wt% 235U are conservative in comparison to well-established single

parameter limits at 5 wt% 235U when combined with an acceptable safety factor. The staff also

determined, based on its previous experience and independent calculations for a spherical

U0 2F2-water system, that the difference between analyzing at 5 and 6 wt% 235U usually results

in conservatism in keff of about 3%. This is significant because it represents additional margin

beyond that provided by the use of a 0.95 keff limit, and provides a high degree of assurance

that processes calculated to be subcritical will be subcritical.

Only product cylinders (30B and 48Y) will use moderation as a controlled parameter (AES,

201 Ob). SAR Table 5.1-2 indicates that product cylinders will be limited to less than 9.3 kg of

water, which represents a safety factor of about 65% when compared to the safe value EREF

calculated for 5 wt% 235U of 14.2 kg of water (AES, 201 Ob) which represents a substantial

margin of safety.

The values listed in SAR Table 5.1-1 and 5.1-2 are adequately subcritical for single units

(AES, 2010b). However, AES has also committed to demonstrate that each process is

adequately subcritical (i.e., keff < 0.95 for normal and credible abnormal conditions) when

multiple units can interact (see Section 5.3.5.2 of this SER) (AES, 2010b).

5-7
Computer Code Validation

NCS analyses are conducted using MONK8A, a well-established and well-known Monte Carlo

computer code used to determine keff. Details about the code and a summary of its validation

are described in SAR Section 5.2.1 (AES, 2010b). Portions of the MONK8A validation report

(AES, 2008) were reviewed by NRC staff to verify that the validation was consistent with the

SAR and was relevant to facility operations. AES commits to report any change in the validation

report to the NRC by letter. Use of computer codes for NCS analysis other than MONK8A

would require NRC approval since the SAR would need to be amended to authorize additional

codes.

The MONK8A validation established a bias, bias uncertainty, and upper subcritical limit (USL) in

accordance with NRC guidance (NRC, 2001). The validation does not use a positive bias, and

the bias uncertainty was determined to be 0.0092. Using a 0.05 margin of subcriticality, which

is widely accepted in the nuclear industry for low-enriched fuel applications, the validation report

established a USL of 0.9408 for everything except the contingency dump system traps and tails

cylinders (AES, 201 Ob). For the contingency dump system traps and tails cylinders, an

additional penalty of 0.0014 was taken since the area of applicability was extrapolated to include

1.5 wt% 235

U (AES, 201 Ob). The USL for the contingency dump system traps and tails cylinders

was therefore determined to be 0.9394 (AES, 2008). Despite establishing the USL as described

in the MONK8A validation report, AES will use 0.95 as its subcritical limit in all its NCS analyses

(Section 5.2.1.2 of (AES, 2010b)). This is equivalent to having a reduced subcritical margin of

0.0394 for the contingency dump system traps and tails cylinders and 0.0408 for all other

systems. The justification for this reduced margin is discussed below.

The validation used well-known benchmark experiments, which consisted of 34 experiments

with uranium enrichment between 4.46-5.64 wt% 235U (Group 1), 30 experiments with uranium

enriched to 9.97 wt% 235

235 U (Group 2), and 29 experiments with uranium enriched to 29.83 wt%

U (Group 3). 'Based solely upon enrichment, the first group of experiments would be most

applicable to the proposed operations at the proposed EREF. The staff questioned whether the

higher enrichment experiments should be included in the validation study, because it was not

readily apparent that they were applicable to operations at the proposed EREF. The inclusion

of inapplicable experiments could decrease the calculated bias or extend the validated area of

applicability inappropriately. Because the applicability of these experiments was not readily

apparent, the staff examined the impact of including the higher enriched experiments in the

validation, by using the 0.95 subcritical limit to calculate an effective subcritical margin assuming

no extrapolation penalty is taken. The goal of this was to determine whether the inclusion of

these experiments would materially change the validation results. When only Group 1 is used in

the validation, the subcritical margin is reduced to 0.0358. When Group 1 and 2 are used, the

subcritical margin is increased to 0.0416. As stated above, when all three groups are used, the

subcritical margin is 0.0408. AES has not provided justification for including experiments with

uranium enrichments well outside the range required for the application; however, the inclusion

or exclusion of the higher enrichment benchmarks produces only a very small difference in the

USL.

A subcritical margin of 0.05 has long been accepted without extensive justification for lowenriched fuel facility processes (as discussed in (NRC, 2006). However AES is requesting a

smaller margin and has included experiments well outside the credible operating parameters

(AES, 2008). Performing NCS analyses at 6 wt% 235U provides a significant additional safety

margin (about 3% in keff) that is sufficient to compensate for the effect of including or excluding

5-8
the higher enriched benchmark experiments, and is also large compared to the difference

between the USL and the 0.95 subcritical limit. The subcritical limit of 0.95 is therefore justified

by the presence of this additional conservatism. The staff also determined that the bias and

bias uncertainty are slowly varying quantities as a function of enrichment, and therefore any

reasonably anticipated change in the USL resulting from extrapolating the area of applicability

(AOA) from the range of Group 1 (up to 5.64 wt% 235U) to 6 wt% would be more than

compensated for by this same conservative margin.

For the contingency dump system traps and tails cylinders, which are the only items analyzed

with 1.5 wt% 235U, significant changes in multiple operating parameters would be required

before the system could approach the subcritical limit. This large difference between operating

conditions and the safety limit for the contingency dump system traps and tails cylinders

provides an additional safety margin that is sufficient to justify the 0.95 subcritical limit for these

systems. Based on the above review, the staff concluded that there will be an adequate margin

of subcriticality for safety to satisfy 10 CFR 70.61(d).

5.3.5.2 NCS Technical Practices

NCS Controls and Controlled Parameters

As described in Section 5.1.1 of the SAR (AES, 2010b), all NCS controls will be designed to

prevent a criticality accident; there are no mitigative NCS controls (AES, 201 Ob). Where

possible, AES will use passive engineered controls to ensure NCS (AES, 2010b). Through its

commitment to NUREG-1520, Section 5.4.3.4.2, the applicant has committed to select controls

in accordance with the following order of preference: (1) passive engineered; (2) active

engineered; (3) augmented administrative; and (4) simple administrative (AES, 2010b).

AES commits to establish NCS controls to ensure that under normal and credible abnormal

conditions all nuclear processes are adequately subcritical (AES, 2010b). This means that NCS

controls must be capable of maintaining the parameter limits established in the NCS analyses

and evaluations (AES, 2010b). AES commits to NUREG-1520, Section 5.4.3.4, thus:

* NCS limits will be derived assuming credible optimal condition (i.e., most reactive

conditions physically possible or limited by written commitment to the NRC) unless

controls are implemented to limit a parameter to a certain range.

" Heterogeneous effects will be considered when evaluating controlled parameters.

* NCS analyses will show that each controlled parameter will be maintained during both

normal and credible abnormal conditions.

The parameters which AES may control for NCS purposes are: mass, geometry/volume,

enrichment, reflection, moderation, and interaction. Commitments associated with these

controlled parameters are discussed in more detail below.

Mass

When mass is used as an NCS controlled parameter, analysis or sampling will be used to verify

the mass of material (AES, 201 Ob). Records will be kept for mass transfers into and out of

containers which are mass controlled (AES, 2010b). Double batching is assumed to be a

credible upset condition whenever only administrative controls are used to limit the mass

5-9
(AES, 2010b). AES explicitly commits to limit the mass of uranium in any single tank in the

Technical Support Building to less than or equal to 12.2 kg uranium (AES, 2010b).

GeometryNolume

When geometry is used as a controlled parameter, subcriticality for a single unit (i.e., a

component or vessel in isolation) will be demonstrated to be independent of all other

parameters assuming 6 wt% 235U. Geometry alone is not sufficient to ensure that a particular

system is subcritical since it is generally not possible to install a single, isolated unit. Thus, AES

will also consider credible interaction and reflection conditions in its NCS determinations

(AES, 2010b). AES will verify all dimensions relied on for NCS prior to the beginning of

operations (AES, 2010b). The safe geometry values, as demonstrated in Table 5.3-1 of this

SER, provide a substantial margin of safety when geometry is used as a controlled parameter.

The applicant based this conclusion on a determination that this margin is sufficient to account

for credible manufacturing tolerances, deformation (e.g., from bulging and corrosion), and

uncertainties in the calculational methods. AES explicitly commits to implement and maintain

centrifuge, chemical trap, and product cold trap diameters and second stage UF 6 pump, uranic

liquid container, vacuum cleaner, and oil container volumes to less than the safe values in Table

5.3-1 of this SER. AES also commits to demonstrate that an isolated first stage UF6 pumps is

subcritical for 6 wt% 2 3 5

U based only upon controlling its geometry (AES, 2010b).

Enrichment

Although AES is requesting authorization to only enrich uranium up to 5 wt% 235U, it commits to

conduct all its NCS analyses, except for the contingency dump system traps and tails cylinders,

using 6 wt% 235U (AES, 2010b). The NCS analysis for the contingency dump system traps and

tails cylinders will be based upon 1.5 wt% 235U (AES, 2010b). Feed and uranium byproduct

cylinders are limited to less than 0.72 wt% 235U, which ensures they will be subcritical under all

credible conditions. In the original SAR, AES had stated that the only exception to conducting

analyses at 6 wt% 235U was for the contingency dump system traps (AES, 2009b). The staff

observed that ISA documents indicated that portions of the primary (tails take-off) dump system

and other parts of the contingency dump systems had been analyzed at 1.5 wt% 231U, which

was inconsistent with the commitment in the SAR (AES, 2010b). In response to the staff's RAI

ISA-14 (AES, 2009a), AES revised the SAR (AES, 201 Ob) to include the tails cylinders in the

exception but otherwise maintained its commitment to conduct all other NCS analyses using 6

wt% 235

u.

Reflection

AES considers full water reflection, but has determined that this is highly unlikely to occur from a

source internal to the facility (AES, 201 Ob). Partial reflection of 2.5 cm of water is assumed

where limited moderating materials may be present (AES, 2010b). This means that calculations

will assume that the outside surface of a component is covered by water 2.5 cm thick. AES

indicated (RAI NCS-2) (AES, 2009a) that using 2.5 cm of water is a common practice in NCS for

simulating room return and human presence. Where present, concrete is modeled in the

analysis since it is a better reflector than water. AES commits to NUREG-1 520, Section 5.4.3.4,

thus the controls to prevent potential reflectors will be identified as IROFS in the ISA Summary

(AES, 2010b).

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In areas where fire sprinkler systems are installed, the analyses will consider credible interstitial

moderation and sheeting assuming the system activates (AES, 2010b). In addition, standpipes

will be installed in areas where their failure will not result in flooding of areas containing enriched

uranium greater than a critical mass (AES, 2010b).

Moderation

Moderation control will be established in accordance with ANSI/ANS-8.22-1998 (NRC, 2005).

Where moderation is the only parameter used for criticality control, two independent controls will

be used to verify moderator content (AES, 201 0b). AES commits to NUREG-1 520, Section

5.4.3.4, thus when process variables can affect moderation, the process variables will be shown

in the ISA Summary to be controlled by IROFS (AES, 201 Ob). As stated previously, control of

moderation will only be applied to product cylinders (AES, 201 Ob).

Optimum moderation or the worst case credible hydrogen to uranium (H/U) ratio will be used in

NCS analyses when moderation is not controlled. AES demonstrated that an H/U ratio of 7 is

the maximum credible moderation that can result from moisture in-leakage (AES, 2010b). This

demonstration assumes the hydrogen is entirely from water molecules. Therefore, when AES

uses the worst case credible moderation (or credible optimum moderation), it will select the H/U

ratio between 0 and 7 that is most reactive for the system being analyzed A(ES, 201 Ob).

Optimum moderation is the moderating condition that is most reactive without any limits on the

H/U ratio.

UF6 and vacuum pumps will use fully fluorinated (non-hydrogenous) perfluorinated polyether

(PFPE) type lubricant (AES, 2010b). AES indicated that it will analyze these pumps with water,

instead of PFPE, which will bound all credible abnormal moderating conditions (AES, 2010b).

Fire sprinkler systems may be installed in some areas of the facility. AES will not install

sprinklers in any area where a critical mass could accumulate in an unsafe geometry (e.g.,

drains) or near sub-atmospheric process systems that could contain a critical mass and require

moderator control (AES, 2010b).

Interaction

NCS analysis and evaluations will consider the effects of interaction (RAI NCS-1 0), including intransit material, using MONK8A to ensure that kff < 0.95. Spacing requirements will be

determined separately for each system (AES, 201 Ob). AES commits to NUREG-1 520, Section

5.4.3.4, thus it will use either engineered controls or augmented administrative controls to

maintain the physical separation between different units (AES, 2010b). In this case, an

administrative control is considered to be augmented if a physical device significantly aids the

operator to ensure that the required action will be performed correctly.

AES will establish controls to maintain the spacing between items in movement and other fissile

material (AES, 2010b). AES commits to have only one item of each type (e.g., one pump and

one trap) in movement at the same time (AES, 201 Ob). The only time the criticality analysis

allows the spacing requirements to be relaxed is when vessels are moved in or out of fixed

positions.

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5.3.5.3 Findings

The staff has reviewed the NCS methodologies and technical practices and finds that they are

acceptable because the applicant has adequately addressed the acceptance criteria in NUREG1520, Sections 5.4.3.4.1, 5.4.3.4.2, 5.4.3.4.4, and 5.4.3.4.5. The applicant commits to the

double contingency principle as required by 10 CFR 70.64(a)(9) (AES, 2010b). The staff finds

that the NCS methodologies and technical practices, if applied as described in the SAR, will

provide an adequate margin of subcriticality for safety as required by 10 CFR 70.61(d).

5.3.6 CAAS

5.3.6.1 SAR Commitments

The acceptance criteria for NRC's review are contained in Section 5.4.3.4.3 of NUREG-1520

(NRC, 2002).

AES commits to maintain documents demonstrating that its CAAS meets the requirements of 10

CFR 70.24 (AES, 201 Ob). The applicant will follow ANSI/ANS-8.3-1997, "Criticality Accident

Alarm System," (ANSI/ANS, 1998) as modified by NRC RG 3.71, Revision 1 (NRC, 2005)

(AES, 201 Ob). The CAAS will be uniform throughout the facility for the type of radiation

detected and the alarm signals (AES, 2010b).

AES commits to have an emergency plan and will comply with ANSI/ANS-8.23-1997, "Nuclear

Criticality Accident Emergency Planning and Response" [Ref. 0] (AES, 201 Ob). Fixed criticality

accident dosimeters will be located within the buildings for accident reconstruction

(AES, 2010b). AES and emergency response personnel will be provided with dosimeters prior

to entering the radiological controlled areas (AES, 2010b).

The CAAS is backed up by emergency power and will be designed to remain operational during

credible events and conditions (AES, 2010b). If CAAS coverage is lost, compensatory

measures [such as limiting access and restricting movement of special nuclear material (SNM)]

will be implemented (AES, 2010b). If coverage is lost and cannot be restored within a specified

number of hours, or if an equivalent level of coverage is not provided, operations in the affected

area will be shut down or quarantined as necessary (AES, 2010b). AES will develop on-site

guidance on when an operation should be safely shut down if CAAS coverage cannot be

restored within the specified number of hours (AES, 2010b).

In the original SAR submittal, AES only committed to provide CAAS coverage to those areas

with greater than a critical mass of fissile material (AES, 201 Ob). However in response to the

staff's concern that this was not sufficient to meet 10 CFR 70.24 (RAI NCS-5) (AES, 2009a),

AES committed to provide CAAS coverage for each area where fissile material is handled,

used, or stored (AES, 201 Ob).

5.3.6.2 ISA Summary

The ISA Summary provided additional information on the planned CAAS (AES, 2010b). CAAS

coverage map was included which shows that each area at EREF where SNM will be handled,

used, or stored is covered by at least two detectors. The coverage map is based off the

planned use of gamma ray detectors with an alarm set point of 1 milligray/hour (mGy/hr). The

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proposed distances in areas with no shielding were compared with hand calculations described

in Appendix B to ANSI/ANS-8.3-1997 (ANSI/ANS, 1997), and found to be consistent. The

proposed range for CAAS detectors near autoclaves was much shorter to account for shielding.

5.3.6.3 Findings

The staff has reviewed the applicant's commitment to the CAAS requirements in 10 CFR 70.24

and finds that it is acceptable because the applicant maintains a CAAS that is capable of

energizing a clearly audible alarm signal if accidental criticality occurs, and the applicant

maintains emergency procedures for each area in which SNM is handled, used, or stored to

ensure prompt personnel evacuation upon the sounding of the alarm. Additionally, the staff

finds that the applicant has adequately addressed the acceptance criteria in NUREG-1 520,

Sections 5.4.3.4.3 and 3.4.3.2(4)(c).

5.3.7 Reporting Requirements

AES will establish a program to evaluate the significance of events, in terms of NCS, using

qualified individuals (AES, 2010b). The criteria for reporting events to the NRC Operations

Center (10 CFR 70.50 and Appendix A to 10 CFR 70) will be incorporated into the facility

emergency procedures. Reports will be made based upon IROFS failure regardless of whether

or not safety limits are exceeded. An event will be reported within one hour-of discovery, when

the reporting criteria in 10 CFR 70, Appendix A, Paragraph (a) apply, or when it cannot be

determined that these criteria do not apply.

The staff has reviewed the applicant's commitments to report NCS events to the NRC

Operations Center and finds that they are acceptable because the applicant has adequately

addressed all the acceptance criteria in NUREG-1 520, Section 5.4.3.4.7(7).

5.3.8 ISA

The acceptance criteria for the NRC's NCS review of the ISA are contained in Sections

5.4.3.4.6 and 3.4 of NUREG-1 520 (NRC, 2002).

The purpose of the ISA review is to determine that there is reasonable assurance that an ISA is

being conducted in accordance with 10 CFR 70.62(c) which will ensure that the requirements of

10 CFR 70.61, as they relate to NCS, will be met. This review is divided into two parts: (1)

review of commitments regarding the ISA and ISA Summary, and (2) review of the ISA

Summary. The conclusions reached in this section are supported by an on-site review of ISA

documents. The non-public portions of the ISA Summary and on-site ISA document review

(horizontal and vertical slice) are discussed in Appendix A.

5.3.8.1 ISA Commitments

The acceptance criteria for the NRC's review of the NCS ISA commitments are contained in

Sections 3.4.3.1.

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ISA Methodology

The ISA conducted to support the design and application was performed by two teams - one for

classified processes and one for unclassified processes. Some ISA team members are on both

teams to ensure consistency. AES commits to have ISA teams which include at least one

member with expertise in NCS that is appropriately trained and qualified as described in Section

5.3.2. During the operations phase, the NCS manager will be responsible for providing NCS

support for the ISA (AES, 201 Ob).

The ISA teams use the hazard and operability (HAZOP) method to identify process hazards

(AES, 2010b). Each process is divided into nodes. Guidewords are used to identify potential

hazards associated with each node. A representative list of guidewords is found in Table 3.1-1

of the SAR (AES, 201 Ob). For each hazard, the ISA team will identify potential causes,

consequences, and safeguards or design features which may prevent or mitigate the hazard.

During the on-site review, AES stated (Ref. 0) that these safeguards and design features are

not credited when evaluating the risk for the uncontrolled situation. Some of these safeguards

and design features are credited when evaluating the controlled accident sequence, in which

case they are designated as IROFS (AES, 2010b).

AES considers criticality accidents to be high consequence events per 10 CFR 70.61(b), and

thus must be determined to be either not credible or highly unlikely. An event is considered to

be credible unless it is determined to meet one of the following criteria:

1. An external event with a frequency of occurrence of less than once in a million years.

2. A process deviation that consists of a sequence of many unlikely human actions or

errors for which there is no reason or motive.

3. A process deviation for which there is a convincing argument based on physical laws

that it is not possible or unquestionably extremely unlikely.

In applying these criteria, AES will not rely on any control or design feature that could credibly

fail to function.

For each credible criticality accident, AES will demonstrate that the accident is highly unlikely as

defined in NUREG-1520 (NRC, 2002). An event is highly unlikely if the frequency of occurrence

is less than 10.5 events per year (based on an approximate order of magnitude). IROFS are

required to meet this criterion unless the uncontrolled event is highly unlikely. More details on

the determination of event frequencies and IROFS failure rates are found in Chapter 3 of this

SER.

The staff concludes that the method used for the HAZOP analysis is sufficient to ensure that all

credible conditions which could impact NCS are considered.

Safe-by-Design IROFS

The SAR defines a special class of NCS controls as safe-by-design IROFS (AES, 2010b). A

safe-by-design IROFS is a passive engineered control which by its geometry and configuration

alone will prevent a criticality accident from occurring (AES, 201 Ob). This means that all NCS

parameters, except geometry, enrichment, and interaction, are assumed to be in the optimum or

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worst case credible condition. By definition, it must be highly unlikely for a safe-by-design

IROFS to fail in a manner that would cause a criticality accident (AES, 2010b). The most

significant requirements to qualify as a safe-by-design IROFS are that:

1. The only credible failure mechanism that could result in a criticality would be to implement

an improper design change.

2. Credible process deviations or events do not adversely impact performance of the safety

function.

3. Quality Assurance Level 1 is applied to the feature.

4. No human actions are required for the component to perform its safety function.

The applicant considers failure of a safe-by-design IROFS due to a loss of configuration control

to be highly unlikely (AES, 201 Gb). The staff determined that such an event could lead to

criticality if an improperly modified safe-by-design IROFS were installed, used, and if an

adequate margin of subcriticality (i.e., keff < 0.95) were not maintained for all credible process

deviations or events. The staff determined, based on commitments made in the SAR (AES,

2010b), that such a failure would require several mistakes to be made in the configuration

management process, which is sufficient to justify that this event is highly unlikely. Thus, the

staff concludes that the configuration management program will ensure that failure or improper

modification of safe-by-design IROFS will be highly unlikely, in accordance with 10 CFR 70.61.

A HAZOP analysis is performed by the ISA team on each safe-by-design IROFS to support the

conclusion that it is not credible for a process deviation or other event to adversely impact the

safety function. The primary focus is on those events which could either change the geometry

or configuration (i.e., physical arrangement and spacing) of the component. The information

provided in Appendices B and C of the ISA Summary (AES, 2010b) indicates that AES will

consider, as appropriate, the worst case credible process deviations and events. Process

deviations considered in the HAZOP analysis include pressure changes, temperature changes,

corrosion, erosion, bulging, leakage, and rupture. Other events considered include impacts,

construction activities, maintenance activities, fires, earthquakes, flooding, and loss of utilities

(AES, 2010b).

The analysis of safe-by-design components assumes the operator is using the required

component and mobile vessels are adequately controlled. Events which could cause one of

these assumptions to be violated are considered to be credible and are analyzed separately in

the ISA.

Safe-by-design IROFS must also have significant margin, as defined in the SAR (AES, 201Gb).

Significant margin is defined as follows: (1) for components based on safe-by-design attributes

of diameter, volume, or slab thickness, a margin of at least 10% during normal and credible

abnormal conditions, and (2) for components based on a calculation, keff < 0.95. The staff

determined that AES's application of significant margin does not provide any additional benefit

to safety, in that it does not provide any actual safety margin beyond what is required for other

passive controls, and it is therefore not relied upon to make any conclusions for the staff's

review.

Based on the above, the staff concludes that components that meet the safe-by-design criteria

will be capable of (1) preventing a criticality accident under normal and credible abnormal

conditions, (2) maintaining an adequate margin of subcriticality for safety, and (3) ensuring that

a criticality accident is highly unlikely to occur.

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5.3.8.2 ISA Summary Review

The acceptance criteria for the NRC's NCS review of the ISA Summary are contained in Section

3.4.3.2 of NUREG-1520 (NRC, 2002).

The purpose of the ISA Summary review is to verify that the applicant complied with the

requirements of 10 CFR 70.65(b) as it relates to NCS. The acceptance criteria (NRC, 2002) for

the ISA Summary regarding completeness and description of processes, accident sequences,

and IROFS is based upon the current stage of the design and not on the final design.

The ISA Summary requirements which are applicable to NCS are 10 CFR 70.65(b)(3) - (9).

Since the ISA Summary contains proprietary, security-related, and export controlled information,

the review of information required to meet 10 CFR 70.65(b)(3), (4), (6), and (8) is discussed in

Appendix A and summarized in this Section. Review of CAAS information required by

70.65(b)(4) is discussed in Section 5.3.6 of this SER. The requirements of 70.65(b)(5) and (9)

are redundant with the commitments discussed in Section 5.3.8.1.

The applicant was unable to provide classified information during the review. The ISA Summary

does not have a classified section; some safety analyses which support the ISA Summary are

classified. The applicant was able to provide the reviewer with an unclassified summary of this

analysis, when it was requested.

10 CFR 70.65(b)(3): Process Description

The staff has reviewed each of the process descriptions in the ISA Summary and finds them

acceptable because they adequately address the acceptance criteria in NUREG-1 520, Section

3.4.3.2(3) as it relates to NCS. Specifically, the applicant provided sufficient information in the

ISA Summary to determine where criticality hazards exist and how operations might impact

NCS.

10 CFR 70.65(b)(4): Performance Requirements

The regulation in 10 CFR 70.61(d) requires that the risk of nuclear criticality accidents be limited

by assuring that under normal and credible abnormal conditions, all nuclear processes are

subcritical. At the proposed EREF, a criticality hazard will exist only in those areas where

enriched uranium is present under normal or credible abnormal conditions (AES, 201 Ob). For

the uranium enrichment levels that will be present at the proposed EREF, a moderator is also

required for a criticality to occur. Common moderators include water, hydrogen fluoride (HF),

and hydrocarbon oils or lubricants. At low enriched fuel facilities like the proposed EREF,

operations where a moderator is intentionally added to enriched uranium require the most NCS

scrutiny. However, possible external sources of a moderator must be considered as a credible

abnormal condition, in accordance with 10 CFR 70.61(d), in the NCS analysis for all operations.

At the proposed EREF, the only operations where significant moderation of fissile material will

normally occur are in the decontamination workshop and the liquid effluent collection and

treatment system (AES, 2010b).

Large portions of the proposed facility rely upon safe-by-design attributes to ensure that

processes are subcritical under normal and credible abnormal conditions (AES, 2010b). The

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staff determined that subcriticality in equipment with safe-by-design attributes relies upon

controlling equipment geometry (i.e., one or more physical dimensions), enrichment, and

interaction. All other parameters are assumed to be in the worst case credible condition for

NCS purposes. 10 CFR 70.61(e) requires that controls on parameters needed to ensure

subcriticality be designated as IROFS. The staff determined during its ISA Summary review

that IROFS have been designated which will ensure that the enrichment limits will not be

exceeded for any the safe-by-design equipment. The staff also determined that geometry

control is part of the safe-by-design IROFS designations. The staff found that some interaction

controls were part of a safe-by-design IROFS, and that some were captured by administrative

IROFS. The staff did not assess whether all interaction controls had been designated as IROFS

because the staff's review of the ISA Summary was on a sampling basis.

AES has identified loss of configuration control as an initiating event that could cause a safe-bydesign IROFS to fail (AES, 2010b). The staff notes that configuration management is a

management measure that is applied to all IROFS to ensure their availability and reliability.

Typically, failure of configuration management is not identified as a separate accident initiator in

the ISA since it is already factored into the availability and reliability of the IROFS. Therefore,

consideration of its failure is already covered in determining that failure of safe-by-design

components is highly unlikely. The applicant has made commitments regarding its configuration

management program (some specific to safe-by-design IROFS) in its SAR (AES, 2010b). Upon

review, the staff determined that those commitments are sufficient to ensure that the failure of

these safe-by-design IROFS is at least highly unlikely. Thus, the staff determined that it is not

necessary for AES to specifically identify failure of configuration management as an accident

sequence.

Because AES considered improper design changes as separate accident sequences, it had

identified safe-by-design IROFS as sole IROFS. However, the staff disagrees, based on the

determination that such failures of configuration management are not required to be considered

as separate accident sequences. In addition, in sequences involving the failure of configuration

management, it is the configuration management program that ensures safety, and not the safeby-design IROFS itself. Designating these controls as sole IROFS is conservative, so this does

not constitute a safety or regulatory concern.

The staff has reviewed the information provided in the ISA Summary and supporting ISA

documents demonstrating compliance with the performance requirements of 10 CFR 70.61 and

finds it is acceptable because it adequately addresses the acceptance criteria in NUREG-1520,

Sections 3.4.3.2(4)(a) and (b) as they relate to NCS. Specifically, the staff has concludes that

for the current state of facility design (1) each process has been evaluated for NCS, (2) criticality

accidents will be at least highly unlikely, and (3) each process will be adequately subcritical

under normal and credible abnormal conditions.

10 CFR 70.65(b)(4): Baseline Design Criteria and Defense-in-Depth

As part of the defense-in-depth requirements, the applicant must show a preference for

engineered over administrative controls in the design of the facility. The widespread application

of safe-by-design IROFS demonstrates this preference since these are passive engineered

controls that require no human actions to prevent a criticality accident.

Another part of the defense-in-depth requirements is that the facility design should incorporate

features which reduce challenges to IROFS. For many reasons, the normal facility operations

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requirements greatly reduce the challenges to IROFS for most operations. The parameter

values at which the safe-by-design components are analyzed represent very conservative

process conditions which are unlikely to ever be approached. For example, the feed, product,

and main processes all use UF6, which necessarily requires that moderator be excluded.

Without a moderator, a criticality accident is not possible for the proposed enrichment limit. This

ensures that the geometry control associated with safe-by-design components will seldom be

relied on to ensure subcriticality. The staff has reviewed the information provided in the SAR

and ISA Summary as it relates to the defense-in-depth practices required by 10 CFR 70.64(b).

The staff finds this information is acceptable because it adequately addresses the acceptance

criteria in NUREG-1520, Sections 3.4.3.2(4)(d) as it relates to NCS. Specifically, the NRC staff

concludes that the applicant has designed the process with a preference for engineered controls

and will operate it in a manner which limits the challenges to NCS IROFS.

10 CFR 70.65(b)(6) and (8): IROFS and Sole IROFS

The ISA Summary identifies four generic safe-by-design IROFS. The first three are for

components where NCS is assured because the diameter, slab thickness, or volume of the

component is less than the safe geometry values for 6 wt% 235U (SAR Table 5.1-1). The other

safe-by-design IROFS is for components which are demonstrated to be subcritical with an

explicit NCS analysis (AES, 2010b). Appendix B and C of the ISA Summary describes the

specific components that will be categorized as safe-by-design IROFS.

The safe-by-design IROFS have been designated as sole IROFS, because they are the only

IROFS listed for a criticality accident initiated by a loss of configuration control. The staff

determined that the loss of configuration control is not required to be listed as an accident

initiator in the ISA summary, for reasons stated above. Therefore, the staff does not concur that

safe-by-design IROFS are necessarily sole IROFS. They are not sole IROFS if the only basis

for calling them sole IROFS is that a hypothetical design change could theoretically result in

criticality. Nevertheless, the staff finds that there is no safety or regulatory concern with the

applicant's conservatively designating them as sole IROFS, since this would result in a higher

level of regulatory oversight. The ISA Summary identifies several enhanced administrative

controls which prevent criticality accidents (AES, 201 Ob). Each of these IROFS requires an

independent verification, which means that a second qualified individual will independently

ensure that the required task is performed correctly. A failure probability index of -3 is assigned

to these IROFS, which is consistent with what has been approved for other facilities.

The ISA Summary identifies six enhanced administrative controls as sole IROFS preventing

certain criticality accident sequences (AES, 2010b). For five out of these six sole IROFS,

several repeat failures would have to occur before a criticality is possible. The remaining sole

IROFS is sufficient, along with an initiating event with a frequency index of at least -2, for the

sequence to meet the criteria for demonstrating that an event is highly unlikely. (The applicant's

risk index methodology is discussed and evaluated in Chapter 3 of this SER.)

The staff has reviewed the descriptive lists of NCS IROFS and sole IROFS in the ISA Summary

and finds them acceptable because they adequately address the acceptance criteria in NUREG1520, Sections 3.4.3.2(6) and (8) as it relates to NCS. For each criticality accident sequence

identified in the ISA Summary, the list included the IROFS needed to render the sequence

highly unlikely. The descriptions were sufficient to understand how the IROFS would prevent a

criticality from occurring.

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5.3.8.3 Conclusion

The staff has reviewed the EREF ISA methodology and ISA Summary and has reasonable

assurance that the applicant has conducted an ISA, based upon the current level of design,

that:

(1) Identified all credible criticality accident scenarios.

(2) Identified controls to prevent each credible criticality accident sufficient to meet the

performance requirements of 10 CFR 70.61(b) and (d) and provide for double

contingency protection and defense-in-depth as required by 10 CFR 70.64(a)(9) and

70.64(b).

(3) Designated as IROFS, as required by 10 CFR 70.61(e), the controls relied on to meet

the performance requirements as they relate to NCS.

(4) Identified management measures which will be applied to NCS-related IROFS to ensure

they will be available and reliable.

Based upon the review of the ISA commitments, ISA Summary, and on-site ISA documents, the

staff concludes that an ISA will be performed and maintained in accordance with 10 CFR

70.62(c) such that the requirements of 10 CFR 70.61, as they relate to NCS, will be met. The

staff concludes that the ISA methodology committed to in the SAR is sufficient to identify all

credible nuclear criticality accidents and the IROFS necessary to make such events highly

unlikely.

5.3.9 Facility Changes and Configuration Management

Configuration management will be provided during all phases of facility operation (AES, 201 Ob).

In accordance with approved procedures, design changes undergo formal review which will

include assessing impacts to the ISA. As discussed in Section 5.2.1.6 of the SAR, any change

to the facility or activities of personnel that may impact the handling, use, or storage of uranium

requires an NCS evaluation and, if necessary, an NCS analysis to be prepared by an NCS

engineer and approved by an NCS manager. Prior to implementing a change it must be

determined that the entire process is subcritical under normal and credible abnormal conditions.

As part of configuration management, the applicant must assess whether NRC approval is

required before a change is implemented. Changes to the site, structures, processes, systems,

equipment, components, computer systems, and activities of personnel may be changed

without prior NRC approval if the criteria in 10 CFR 70.72(c) are met.

NRC approval is required before AES can make a change which alters a sole IROFS identified

in the ISA Summary that is necessary to prevent a criticality accident (10 CFR 70.72(c)(3)).

Altering a sole IROFS consists of making a change that modifies, positively or negatively, any of

the attributes associated with the safety function of the IROFS.

NRC staff has determined that the SAR contains the information and commitments required by

10 CFR 70.22(a) and 70.65(a) to be included as part of the license application. In addition,

10 CFR 70.61(d) requires that the margin of subcriticality for safety to be approved by the NRC.

The information and commitments regarding NCS methodologies and technical practices define

the margin of subcriticality that AES will use. Other NCS, ISA, and management measure

program information and commitments are necessary to ensure that the margin of subcriticality

will be maintained. As discussed in Section 1.2.4.2.2 of this SER, the applicant has requested

authorization to make changes to license commitments. The authorization has two parts

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covering changes requiring prior approval and changes not requiring prior approval. Changes

to the license application that will require prior NRC approval are those that decrease the

effectiveness of any safety commitments. For a change that decreases the effectiveness of any

safety commitment, the applicant will submit the change to the NRC as a license amendment

and will not implement the change until NRC has reviewed and approved the amendment.

Changes to the license application that could be made without prior NRC approval are those for

which there is no degradation in the safety commitments in the license and for which the

change, test, or activity does not conflict with any condition specifically stated in the license.

The staff will impose the license condition stated in Section 1.2.4.2.2 of this SER so that

commitments in the SAR, including those required by 10 CFR 70.22(a) and 70.65(a), cannot be

changed except by license amendment.

The staff has reviewed the applicant's commitments for configuration management as it relates

to NCS and finds that it is acceptable because the applicant has adequately addressed the

acceptance criteria in NUREG-1 520, Section 5.4.3.4.7. In addition, the staff finds the

applicant's method for assessing changes to safe-by-design IROFS to determine if they are

alterations per 10 CFR 70.72(c)(3) is acceptable. The staff does not find the applicant's initial

proposal to maintain the SAR as a "living document" under the configuration management to be

acceptable. Instead, the staff will impose a license condition covering changes requiring prior

NRC approval and changes not requiring prior approval to ensure that commitments in the SAR,

including those required by 10 CFR 70.22(a) and 70.65(a), cannot be changed except by

license amendment. This license condition is provided in Section 1.2.4.2.2 of this SER.

5.4 Evaluation Findings

The staff has reviewed the SAR, ISA Summary, and on-site ISA documents for the proposed

EREF as it relates to NCS. Provided that the applicant implements and maintains its safety

programs as described in the SAR, the staff has reasonable assurance that:

(1) The applicant will have a staff of managers, supervisors, engineers, process operators,

and other support personnel who are qualified to develop, implement, and maintain the

NCS program.

(2) The applicant's conduct of operations will be based on NCS methodologies and

technical practices, which will ensure that fissile material will be possessed, stored, and

used safely according to the requirements in 10 CFR Part 70.

(3) The applicant's NCS methodologies and technical practices will provide an adequate

margin of subcriticality for safety as required by 10 CFR 70.61(d).

(4) The applicant's safety programs and management measures will ensure that the margin

of subcriticality is maintained such that processes will meet the subcriticality

requirements of 10 CFR 70.61(d) and the requirements for new facilities specified in 10

CFR 70.64.

(5) The applicant will develop, implement, and maintain a CAAS in accordance with the

requirements in 10 CFR 70.24 and the facility emergency management program.

(6) The applicant will perform and maintain an ISA in accordance with 10 CFR 70.62(c)

such that the requirements of 10 CFR 70.61, as they relate to NCS, will be met.

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NRC staff has determined that the SAR contains commitments required by 10 CFR 70.22(a)

and 70.65(a) and needed to define the approved margin of subcriticality to be included as part

of the license application. The staff will impose the license condition provided in

Section 1.2.4.2.2 of this SER so that commitments cannot be changed except by license

amendment.

The NRC staff concludes that if the proposed EREF NCS program is implemented in

accordance with the statements made in the SAR it will 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.

5.5 References

(AES, 2010a) AREVA Enrichment Services LLC, Eagle Rock Enrichment Facility,

"Supplemental Response to Requests for Additional Information," May 12, 2010.

(AES, 2010b) AREVA Enrichment Services, Eagle Rock Enrichment Facility, "Revision 2 to

License Application for the Eagle Rock Enrichment Facility," April 30, 2010.

(AES, 2009a) AREVA Enrichment Services, Eagle Rock Enrichment Facility, "Response to

Requests for Additional Information-AREVA Enrichment Services LLC License Application for

the Eagle Rock Enrichment Facility," September 28, 2009.

(AES, 2009b) AREVA Enrichment Services, Eagle Rock Enrichment Facility, "Revision 1 to

License Application for the Eagle Rock Enrichment Facility," April 23, 2009.

(AES, 2008a) AREVA Enrichment Services, Eagle Rock Enrichment Facility, "Application for a

Uranium Enrichment Facility," Rev. 0, December 30, 2008.

(AES, 2008b) AREVA Enrichment Services, Eagle Rock Enrichment Facility, "MONK8A

Validation and Verification," December 2008.

(AES, undated) AREVA Enrichment Services, Eagle Rock Enrichment Facility, "MONK8A

Validation and Verification," Rev. 2, undated.

(ANSI/ANS, 2005) American National Standards Institute/American Nuclear Society,

ANSI/ANS-8.19-2005, "Administrative Practices for Nuclear Criticality Safety."

(ANSI/ANS, 1998) American National Standards Institute/American Nuclear Society,

ANSI/ANS-8.1-1998, "Nuclear Criticality Safety in Operations with Fissionable Material Outside

Reactors."

(ANSI/ANS, 1997a) American National Standards Institute/American Nuclear Society,

ANSI/ANS-8.3-1997, "Criticality Accident Alarm System."

(ANSI/ANS, 1997b) American National Standards Institute/American Nuclear Society,

ANSI/ANS-8.22-1997, "Nuclear Criticality Safety Training Based on Limiting and Controlling

Moderators."

(ANSI/ANS, 1997c) American National Standards Institute/American Nuclear Society,

ANSI/ANS-8.23-1997, "Nuclear Criticality Accident Emergency Planning and Response."

(ANSI/ANS, 1991) American National Standards Institute/American Nuclear Society,

ANSI/ANS-8.20-1991, "Nuclear Criticality Safety Training"

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(ANSI/ANS, 1995) American National Standards Institute/American Nuclear Society,

ANSI/ANS-8.21-1995, "Use of Fixed Neutron Absorbers in Nuclear Facilities Outside Reactors."

(ANSI/ANS, 1983) American National Standards Institute/American Nuclear Society,

ANSI/ANS-8.10-1983, "Criteria for Nuclear Criticality Safety Controls in Operations with

Shielding and Confinement."

(NRC, 2010a) U.S. Nuclear Regulatory Commission, Teleconference Summary, February 25,

2010.

(NRC, 2010b) U.S. Nuclear Regulatory Commission, Teleconference Summary, March 8, 2010.

(NRC, 2010c) U.S. Nuclear Regulatory Commission, Draft Regulatory Guide DG-3037,

"Guidance for Fuel Cycle Facility Change Processes," June 2009.

(NRC, 2009) U.S. Nuclear Regulatory Commission, July 13-15, 2009, Meeting Summary,

"AREVA Enrichment Services Eagle Rock License Application - Integrated Safety Analysis InOffice Review," 2009.

(NRC, 2006) U.S. Nuclear Regulatory Commission, FCSS-ISG-10, "Justification for Minimum

Margin of Subcriticality for Safety," 2006.

(NRC, 2005) U.S. Nuclear Regulatory Commission, Regulatory Guide 3.71, "Nuclear Criticality

Safety Standards for Fuels and Material Facilities," Rev. 1, 2005.

(NRC, 2002) U.S. Nuclear Regulatory Commission, NUREG-1 520, "Standard Review Plan for

the Review of a License Application for a Fuel Cycle Facility," March 2002.

(NRC, 2001) U.S. Nuclear Regulatory Commission, NUREG/CR-6698, "Guide for Validation of

Nuclear Criticality Safety Calculational Methodology," January 2001.

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CHAPTER 6.0 CHEMICAL PROCESS SAFETY

The purpose of the U.S. Nuclear Regulatory Commission's (NRC's) review of the applicant's

chemical safety program and the design of the proposed Eagle Rock Enrichment Facility

(EREF) is to evaluate whether the applicant's chemical process safety program will adequately

protect workers, the public, and the environment during normal operations against chemical

hazards of licensed material and its byproducts. The chemical process safety program and the

facility's design must also protect against facility conditions and/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 Title 10 of the Code of Federal

Regulations (10 CFR) 70.22 and 70.65. In addition, the chemical process safety review should

provide reasonable assurance of compliance with 10 CFR 70.61, 70.62, and 70.64.

6.2 Regulatory Guidance and Acceptance Criteria

The guidance applicable to NRC's review of chemical process safety for the proposed facility is

contained in Chapter 6 of "Standard Review Plan for the Review of a License Application for a

Fuel Cycle Facility," NUREG-1520 (NRC, 2002). This chapter is applicable in its entirety. The

staff also uses the following as guidance documents for this review: NUREG-1601, "Chemical

Process Safety at Fuel Cycle Facilities," (NRC, 1997a), NUREG-1513, "Integrated Safety

Analysis Guidance Document," (NRC, 2001). The acceptance criteria applicable to this review

are contained in Section 6.4.3 of NUREG-1 520 (NRC, 2002).

6.3 Staff Review and Analysis

The NRC staff reviewed the Safety Analysis Report (SAR) (AES, 2009a) and the Integrated

Safety Analysis (ISA) Summary (AES, 2009b), submitted by the applicant and considered the

following areas:

1. Process Description;

2. Chemical Accident Sequences;

3. Chemical Accident Consequences;

4. Chemical Process items relied on for safety (IROFS)

5. Management Measures;

6. Emergency Management; and

7. Baseline Design Criteria (BDC).

The staff reviewed the applicant's responses to requests for additional information and the ISA

documents during an in-office review (NRC, 2009), 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. Compliance with these regulations is

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discussed in more detail in Chapter 3, Appendix A, and Appendix B of this Safety Evaluation

Report (SER). The staff's evaluation and general information about the proposed EREF

process are summarized in the following sections.

6.3.1 Process Description

The applicant describes the gas centrifuge process in Section 1.1.3 of the SAR (AES, 2009a).

The process is entirely physical in nature and mechanically separates the uranium isotopes

using a fast rotating cylinder (centrifuge). This separation occurs because there is a difference

in centrifugal forces between the isotopes of uranium since they have different molecular

weights. Neither nuclear reactions nor significant chemical changes are expected to occur

during normal operations. The feed, product, and tails streams are all in the form of uranium

hexafluoride (UF 6).

6.3.1.1 Gas Centrifuge Process

The enrichment process proposed by the applicant, housed in the Separations Building, is

comprised of four major systems: a UF 6 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 the Product Blending System. Supporting functions include

sample analysis, equipment decontamination and rebuild, liquid effluent treatment, and solid

waste management (AES, 2009b).

The major equipment used in the UF6 Feed System are the Solid Feed Stations. UF6 will be

delivered to the proposed EREF in the American Nuclear Standards Institute (ANSI) N14.1

(ANSI, 1995) standard 48Y international transit cylinders (AES, 2009a). Feed cylinders will be

loaded into Solid Feed Stations; vented for removal of light gases, primarily air and hydrogen

fluoride (HF); and heated air will be circulated around the feed cylinder to sublime the UF6 .

The light gases and UF6 gas generated during feed purification will be routed to the Feed

Purification Subsystem (AES, 2009b). The major pieces of equipment in the Feed Purification

Subsystem will be UF 6 Cold Traps, a Vacuum Pump/Chemical Trap Set, and a LowTemperature Take-off Station (LTTS) (AES, 2009b). The Feed Purification Subsystem will

remove any light gases, such as air and HF from the UF6 feed prior to introduction into the

cascades (AES, 2009b). The UF6 will be captured on UF6 Cold Traps and ultimately recycled

as feed, while HF will be captured on chemical traps (AES, 2009b).

After purification, the gaseous UF6 will be 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. Therefore, the

centrifuges are grouped in series and in parallel to form arrays known as cascades. A typical

cascade is compromised 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 leave the centrifuge through the central post. Control valves, restriction

orifices, and controllers provide uniform flow of product and tails.

Depleted UF6 exiting the cascades will be transported from the high vacuum of the centrifuge for

desublimation into cylinders at subatmospheric pressure. Tails material will be desublimed into

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48Y cylinders. The primary equipment of the Tails Take-off System is the vacuum pump and

the Tails LTTS. Chilled air will flow over cylinders in the Tails LTTS to effect desublimation.

Filling of the cylinders will be monitored with a load cell system, and filled cylinders will be

transferred outdoors to the Full Tails Cylinder Storage Pad.

Enriched UF6 from the cascades will be desublimed in a Product Take-off System comprised of

vacuum pumps, Product LTTS, UF 6 Cold Traps, and Vacuum Pumps/Chemical Trap Sets (AES,

2009b). Product material will be desublimed into 48Y or smaller 30B cylinders (AES, 2009b).

The vacuum pumps will transport the UF6 from the cascades to the Product LTTS at

subatmospheric pressure (AES, 2009b). The heat of desublimation of the UF 6 will be removed

by cooling air routed through the LTTS (AES, 2009b). The product stream normally contains

small amounts of light gases that may have passed through the centrifuges. Therefore, a UF6

Cold Trap and Vacuum Pump/Chemical Trap Sets will be provided to vent these gases from the

product cylinder (AES, 2009b). Any UF6 captured in the cold trap will be periodically transferred

to another product cylinder for use as product or blending stock. Filling of the product cylinders

will be monitored with a load cell system, and filled cylinders will be transferred to the Product

Liquid Sampling System for sampling.

Sampling will be performed to verify product assay level (weight percentage 235U) (AES, 2009b).

The Product Liquid Sampling Autoclave is an electrically heated, closed pressure vessel used to

liquefy and allow collection of a sample (AES, 2009b). After the UF 6 is liquefied, the contents of

the product cylinder will be allowed to homogenize for a period of time. The autoclave will be

fitted with a hydraulic tilting mechanism that elevates one end of the autoclave so that liquid UF6

will pour into a sampling manifold connected to the cylinder valve. After sampling, the autoclave

will be brought back to the horizontal position; and the autoclave and cylinder will be cooled

down by a chiller unit mounted on the interior of the pressure vessel with the refrigerant

compression and heat rejection components on the exterior.

The facility will have the capability to blend enriched UF 6 from donor cylinders of different

assays into a product receiver cylinder. The Product Blending System will be comprised of

Blending Donor Stations for two donor cylinders and Blending Receiver Stations for the receiver

cylinder. The Donor Stations are similar to the Solid Feed Stations and the Receiver Stations

are similar to the LTTS. The Solid Feed Stations and the LTTS are described above.

The entire enrichment and product blending processes will operate at subatmospheric pressure

with the exception of product liquid sampling operations. This safety feature helps ensure that

releases of UF6 or HF are minimized because leakage would be typically inward into the

system. During sampling operations, the UF6 will be liquefied within an autoclave. The

autoclave is an American Society of Mechanical Engineers (ASME), Section VIII, Division I,

"Boiler and Pressure Vessel Code" rated pressure vessel that serves as a secondary

containment for the UF6 product cylinder while the UF 6 is in a liquid state.

6.3.1.2 Chemical Screening and Classification

The applicant classifies all site chemicals into one of three categories: Chemicals of Concern

(EREF Class 1), Interaction Chemicals (EREF Class 2), or Incidental Chemicals (EREF

Class 3).

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Chemicals of Concern (EREF Class 1)

The applicant determined the Chemicals of Concern (EREF Class 1) based on one or more

characteristics of the chemical and/or the quantity in storage/use at the facility. For licensed

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). UF6 is the only licensed material-related chemical of concern

(EREF Class 1) that will be used at the facility.

Chemicals of concern that are not related to licensed materials are those that are listed and

handled above threshold quantities set forth by either of the following standards: 29 CFR

1910.119, "Occupational Safety and Health Administration (OSHA) Process Safety

Management," or 40 CFR Part 68, "U.S. Environmental Protection Agency (EPA) Risk

Management Program."

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.

The applicant states in Section 6.1.1.1 of the SAR (AES, 2009a) that there are no non-licensed

chemicals of concern at the facility.

Interaction Chemicals (EREF Class 2)

Interaction chemicals (EREF Class 2) are those chemicals/chemical systems that require

evaluation for the potential to precipitate or propagate accidents in chemical of concern

(EREF Class 1) systems, but by themselves are not chemicals of concern. The EREF Class 2

chemicals are listed below (AES, 2009a):

* Perfluorinated polyether (PFPE) oil

" Activated carbon (C)

* Alumimum oxide (A120 3)

* Sodium fluoride (NaF)

* Citric acid (C6H80 4)

* Nitrogen (N2)

* Polydimethylsiloxane (silicone oil)

Incidental Chemicals (EREF Class 3)

The facility will use other chemicals that are neither chemicals of concern nor interaction

chemicals. Some of these incidental chemicals (EREF 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 which, via their nature, quantity, and/or use have

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no potential for impacting chemicals of concern (EREF Class 1). EREF Class 3 chemicals are

listed below (AES, 2009a), but do not impact the performance requirements of 10 CFR 70.61.

* Paper, polymers

* Potassium hydroxide (KOH)

" Phosphate

* Scrap metals

* Sodium hydroxide (NaOH)

* Hydrocarbon oils / greases

* Hydrocarbon sludges

* Methylene chloride (CH2CI2)

* Hydrocarbon I polar solvents and liquids

* Nitric acid (HNO 3)

* Hydrofluoric acid (HF(aq))

" Hydrogen peroxide (H20 2)

" Sulfuric acid (H2SO4)

• Phosphoric acid (H3PO4)

* Diesel fuel

* Deionized water (H20)

* Hydrofluorocarbons

* Propane (C3H8)

* Hydrogen (H2)

* Acetylene (C2H2)

* Oxygen (02)

* Argon (Ar)

* Helium (He)

6.3.1.3 Hazardous Chemicals and Chemical Interactions

The only chemical present in significant quantities in the facility is UF6, and it constitutes the

main hazard in the EREF. Any UF6 that is released to the environment will react exothermically

with moisture in the air, producing solid uranyl fluoride (U0 2F2) and HF gas. The reaction of

gaseous UF6 with water vapor at elevated temperatures is shown in Equation 1.

UF 6 (gas) + 2 H20(vapor) -) U0 2 F 2 (solid) + 4HF(gas) + heat (Eq. 1)

At room temperature, depending on the relative humidity of the air, the products of this reaction

are U0 2F2, hydrates and HF-H20 fog, which will be seen as a white cloud. A typical reaction

with excess water is given in Equation 2.

UF6(gas) + ( 2 +4x)H20(vapor) -> U0 2F2 * 2H20 (solid) + 4HF * xH20(fog) + heat (Eq. 2)

These reactions, if occurring in the gaseous phase at ambient or higher temperatures, are very

rapid, near instantaneous. Usually, U0 2F2 compounds are deposited or precipitated close to

the point of the release.

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6.3.1.4 UF6 and Interaction Chemicals

The following discussion describes the potential interactions between UF6 and EREF Class 2

chemicals, halocarbon refrigerants, and centrifuge cooling water.

Perfluorinated Polvether (PFPE) Oil

Besides the interaction with moisture in the air, UF 6 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. Fluorinated hydrocarbons do not react with UF6, thus the applicant will

not be using non-fluorinated hydrocarbon lubricants in any UF 6 system at the EREF.

The UF6 vacuum pumps are lubricated using PFPE oil. PFPE 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

PFPE 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.

The used PFPE oil will not be recovered for reuse at the proposed EREF. Instead, the used

PFPE oil will be collected, packaged, and shipped offsite for disposal at a licensed, low-level

radioactive waste facility (AES, 2009b).

Activated Carbon, Aluminum Oxide, and Sodium Fluoride

The applicant will use chemisorption for the removal of UF6 and HF from gaseous effluent

streams. Chemisorption occurs when a gas is captured on the surface of an activated solid

caused by the chemical reaction between the gas and the activated solid. The applicant also

will use chemisorption to remove oil mist from vacuum pumps operating upstream gaseous

effluents ventilation systems. The applicant places absorbent materials on stationery 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 (AES, 2009a).

The applicant will be using two different types of traps. The first type of trap contains a charge

of activated carbon to capture small amounts of UF6 that escape desublimation (AES, 2009a).

A second type of trap is used to absorb HF, since HF is not fully absorbed on carbon at low

pressure (AES, 2009a). The second type of trap contains a charge of aluminum oxide (A120 3)

(AES, 2009a).

Activated carbon cannot be used in the 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) will be installed in the contingency dump flow path to trap UF6. NaF will

be used because the heat of UF6 chemisorption on NaF is significantly lower than the heat of

UF6 chemisorption on activated carbon.

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Citric Acid

The applicant will use citric acid to decontaminate components (e.g., pumps, valves, and

piping), sample bottles, and flexible connectors from residual U0 2F2 compound layers that are

present in the surfaces once these are removed from the process areas (AES, 2009a). The

reaction of the uranium compounds with the citric acid solution produces various uranyl citrate

complexes. The applicant will use personnel protective features for the safe handling of

decontamination chemicals and byproducts.

NitroQen

The applicant will use gaseous nitrogen in the UF 6 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 (AES, 2009a).

Nitrogen is not reactive with UF6 in any operational condition at the facility.

Silicone Oil

The applicant will use silicone oil as a heat exchange medium for heating/chilling of various cold

traps. In all cases, this oil will be external to the UF 6 process stream and is not expected to

interact with UF6 (AES, 2009a).

Halocarbon Refrigerants

The applicant will use halocarbon refrigerants (including R23 trifluoromethane, R404A

fluoromethane blend, and R507 penta/trifluoromethane) in individual package chillers that will

provide cooling of UF6 cylinders and/or silicone oil heat exchange media for take-off stations

and cold traps (AES, 2009a). The applicant will use these halocarbon refrigerants because they

have good heat transfer properties, they satisfy environmental restrictions regarding ozone

depletion, and they are non-flammable. In all cases, the halocarbon refrigerants will be external

to the UF6 process stream and are not expected to interact with UF6.

Centrifuge Cooling Water

The applicant will provide centrifuge cooling water from the Centrifuge Cooling Water

Distribution System (AES, 2009a). The applicant will use this system to provide a supply of

deionized cooling water to the cooling coils of the centrifuge (AES, 2009a). This system will

provide stringent control over the operating temperature of the centrifuge to enable its efficient

operation. In all cases, centrifuge cooling water will be external to the UF6 process stream and

is not expected to interact with UF6.

6.3.1.5 UF6 and Construction Materials

UF6 is a fluorinating agent that reacts with most metals. The reaction between UF6 and metals

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such as nickel, copper, and aluminum will produce a protective fluoride film over the metal that

inhibits further reaction. These materials are therefore relatively inert to UF6 corrosion after

passivation and are suitable for UF6 service.

Resistant metals such as stainless steel will be used in valve bellows and flex hoses. Aluminum

piping will be bent to minimize the use of fittings. Connections will be welded to minimize the

use of flanges and gaskets. The use of sealant materials will be minimized to reduce the

number of potential leak paths.

Non-metallic materials are required to seal connections in UF6 systems to facilitate valve and

instrument replacement as well as cylinder connections. The applicant will confirm that all

gasketing and packing material are appropriate for UF6 services. Typical materials that are

resistant to UF6 through the range of the facility operating conditions include butyl rubber,

Teflon, Viton, and Kel-F. 6

The applicant selected materials of construction compatible with the process. The material

compatibility is demonstrated by the corrosion rates provided in Table 6.2-2 of the

SAR (AES, 2009a). For instance, take the corrosion rate of stainless steel at 1 000C (i.e, 0.03

millimeters (mm) per year) (AES, 2009a) and the nominal wall thickness of the UF 6 piping from

Table 6.2-3 of the SAR (i.e., 3.7 mm) which will be constructed of aluminum and stainless steel

(AES, 2009a). These data suggest that it would take approximately 123 years to corrode the

nomimal wall thickness of the UF6 piping. Therefore, these corrosion rates indicate that these

materials are acceptable for UF6 service over the life of the facility.

The cylinders used to store and transport UF6 are made out of carbon steel and are standard

Department of Transportation (DOT)-approved containers, designed and fabricated in

accordance with ANSI N14.1 (ANSI, 1995) (AES, 2009a). Tails cylinders will be stored outside

in open air where they will be exposed to the elements. Feed and product cylinders will be

subject to short duration exterior storage (months) and will be inspected in accordance with

requirements of DOT regulations upon receipt and prior to shipment.

The nominal and minimum wall thicknesses for cylinders are listed in Table 6.2-3 of the SAR

(AES, 2009a). The carbon steel storage cylinders are painted to provide a corrosion barrier to

external elements. Tails cylinders will be periodically inspected to assess corrosion and

corrosion rate. This provides reasonable assurance that there is not undue risk to members of

the public from the storage of tails cylinders.

6.3.1.6 Process Description Conclusion

The staff finds that the applicant has provided process descriptions that are sufficiently detailed

to allow an understanding of the chemical process hazards, and the chemical hazards that

could result from potential chemical interactions. The rational for these findings is that the

applicant provided expected process operating conditions, interactions between chemicals, and

interactions between process chemicals and the material of the equipment that will be used to

contain these process chemicals. Also, the information provided by the applicant allowed the

6Teflon, Viton, and KeI-F are tradenames for different fluorocarbon-based polymers.

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development of potential accident sequences. Therefore, the information that the applicant

provided, as described above, meets the guidance in Section 6.4.3.1 bullets (1) and (2) of

NUREG-1520 (NRC, 2002) and is acceptable.

6.3.2 Chemical Accident Sequences

The ISA Summary Section 3.7, Table 3.7-1 lists the chemical accident sequences, and Table

3.7-2 provides a narrative description (AES, 2009b). The applicant determined that the

consequences of the chemical accident sequences identified in these tables have the potential

to exceed the performance requirements of 10 CFR 70.61. The chemical accident sequences

covered the following areas:

(a) UF6 Feed System,

(b) Tails Take-off System,

(c) Product Take-off System,

(d) Product Blending System,

(e) Product Liquid Sampling System,

(f) Cylinder Preparation Room,

(g) Dump System,

(h) Decontamination System,

(i) Sub-sampling System,

(j) Separation Building Module Gaseous Effluent Ventilation System,

(k) Centrifuge Test/Centrifuge Post Mortem, and

(I) Ventilated Room System.

The applicant identified a total of 42 chemical accident sequences. The chemical accident

sequences covered the range of events that could result in a loss of confinement of UF 6 and the

hazardous chemicals produced from UF6 (i.e., U0 2F2 and HF). The accident sequences

addressed both high- and intermediate-consequence events (AES, 2009b). High- and

intermediate-consequence events are defined in 10 CFR 70.61 (b) and 70.61(c), respectively.

A high-consequence event is the release of licensed material or hazardous chemicals produced

from licensed material that, if an individual is exposed, it could endanger the life of a worker, or

lead to irreversible or other serious, long-lasting health effect to a member of the public. Also, a

high-consequence event includes an intake of 30 milligrams of soluble uranium by a member of

the public. An intermediate-consequence event is the release of licensed material or hazardous

chemicals produced from licensed material that, if an individual is exposed, it could lead to

irreversible or other serious, long-lasting health effects to a worker, or could cause mild transient

health effects to a member of the public. The staff performed a risk-informed review of selected

chemical accident sequences and also performed four vertical slice reviews. Chemical accident

sequences are discussed in Chapter 3, and Appendix A of this SER.

The staff concludes that the applicant has identified appropriate chemical accident sequences

based on the applicant's use of one of the recommended hazard evaluation methods (i.e.,

Hazard and Operability [HAZOP] Analysis) contained in NUREG-1 513 (NRC, 2001) to identify

those sequences and the results of the above staff review. In addition, the information provided

by the applicant, as described above and in Chapter 3, and Appendix A of this SER, includes a

list of the accident sequences with their respective consequence and likelihood that were

identified in the ISA Summary (AES, 2009b) that involve hazardous chemicals produced from

licensed material and chemical risks of plant conditions that affect the safety of licensed

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material. Also, the applicant described the postulated high-consequence events, how they will

be detected, and the mitigative measures on Section 2.0 of the Emergency Plan (AES, 2009d).

The actions described on the Emergency Plan (AES, 2009d) are consistent with the

consequences of the accident sequences identified in the ISA Summary (AES, 2009b). Based

on the above, the information provided by the applicant meets the guidance in Section 6.4.3.1

bullet (2) of NUREG-1520 (NRC, 2002) and is, therefore, acceptable.

6.3.3 Chemical Accident Consequences

The chemical exposure limits proposed by the applicant for HF and UF6 are based on the Acute

Exposure Guideline Limit (AEGL) values. Since U0 2F2 is a soluble uranium compound, the

applicant used the values presented in Table 2 of NUREG-1391 (NRC, 1991) to evaluate

soluble uranium exposure in terms of both chemical toxicity and radiological dose. The AEGL

values for HF and UF6 , and the values for soluble uranium intake are provided in Chapter 3 of

this SER. These exposure limits and the categorization of the severity of the chemical accident

consequences proposed by the applicant are consistent with NUREG-1 520, Table A-5,

"Consequence Severity Categories Based on 10 CFR 70.61." The applicant proposed to use

the 10-minute AEGL values for exposures to workers whose duration was 10 minutes or less.

For the public, the exposure duration was assumed to be 30 minutes. This is consistent with

self-protective criteria for UF6HF plumes listed in NUREG-1 140 (NRC, 1988). The staff finds

that the use of AEGL values is consistent with the guidance in Section 6.4.3.1 bullet (6) of

NUREG-1 520 (NRC, 2002) and is, therefore, acceptable for the determination of compliance

with the performance requirements of 10 CFR 70.61.

The applicant used the methods prescribed in NUREG/CR-6410 (NRC, 1998) to determine the

source terms. The source term was estimated using a five factor formula contained in section

3.2.5.2 of NUREG/CR-6410 (NRC, 1998):

ST = MAR x DR x ARF x RF x LPF

ST = Source term, UF6 or HF available to cause consequences, grams

MAR = Material at risk, the amount of UF6 that potentially could be impacted by the

accident, grams

DR = Damage ratio, fraction of the MAR actually impacted by the accident, unitless

ARF = Airborne release fraction, the fraction of material affected by the accident that

becomes airborne, unitless

RF = Respirable fraction, the fraction of material that becomes airborne with a

particle size that can be inhaled, unitless

LPF = Leak path factor, the fraction of the respirable material that leaves any

confinement or containment barrier, unitless

The staff's review of the ISA and supporting documentation found the source terms values to be

reasonable because the applicant provided the MAR values for each building and followed the

methodology described in NUREG/CR-641 0 (NRC, 1998). The values for the MAR can be

found in Table 2-14 of Appendix E to the ISA Summary (AES, 2009b). Site boundary

atmospheric dispersion factors were generated based on Regulatory Guide 1.145 (NRC, 1982).

Meteorological data collected at Argonne National Lab-West (EBR) - which is now identified as

MFC (Materials and Fuels Complex) - a mesonet station on the Idaho National Laboratory

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property that is located 18 kilometers (11 miles) west of the EREF site, during 2003-2007, were

used. The applicant also used modeling methods for source term determination, release

fraction, dispersion factors, and meteorological conditions as prescribed in Regulatory Guide

1.145 (NRC, 1982). The information provided by the applicant meets the guidance of Section

6.4.3.1 bullets (3) and (5) of NUREG-1520 (NRC, 2002) because, as described in Appendix E to

the ISA Summary (AES, 2009b), the applicant identified and used appropriate techniques in

estimating the concentration of hazardous chemicals produced from licensed material, used the

performance requirements criteria of 10 CFR 70.61, and the consequence analysis conform to

the guidance in NUREG/CR-6410 (NRC, 1998). Therefore, 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 Table 3.7-2 (AES, 2009b) describes each of the accident sequences identified by

the applicant and the specific IROFS that are applied to prevent the accident sequence or

mitigate its consequences. The ISA Summary Table 3.8-1 (AES, 2009b) 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 Sections 3.4 and 3.5 of the ISA Summary to identify where each IROFS

would be used and how the IROFS would function to prevent the accident sequence or mitigate

its consequences. The identified IROFS provide protection to prevent a loss of confinement of

licensed material during operation of the facility. Based on this system level review, the staff

concludes that the applicant has met the guidance of Section 6.4.3.2 bullet (2) of NUREG-1520

(NRC, 2002) because the ISA Summary (AES, 2009b) identified chemical process IROFS to

prevent or mitigate the consequences of accident sequences that involve the chemical hazards

of licensed material and hazardous chemicals produced from licensed material. In addition, the

ISA Summary (AES, 2009b) identified the hazards being mitigated and the risk category of each

accident sequence.

6.3.4.2 Management Measures

The applicant identified management measures to ensure that chemical safety IROFS would be

available and reliable to perform their safety function when needed. A Quality Assurance (QA)

Level will be applied to IROFS based in the following criteria described in the Quality Assurance

Program Description (QAPD) (AES, 2009c):

QA Level 1 Items whose failure or malfunction could directly result in a condition that adversely

affects the public, the worker, and the environment as described in 10 CFR 70.61.

The failure of a single QA Level 1 item could result in a high or intermediate

consequence.

QA Level 2 Items whose failure or malfunction could indirectly result in a condition that

adversely affect the public, the worker, and the environment as described in 10

CFR 70.61. The failure of a QA Level 2 item, in conjunction with the failure of an

additional item, could result in a high or intermediate consequence.

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QA Level 3 Items that are not classified as QA Level 1 or QA Level 2. QA Level 3 items are

controlled in accordance with standard commercial practices.

The following sections briefly discuss the management measures applied to chemical safety.

Chapter 11 of this SER provides an evaluation of the management measures applied to the

proposed EREF.

6.3.4.2.1 Configuration Management

The configuration management proposed for the facility includes those controls that ensure that

the facility design basis is thoroughly documented and maintained, and that changes to the

design basis are controlled. This includes the following:

* That management commitment and staffing is appropriate to ensure that configuration

management is maintained;

* That proper QA is in place for design control, document control, and records management;

and

* That all structures, systems, and components, including IROFS, are under appropriate

configuration management. (AES, 2009a)

6.3.4.2.2 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 (AES, 2009a). These controls

include planned and scheduled maintenance of equipment and controls so that design features

will function when required (AES, 2009a). 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.

Maintenance activities generally fall into the following categories:

1. Surveillance/monitoring,

2. Corrective maintenance,

3. Preventive maintenance, and

4. Functional testing.

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6.3.4.2.3 Training

The applicant proposes to provide training to individuals who handle licensed materials and

other chemicals at the facility (AES, 2009a). 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 the event of a severe chemical release;

and

7. Coordination of chemical process safety training curriculum with that of other areas

including, radiological safety, criticality safety, facility operations, emergency response,

and related areas (AES, 2009a).

6.3.4.2.4 Procedures

The applicant proposes to use four types of facility 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:

1. Directions for normal operations, including startup 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;

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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. (AES, 2009a)

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. Recordkeeping and document control; and

9. Reporting (AES, 2009a).

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 (AES, 2009a).

Maintenance procedures address:

1. Preventive and corrective maintenance of IROFS;

2. Surveillance (including calibration, inspection, and other surveillance testing);

3. Functional testing of IROFS; and

4. Requirements for pre-maintenance activity involving review of the work to be performed

and review of procedures (AES, 2009a).

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Emergency procedures address the pre-planned actions of operations and other facility

personnel in case of an emergency (AES, 2009a).

6.3.4.2.5 Audits and Assessments

The applicant proposes to conduct audits to determine that facility operations are performed in

compliance with regulatory requirements, license conditions, and written procedures (AES,

2009a). The applicant assesses activities related to radiation protection, criticality safety

control, hazardous chemical safety, fire protection, and environmental protection.

The applicant will perform audits in accordance with a written plan, which will identify and

schedule audits to be performed. Audit team members have no direct responsibility for the

function area being audited. Team members will have technical expertise or experience in the

function area being audited and will be trained in audit techniques. The applicant will conduct

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 productions activities

will be used to perform routine surveillance/assessments (AES, 2009a). Deficiencies noted

during the inspection, requiring corrective action, will be forwarded to the manager of the

applicable area or function for action (AES, 2009a). Future surveillance/assessments will

include a review to evaluate if corrective actions have been effective (AES, 2009a).

6.3.4.2.6 Incident Investigation and Corrective Actions

The applicant has a facility-wide incident investigation process that includes chemical process

related incidents (AES, 2009a). This process is available for use by any person at the facility for

reporting abnormal events and potentially unsafe conditions or activities (AES, 2009a). Events

that potentially threaten or lessen the effectiveness of health, safety, or environment protection

will be identified, reported to, and investigated by the Environmental Health, Safety & Licensing

Manager (AES, 2009a). Each event will be considered in terms of its requirements for reporting

in accordance with regulations and will be evaluated to determine the level of investigation

required (AES, 2009a). These evaluations and investigations will be conducted in accordance

with applicant-approved procedures. The depth of the investigation will depend upon the

severity of the classified incident in terms of the levels of uranium/chemical released and/or the

degree of potential for exposure to workers, the public, or the environment.

6.3.4.2.7 IROFS and Management Measures Conclusion

The information the applicant provided, as described above and in Appendices A and B of this

SER, meets the guidance in Section 6.4.3.2 bullet (2) of NUREG-1520 (NRC, 2002) and is

acceptable because the applicant has adequately identified the administrative and engineered

controls (IROFS) to prevent chemical accident sequences or mitigate their consequences at the

proposed facility. Also, the applicant identified the hazards being mitigated and the risk

category.

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In addition, the information the applicant provided, as described above and in Chapter 11 of this

SER, meets the guidance in Section 6.4.3.2 bullet (3) of NUREG-1 520 (NRC, 2002) and is

acceptable because the applicant provided sufficient description of its procedures to ensure the

reliable operation of engineered controls and that administrative controls will be correctly

implemented.

6.3.5 Emergency Management

The proposed facility has an emergency plan and program which include response to mitigate

the potential impact of any process chemical release, including requirements for notification and

reporting of accidental chemical releases (AES, 2009a). The EREF fire brigade/emergency

response team will be outfitted, equipped, and trained for hazardous material response; and

local agencies can supplement with additional response teams (AES, 2009a). The City of Idaho

Falls, Idaho, Fire Department is the nearest offsite response agency that can supplement the

proposed EREF with additional response teams. The applicant states in Section 4.4.2 of the

Emergency Plan (AES, 2009d) that offsite response and support groups will be offered annual

radiological and chemical response training specific to the proposed EREF.

In Chapter 10 of the Emergency Plan (AES, 2009d) for the proposed EREF, the applicant

certifies that the facility has met all the responsibilities under the Emergency Planning and

Community Right to Know Act of 1986, Title Ill, Public Law 99-499, as required by 10 CFR

70.22(i)(3)(xiii). The applicant states that the Material Safety Data Sheets are required for all

received hazardous chemicals, controlled, and made available at one or more locations within

the facility to ensure that they are readily accessible to all work shifts and for use in an

emergency (AES, 2009d).

The information the applicant provided, as described above, meets the guidance in Section

6.4.3.1 bullet (2) of NUREG-1 520 (NRC, 2002), and is acceptable because the applicant

described the postulated high-consequence events, how they will be detected, and the

mitigative measures on Section 2.0 of the Emergency Plan (AES, 2009d). The actions

described on the Emergency Plan (AES, 2009d) are consistent with the consequences of the

accident sequences identified in the ISA Summary (AES, 2009b).

6.3.6 Baseline Design Criteria

The applicant provides design bases information for chemical process safety IROFS for the

proposed facility in the SAR (AES, 2009a) and the ISA Summary (AES, 2009b). 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 material."

The only chemical of concern (EREF Class 1) is UF6. Details of design and safety features of

all chemical process systems are found in Chapter 3 of the ISA Summary (AES, 2009b). 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:

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" Managing the arrangement and size of material containers and processes;

* Selection and use of materials 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 the ISA Summary Sections 3.3 through 3.5, and the process hazards

description in Section 3.6 (AES, 2009b). The staff notes that the uranium enrichment process is

basically a physical process that separates the 235U isotope from the 238U isotope based on their

difference in mass. The uranium is in the chemical form of UF 6. The entire process, with the

exception of the product liquid sampling step, will be conducted under a significant vacuum.

Furthermore, the process design involves very limited inventories of UF 6 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 moisture in the air, discussed in Section 6.3.1.3 of this SER) are such that most uraniumbearing material is likely to accumulate near any process breach. Because 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 centrifuges. 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 bases

provide for adequate protection against chemical risks.

The staff notes that the applicant's design of product liquid sampling system uses an

ANSI N14.1 qualified cylinder as the primary confinement vessel and an American Society of

Mechanical Engineers Code pressure vessel as a secondary confinement system (AES,

2009b). The staff concludes that this design approach for the liquid portion of the process is

acceptable because it uses recognized nuclear fuel cycle industry codes and standards.

The staff reviewed the results of the applicant's HAZOP analysis as discussed in Chapter 3 of

this SER. This method is widely used in the chemical industry during the design phase to

identify operability and safety issues, and is identified as an acceptable method in Section 2.4 of

NUREG-1513 (NRC, 2001). As applied to the gas centrifuge uranium enrichment process, the

applicant's HAZOP analysis considered a variety of internal process, facility, and external

hazards that could breach the process and release licensed material and hazardous chemicals

produced from licensed material. The results of the applicant's ISA are presented in Table 3.7-1

of the ISA Summary (AES, 2009b). The table contains information concerning the accident

sequences identified as a result of the applicant's HAZOP analysis, the unmitigated risk of each

applicant-identified accident sequence, and the IROFS applied to prevent the accident

sequence or mitigate its consequences. 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.

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The staff concludes that the information the applicant provided and the applicant's proposed

design meets the guidance in Section 6.4.3.3 of NUREG-1 520 (NRC, 2002) because the

applicant addressed the list of items in Section 2.4 of NUREG-1601 (NRC, 1997). These items

should be considered in an adequate facility design. Also, the applicant described in the ISA

Summary how the ISA was performed and how it satisfies the performance requirements of 10

CFR 70.61. Based on the information in this section and the process descriptions of the ISA

Summary (AES, 2009b), the applicant applied defense-in-depth to the entire process (i.e.,

several alarms before reaching IROFS set points, vacuum operation for most parts of the

process and two independent confinement barriers for sampling operations). The applicant's

information also provides for adequate protection against chemical risks produced from licensed

materials, facility conditions which affect the safety of licensed material, and hazardous

chemicals produced from licensed material; 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 SAR (AES, 2009a), the NRC staff has concluded that the applicant has described and

assessed accident sequences that can result from the handling, storage, or processing of

licensed materials and that can potentially have significant chemical consequences and effects.

The applicant has prepared a hazard analysis that identifies and evaluated those chemical

process hazards and potential accidents and established safety controls providing reasonable

assurance of safe facility operation. To ensure that the performance requirements in

10 CFR Part 70 are met, the applicant has stated that controls are maintained, available, and

reliable to perform their safety-related functions when needed. The staff has reviewed and

independently verified these safety controls and the applicant's plan to 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 for 10 CFR Part 70, and provides

reasonable assurance that the public health and safety and the environment will be protected.

6.5 References

(AES, 2009a) AREVA Enrichment Services, "Eagle Rock Enrichment Facility Safety Analysis

Report," Revision 1, April 2009.

(AES, 2009b) AREVA Enrichment Services, "Eagle Rock Enrichment Facility Integrated Safety

Analysis Summary," Revision 1, April 2009.

(AES, 2009c) AREVA Enrichment Services, "Quality Assurance Program Description for

Design, Construction, Operation and Decommissioning of the Eagle Rock Enrichment Facility,"

Revision 1, April 2009.

(AES, 2009d) AREVA Enrichment Services, "Eagle Rock Enrichment Facility Emergency Plan,"

Revision 1, April 2009.

(ANSI, 1995) American National Standards Institute, ANSI N14.1, "Uranium Hexafluoride -

Packaging for Transportation," 1995.

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(NRC, 1982) U.S. Nuclear Regulatory Commission, Regulatory Guide 1.145,

Revision 1, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments

at Nuclear Power Plants," November 1982.

(NRC, 1988) U.S. Nuclear Regulatory Commission, NUREG-1140, "A Regulatory Analysis on

Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees," January

1988.

(NRC, 1991) U.S. Nuclear Regulatory Commission, NUREG-1391, "Chemical Toxicity of

Uranium Hexafluoride Compared to Acute Effects of Radiation," February 1991.

(NRC, 1997) U.S. Nuclear Regulatory Commission, NUREG-1601, "Chemical Process Safety

at Fuel Cycle Facilities," August 1997.

(NRC, 1998) U.S. Nuclear Regulatory Commission, NUREG/CR-6410, "Nuclear Fuel Cycle

Accident Analysis Handbook," March 1998.

(NRC, 2001) U.S. Nuclear Regulatory Commission, NUREG-1 513, "Integrated Safety Analysis

Guidance Document," May 2001.

(NRC, 2002) U.S. Nuclear Regulatory Commission, NUREG-1 520, "Standard Review Plan for

the Review of a License Application for a Fuel Cycle Facility," March 2002.

(NRC, 2009) U.S. Nuclear Regulatory Commission, "Memorandum from M. Breeda Reilly to

Thomas G. Hiltz Regarding the July 13-15, 2009, AREVA Enrichment Services Eagle Rock

License Application-Integrated Safety Analysis In-Office Review," August 2009.

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CHAPTER 7.0 FIRE SAFETY

The purpose of this review is to determine, 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 Title 10 of the Code of Federal Regulations (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-1 520 (NRC, 2002). The

following section, "Staff Review and Analysis," provides details on the acceptance criteria and

describes how the applicant satisfies them.

7.3 Staff Review and Analysis

This chapter addresses the staffs review of facility fire protection, including fire safety

management measures, fire hazards analysis, facility fire protection, process fire safety, and fire

safety and emergency response, as presented in the Safety Analysis Report (SAR)

(AES, 2009a) and supporting documents (AES, 2009b and AES, 2010).

The Eagle Rock Enrichment Facility (EREF) and its Fire Safety program were reviewed to

determine applicability and level of compliance with the National Fire Protection Association

Standard (NFPA) 801 (NFPA, 2008c) and applicable standards referenced within. Section 7.6

of the SAR (AES, 2009a) lists the fire codes and standards considered by the applicant to be

applicable to the facility. The staff finds the use of these consensus codes and standards to be

in accordance with the guidance of Section 7.4.3 of NUREG-1520 (NRC, 2002) in regard to

nationally recognized codes and standards that may be used to measure reasonable assurance

of fire safety. Therefore, the staff considers the use of the above codes and standards to satisfy

the requirements of 10 CFR 70.64(a) Baseline Design Criterion (3), "Fire Protection."

7.3.1 Fire Safety Management Measures

The applicant will implement fire safety management measures as described in Chapter 11 of

the SAR (AES, 2009a). Management measures applicable to fire safety include: configuration

management; maintenance; training; procedures; audits and assessments; incident reporting

and investigations; and records management. These measures will ensure that fire protection

IROFS are available and reliable. The applicant will follow the codes and standards, as listed in

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Section 7.6 of the SAR (AES, 2009a), which are applicable to the individual fire safety

management measures. Management measures are evaluated in Chapter 11 of this Safety

Evaluation Report (SER).

7.3.1.1 Management Policy and Direction

The Environmental Health, Safety and Licensing Manager is responsible for fire protection and

is assisted by the Safety, Security, and Emergency Preparedness Manager, who is responsible

for day-to-day safe operation of the facility, including fire safety. The Safety, Security, and

Emergency Preparedness 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 recordkeeping, 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 combustion

generated smoke is not used for leak-testing, (d) conduct periodic fire prevention inspections,

(e) perform periodic housekeeping inspections, and (f) implement a system to control the

disarming of the various types of fire detection or fire suppression systems. Further information

concerning the fire detection and suppression systems can be found in Section 7.3.3 of this

chapter. The inspection, testing, and maintenance of fire protection systems will comply with

nationally recognized industry standards. The applicant's 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.

7.3.1.2 Fire Safety Management Measures Conclusions

Based on its review, the staff reached the following conclusions:

Consistent with the acceptance criteria from Section 7.4.3.1 of NUREG-1 520 (NRC, 2002), the

applicant's fire safety management measures identify a senior level manager who has the

authority and staff to ensure that fire safety receives appropriate priority, and are, therefore,

acceptable.

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Consistent with the acceptance criteria from Section 7.4.3.1 of NUREG-1520 (NRC, 2002), the

applicant's fire safety management measures identify a facility safety committee staffed by

managers of different disciplines to integrate facility modifications, and are, therefore,

acceptable.

Consistent with the acceptance criteria from Section 7.4.3.1 of NUREG-1520 (NRC, 2002), the

applicant's fire safety management measures include fire prevention; inspection, testing, and

maintenance of fire protection systems; Emergency Response Organization (ERO)

qualifications, drills, and training; and pre-fire plans as recommended by NFPA 801 (NFPA,

2003e) and are, therefore, acceptable.

Consistent with the acceptance criteria from Section 7.4.3.1 of NUREG-1 520 (NRC, 2002), the

applicant's fire safety management measures are documented in sufficient detail to identify their

relationship to, and functions for normal operations; anticipated (off-normal) events; and

accident safety (i.e., IROFS), and are, therefore, acceptable.

Consistent with the acceptance criteria from Section 7.4.3.1 of NUREG-1 520 (NRC, 2002), the

applicant's fire safety management measures will ensure that the IROFS, as identified in the

ISA Summary, are available and reliable; and will ensure that the facility maintains fire safety

awareness among employees, controls transient ignition sources and combustibles, and

maintains a readiness to extinguish or limit the consequences of fire. Therefore, these

measures are acceptable.

Based on the above conclusions, the staff has reasonable assurance that the applicant's fire

safety management measures would meet the requirements of 10 CFR 30.33, 40.32, 70.22,

70.61, 70.64, and 70.65 as they pertain to the fire protection aspects of the facility.

7.3.2 Fire Hazards Analysis (FHA)

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 Cylinder Receipt and Shipping Building (CRSB);

" Fire involving Cylinder Transporters/Movers;

• Fire inside the Cascade Halls;

" Fire inside the uranium hexafluoride (UF6) handling area/Blending, Sampling, and

Preparation Building;

* Fire inside the Technical Support Building (TSB); and

• Fire affecting cylinder storage pads.

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Exterior and interior building explosions as initiating events to accident sequences were

evaluated and found to be highly unlikely, without the need for IROFS. These initiating events

had a frequency of less than 106/year which satisfies the performance criteria requirements in

10 CFR 70.61, therefore no IROFS are required.

7.3.2.1 IROFS Related to Fire Safety

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 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 assigned to these measures,

which are in accordance with the NRC guidance provided in Table A-9 of NUREG-1 520. All of

the listed IROFS will also be supported by the general management measures, as described in

Section 3.1.8.3 of the ISA Summary. Further analysis of the ISA methodology can be found in

Chapter 3 of this SER. In conclusion, the staff finds the selection of accident sequences and

the determination of the IROFS related to fire protection to be acceptable to satisfy the

performance requirements of 10 CFR 70.61.

The remaining features of fire protection that are described in the license application are fire

protection measures that provide overall defense-in-depth protection of fire safety for

operations. All fire protection measures are further discussed and evaluated in the following

subsections discussing each fire related accident scenario. The applicant's ISA Summary has

adequately addressed fire risks in accordance with the regulation and the guidance established

in NUREG-1 520 (NRC, 2002).

7-4
Table 7.3-1 Identification of IROFS

Failure

IROFS probability Description of IROFS

index

Automatic closure of fire rated barrier opening protections (e.g.,

IROFS42 -3 doors, dampers, penetration seals) to ensure that the integrity of

area fire barriers prevents fires from propagating into areas

containing uranic material.

Administratively limit transient combustible loading in areas

IROFS43 -3 containing uranic material to ensure integrity of uranic material

components/ccntainers, and limit the quantity of uranic material at

risk, to ensure that consequences to the public are low.

Administratively limit fire exposure to 30B product cylinders

containing > 0.1 kg of UF6 from a semi-tractor trailer fire during the

shipping process. This will be controlled by requiring that all

IROFS44 -3 outgoing 30B

product cylinders are loaded into their required U.S. Department of

Transportation (DOT) overpacks prior to placement on the semitractor trailer.

IROFS45 -3 Administratively limit storage of UF6 cylinders in the CRSB, to ensure

> 1-m (3-ft) setback from the edge of the loading dock.

Automatic hardwired, fail-safe trip of the Chemical Trap Workshop

heating, ventilation, and air conditioning and isolation from Technical

IROFS46 -2 Support Building (TSB) gaseous effluent ventilation system on

smoke detection with low exfiltration from the room to ensure low

consequences to the public.

Administratively limit vehicle approach to bare 30B UF6 cylinders

IROFS47 -3 containing greater than heel quantity-1 1.3 kg (25 Ib) for all exterior

storage and handling areas.

Administrative control of diesel fuel delivery vehicle route and

staging.

Administratively limit exposure by requiring worker action to

IROFS49 -3 evacuate the area(s) of concern to ensure worker consequences of

inhalation of uranic material and Hydrogen Fluoride (HF) are low.

IROFS67 -3 Passive engineered control of intrasite cylinder movement vehicles

to ensure they do not result in rupture of 30B or 48Y cylinders.

Administrative controls to ensure that only authorized vehicles

IROFS73 -3 operate in areas proximate to exterior locations where bare 30B

product cylinders (> heel quantity) are stored or handled.

Automatic suppression of a fire to ensure consequences to the

public is low. Included within the boundary of IROFS is an

IROFS100 -2 administrative control, independent from IRFOS43, to maintain

combustibles at acceptable levels in accordance with NFPA 13

(NFPA, 2007b).

7-5
7.3.2.2 Fire in the CRSB

All UF6 feed cylinders and empty product and byproduct cylinders will enter the facility through

the CRSB. The applicant considered three separate fire scenarios in the CRSB. These fire

scenarios were a fire at the CRSB loading dock, a fire in the CRSB general areas, and a fire

involving a cylinder delivery vehicle.

At the CRSB loading dock, UF6 is contained in 48Y 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

dock. The 1 m (3.3 ft) setback is an administrative control (IROFS45). 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 and agrees with this assessment, given

the calculation was performed utilizing the guidance provided in NUREG-1805 (NRC, 2004).

The CRSB General Areas contain UF6 in 48Y and 30B cylinders. Combustible loading is

expected to be very low, and transient combustibles will be controlled with IROFS43. No liquid

combustibles are listed for the CRSB 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 (IROFS42) between adjacent fire areas and the

CRSB 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 agrees with this

assessment.

The applicant also evaluated a fire involving UF6 cylinders present on a delivery vehicle. The

Department of Transportation's (DOT's) 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 offsite UF6 shipments. Hence,

both incoming cylinders and outgoing cylinders will be protected by approved thermal

protection. The handling practice for incoming cylinders containing UF6will 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 30 minutes. Approved protective assemblies (IROFS44) 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 calculations

and agrees with the results, given the calculations were performed utilizing guidance previously

approved by the NRC (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.

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The staff concludes 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 CRSB.

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 (48Y or 30B) in transit. Only electric drive cylinder transporters with preventative

measures engineered to prevent a rupture of a cylinder will be used for cylinder transport at the

proposed facility (IROFS67). When filled 30B cylinders are transported outside of the buildings,

they are protected by DOT approved overpacks (IROFS44).

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 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 Separations Building

The eight cascade halls containing the centrifuges are located in the Separations Building. The

fire scenario inside the cascade halls is assumed to take place inside a module that holds

twelve cascades, each cascade containing hundreds of centrifuges. If the entire module was

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 (IROFS42) and by

automatic fire suppression (IROFS1 00) in the Process Services Corridor of the Separations

Building. A fire originating inside the module is presumed to involve the cables feeding the

centrifuge drive motors. If transient combustibles are controlled (IROFS43), this fire, at worst, is

expected to cause failures in the aluminum piping manifold releasing 100% of the inventory

feeding one assay, resulting in consequences below the threshold of the 10 CFR 70.61(c)

intermediate-consequence limit. This analysis was based upon the guidance provided in

NUREG-1805 (NRC, 2004). Based on the above, the applicant has determined that the

likelihood of a fire being initiated and the IROFS failing, so that a release exceeding the

consequence threshold of 10 CFR 70.61(b) or (c) occurs is highly unlikely.

The staff concludes 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 Separations

Building.

7.3.2.5 Fire in the UF6 Handling Area or Blending, Sampling, and Preparation Building

The UF6 Handling Area contains the UF6 Feed System, Product Take-Off System, and the Tails

Take-Off System. The Blending, Sampling, and Preparation Building 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 UF6from the various feed, product, and tails

system cold traps. A fire is prevented from propagating into the UF6 Handling Area, and

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Blending and Liquid Sampling Area, by fire barriers (IROFS42) and the automatic fire

suppression system (IROFS100). 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 10 CFR 70.61(c) intermediate-consequence limit. Severe

fires will be prevented by IROFS43, which will control the location and the amount of transient

combustibles in the area, and by the automatic fire suppression system (IROFS1 00). 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 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 or the Blending, Sampling, and Preparation Building.

7.3.2.6 Fire in the TSB

The TSB contains support areas for the facility, such as the Solid Waste Collection Room,

Chemical Trap Workshop, Sample Bottle Storage Room, various laboratories, 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, Chemical Trap Workshop, and Sample Bottle

Storage Room.

The uranium inventory in the Solid Waste Collection Room is contained in 12-L (3.2-gallon [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 (IROFS42) and automatic fire suppression

(IROFS100). 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 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

(IROFS43). Additionally, automatic fire suppression will be provided (IROFS100). 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 Chemical Trap Workshop 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 room

by fire rated barriers (IROFS 42) and automatic fire suppression (IROFS100). For a fire

originating in the area, and involving expected in-situ and transient combustibles, the applicant

postulates that several kilograms of uranic materials would be present in open containers or

drums during transfer/bulking 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 (IROFS43). Additionally, to mitigate the

severity to low consequence, automatic fire suppression will be provided (IROFS1 00) and

7-8
smoke detection interlocked to isolate the room's ventilation systems will also be provided

(IROFS46). 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 TSB Sample Bottle 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 (IROFS42) and automatic fire suppression

(IROFS100). For a fire originating in the Sample Storage Room with controls in place

(IROFS43), with expected insitu and transient combustibles, the consequences would be low.

The fire would not have sufficient combustibles to fail a sample cylinder. Additionally, automatic

fire suppression will be provided (IROFS100) consistent with criticality safety requirements.

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 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.

7.3.2.7 Cylinder Storage Pads

The Cylinder Storage Pads provide storage for full product cylinders, full feed cylinders, tails

cylinders, and 6 months of empty product and feed cylinders. The Storage Pads occupy

approximately 17 hectares (42 acres) and are sized to accommodate enough cylinders for

30 years of operation.

For the full product cylinder storage pad, fires were postulated on the transporter and from other

vehicles near the pad. Only electric drive cylinder transporters with preventative measures

engineered to prevent a rupture of a cylinder will be used for cylinder transport at the proposed

facility (IROFS67). In order to prevent exposure from a potential vehicle fire of fuel spill fire,

preventative measures are to place cylinders on the pad behind a vehicle impact barrier

(IROFS47). The barrier also provides separation distance to prevent fuel fires from migrating

onto the cylinder storage pad and to provide radiant heat reduction from cylinders from vehicle

and fuel spill fires. The combination of barriers and separation distance values was shown to

prevent cylinder failure from fire exposure. The cylinder target wall temperature was less than

600°C (11 12°F). Additionally, vehicle operation is administratively controlled (IROFS73) to

allow only authorized vehicles and operators within exterior areas of bare 30B cylinders. 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, in accordance with the guidance provided in

NUREG-1 805 (NRC, 2004), to determine the possible effects of a diesel fuel delivery vehicle on

the storage pads and staging areas. To prevent exposure to a potential diesel fuel delivery

vehicle fire or fuel spill fire from this vehicle near the Full Tails Cylinder Storage Pad, the Full

Feed Cylinder Storage Pad, the Empty Cylinder Storage Pad, the Truck Entry Staging Area, and

the Cylinder Transporter Path, preventive measures are to designate a specific route for fuel

deliveries that is remote from these areas and to require all fuel delivery vehicles to be

accompanied onsite to ensure delivery only via the designated route (IROFS48). Based on the

7-9
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 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 near the Cylinder

Storage Pads.

7.3.2.8 FHA Conclusions

The NRC staff concludes 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 preventing

or mitigating fire 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-1 520 (NRC, 2002).

Consistent with the acceptance criteria from Section 7.4.3.2 of NUREG-1520 (NRC, 2002), an

FHA was performed for each process area and each FHA included a description by fire area of

the fuel loading, fire scenarios, methods of consequence analysis, the potential consequences,

and a description of mitigative controls, and is, therefore, acceptable. In addition, based on its

review of the above information, the staff concludes that the applicant's performance of their

FHA meets the requirements of 10 CFR 30.33, 40.32, 70.22, 70.61, 70.64, and 70.65, as they

pertain to the fire protection aspects of the facility.

7.3.3 Facility Design

7.3.3.1 Facility Passive-Engineered Fire Protection Systems

Buildings containing UF6 are the TSB; CRSB; Cascade Halls; and the Blending, Sampling, and

Preparation Building, which have protected structural steel columns and trusses with built-up

composite roofing on metal deck. This construction is classified as Type 11-222, in accordance

with NFPA 220 (NFPA, 2006). 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 2 hours. Floors have

a resistance of at least 2 hours and roofs have a fire resistance of at least 1 hour.

The Centrifuge Assembly Building (including the Centrifuge Test and Centrifuge Post Mortem

Area), which also contains UF 6, will have insulated metal panel exterior walls with built-up

composite roofing on metal deck. This construction is classified as Type 11-000 in accordance

with NFPA 220 (NFPA, 2006). This means that all structural members are of non-combustible

or limited combustible construction, but no fire rating is required.

The remaining utility and non-process-related areas are classified as Type 11-000, in accordance

with NFPA 220 (NFPA, 2006). 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 passive

7-10
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.

7.3.3.2 Facility Active-Engineered Fire Protection Systems

Electrical Installation and Fire Alarm System

All electrical systems at the facility are installed in accordance with NFPA 70 (NFPA, 2008b).

Switchgear, motor control centers, panel boards, variable frequency drives, uninterruptible

power supply systems, and control panels are mounted in metallic enclosures and contain

limited amounts of combustible material. Cable trays and conduits are metallic and the cable in

cable trays are flame retardant and tested in accordance with industry guidance.

Each building of the facility is equipped with a listed modular, multi-zone fire alarm control panel

installed in accordance with NFPA 72 (NFPA, 2007d). Each panel has a dual power supply

consisting of normal building power and backup power, by either a 24-hour battery or the facility

uninterruptible power supply (UPS). The sprinkler system and hose-station water-flow detection

are connected to separate control-panel zone modules. Fire detector and manual-pull stationalarm 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 a 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:

* 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 diesel and electric fire 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.

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Portable Fire Extinguishers

Portable fire extinguishers are installed throughout all buildings in accordance with NFPA 10

(NFPA, 2007a). Multipurpose fire extinguishers are provided generally for Class A (ordinary

combustibles), Class B (flammable and combustible liquids), and Class C (electrical equipment)

fires. Specialized extinguishers are located in areas requiring protection of particular hazards.

Wheeled extinguishers are provided for use in water-exclusion areas. In areas with moderator

control issues, the extinguishers are filled with carbon dioxide or dry chemical.

Fire Water Supply

The facility fire water supply consists of two 757,082-L (200,000-gal) water storage tanks

designed and constructed in accordance with NFPA 22 (NFPA, 2008a). Within each tank,

681,374-liter (L) (1 80,000-gal) is reserved for fire protection. Fill and makeup to the tanks are

from the well water supply to the site, and the water supply is capable of filling fire protection

water inventory in a single tank within an 8-hour period. The fire pumps consist of one electricdriven pump and one diesel-driven pump, both rated for 5678-liters per minute (Ipm) (1500-

gallons per minute [gpm]), at 10.35-bar (150-pounds-force per square inch absolute [psia]),

pumps. Both pumps are horizontal centrifugal pumps designed and installed in accordance with

NFPA 20 (NFPA, 2007c). The maximum anticipated fire demand is 5678 1pm (1500 gpm),

based on 3785 Lpm (1000 gpm) from a building-sprinkler system, plus 1892 Ipm (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. The tanks are arranged such that one will

be available for suction at all times. In addition to fixed standpipes and fire hose stations, the

facility will be provided with fire hose on mobile apparatus or at strategic locations throughout

the facility. The amount of hose provided will be sufficient to ensure that all points within the

facility will be able to be reached by at least two 38-mm (11⁄2-inch [in]) diameter hoses and one

64-millimeters (mm) (2 1⁄2-in)-diameter backup hose, consistent with NFPA 1410 (NFPA, 2005b).

These lines will have a minimum nozzle pressure of 4.5 bar (65 pounds-force per square inch

gauge [psig]) for the 38-mm (11⁄2-in) hose, and 6.9 bar (100 psig) for the 64-mm (2 1⁄2-in) hose.

Engineered Automatic Fire Suppression Systems

Automatic pre-action sprinkler systems are provided in the following buildings, subject to

moderator control restrictions:

* Process Service Corridor in the Separations Building Module;

" UF6 Handling Area,

* Technical Support Building, and

* Blending, Sampling, and Preparation Building.

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Automatic wet pipe sprinkler systems are provided in the following buildings:

* Administration Building,

* Security and Secure Administration Building,

* Long-Term Warehouse,

* Fire Pump House,

* Centrifuge Assembly Building,

* Operations Support Building, and

* Short-Term Warehouse.

These systems are designed and tested in accordance with NFPA 13 (NFPA, 2007b).

Sprinkler-system control valves are monitored under a periodic inspection program, and their

proper positioning is supervised in accordance with NFPA 801 (NFPA, 2008c).

The NRC staff concludes that the applicant has reasonably determined the required fire

protection 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.

7.3.3.3 Facility Design Conclusions

The applicant has addressed building construction, fire area determination, electrical

installation, life safety, drainage, and lightning protection adequately in the application. A

description of ventilation characteristics, as they relate to fire protection and fire hazards, will be

provided in the FHA. As stated in the introduction to Section 7.3 of this SER, the applicant

meets the baseline design criteria through compliance with accepted consensus standards.

The applicant meets defense-in-depth requirements as evaluated in Chapter 7 of this SER.

Consistent with the acceptance criteria from Section 7.4.3.3 of NUREG-1 520 (NRC, 2002), the

application documents the fire safety considerations used in the general design of the facilities

containing licensed material, or facilities that impose an exposure threat to radiological facilities

and is, therefore, acceptable. In addition, the applicant's facility's fire protection features meet

the requirements of 10 CFR 30.33, 40.32, 70.22, 70.61, 70.64, and 70.65 as they pertain to the

fire protection aspects of the facility.

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7.3.4 Process Fire Safety

The applicant plans to use gas centrifuge machines to enrich uranium up to 5 weight (wt.)

percent uranium-235 (

235U). The feed material will be UF6, which will be enriched by using 96

enrichment cascades with hundreds of gas centrifuge machines per cascade. The UF6 normal

feed is 0.711 wt. percent 235U; the expected product is a maximum of 5.0 wt. percent 235U; and

the depleted tails are typically 0.2 wt. percent 235U. Enriched and depleted UF6streams are

withdrawn from the cascade by pumps and returned to a solid phase in product and tails lowtemperature take-off stations, respectively. The remainder of this section describes the NRC's

staff review of key fire hazards and risks associated with the proposed facility and the gas

centrifuge enrichment process.

7.3.4.1 Process Descriptions

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 perfluoropolyether (PFPE) oil that is noncombustible. The used PFPE oil will be collected, packaged, and shipped offsite for disposal at

a licensed low-level radioactive waste facility.

Hydrogen Fluoride (HF): HF is a byproduct of the chemical reaction of UF6with water vapor.

HF is extremely reactive in both gaseous and aqueous form. HF alone is not flammable nor

combustible. It can, however, react exothermically with water to generate sufficient heat to

ignite nearby combustibles.

Uranyl Fluoride (U02F2): U02F2 is also a byproduct of the chemical reaction of UF6with water

vapor. U02F2 is stable in air to 300' C (5720 F). It is neither 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 an 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 the National Fire

Protection Association (NFPA) Standard 10 (NFPA, 2007a). 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.

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Storage and Handling of UF6: UF6 cylinders are stored or handled in the cylinder storage pads,

the CRSB, the UF6 handling areas, and the blending and liquid-sampling areas. On the storage

pads, fire concerns include the cylinder transport vehicle, a fire exposure from nearby

vegetation, and fire exposure from a nearby vehicle accident. The applicant 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

by IROFS discussed in Section 7.3.2 of this chapter. In the CRSB, the primary fire concern was

from a truck fire at the loading dock. The applicant also analyzed this and determined that the

cylinders could be adequately protected by storing them at least 1 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 via IROFS43.

Therefore, any fire originating in these areas will be limited.

Hydrogen Control: Hydrogen is used within the laboratories and may be generated at batterycharging 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

* Natural or mechanical ventilation will be provided to ensure that hydrogen

concentrations do not exceed 25 percent of the lower explosive limit. 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 the American National Standards Institute (ANSI)

Standard C2 (ANSI, 1981).

" Combustible Material Hazards: Materials of construction for the centrifuge process

building, the supporting buildings, and centrifuge machines and components are

predominantly non-combustible (e.g., steel, aluminum, and concrete floors). A minimum

of fixed combustibles is expected to be present in the operations areas, and the

applicant plans to control transient combustibles to minimize potential fire hazards by

utilizing IROFS43 and general housekeeping procedures. 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 50 L (13.2 gal) of oil. The oil for each heater/chiller system will

remain at a maximum temperature below the oil flash point temperature (with sufficient

margin). High temperature switch cutout controls for individual units are set below the

flash point.

" Oxygen gas (oxidizer), acetylene gas, and propane gas in the TSB workshops and

laboratories;

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" Acetone, toluene, petroleum ether, and petroleum ether in the TSB laboratories; and

" Primus gas, degreaser solvent, penetrating oil, and cutting oil in the TSB workshops.

7.3.4.2 Process Fire Safety Conclusions

Consistent with the acceptance criteria from Section 7.4.3.4 of NUREG-1 520 (NRC, 2002), the

application identifies the hazardous chemicals, processes, and design standards used to ensure

safety in areas that have fire hazards that may threaten licensed material, and is, therefore,

acceptable. There were no fire safety IROFS directly inherent to the process design. The fire

safety IROFS used to protect the licensed material within the processes are analyzed in the

individual discussions in Section 7.3.2 of this chapter. In its review, the staff has taken into

account the potential presence of identified combustibles in the various accident scenarios. The

identification of fire hazards and related analyses are documented in the applicant's ISA

Summary (AES, 2009b and AES, 2010) that provides the supporting safety basis for the SAR

(AES, 2009a).

7.3.5 Fire and Emergency Response

The facility will maintain a fire brigade consisting 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, 2005a) 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 offsite assistance organization personnel in the facility

emergency training procedures. Facility emergency response personnel meet at least annually

with each offsite assistance group to accomplish training and review items of mutual interest,

including relevant changes to the program. The primary agency that will be available for this

response is the City of Idaho Falls, Idaho Fire Department. This agency is a signatory to the

Bonneville County Mutual Aid agreement and can request assistance from other signatory

agencies including adjacent municipal fire departments and the fire/emergency response

services of the US Department of Energy's Idaho National Laboratory as warranted. The

applicant has received a letter from the Bonneville County Fire Protection District 1 (which

includes response by the Idaho Falls Fire Department) including a commitment to fire protection

and emergency response to the Eagle Rock Enrichment Facility (EREF) (AES, 2009a). A copy

of this letter is included in Attachment 2 to the EREF Emergency Plan (AES, 2009a). The

training and conduct of emergency drills is discussed in the EREF Emergency Plan and defines

the fire protection and emergency-response commitments between the organizations. The

Idaho Falls Fire Department is comprised of a roster of approximately 100 paid personnel-24

response personnel per shift, staffing five fire stations in a three-shift rotation. The department

has five front-line engine companies (pumpers) and five in reserve, one 30 m (100 ft)

telescoping platform, one heavy rescue truck and four light duty rescue/wildland trucks, three

water tenders (6813 L [1800 gal], 11,356 L [3000 gal], and 12,114 L [3200 gallon]) tankers, one

hazmat response vehicle, several command vehicles and ten ambulances equipped to provide

advanced level life support. Six ambulances are staffed per shift with four in reserve.

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The estimated response time to the EREF for a basic life support ambulance is 26 minutes with

a second ambulance available within an additional 1-3 minutes. The EREF personnel will be

trained and equipped to provide first aid and circulatory/respiratory support in the interim (e.g.,

provide cardiopulmonary resuscitation (CPR), apply automatic external defibrillation, and

administer oxygen). The estimated response time to EREF for a structural fire engine and a full

structural crew is 28 minutes with a second engine company within an additional 1-3 minutes.

The initial response for a structural first alarm would be three engines, a rescue truck, an

ambulance, and a staff officer. In the event of a fire, the EREF fire brigade will respond and the

Idaho Falls Fire Department will be notified to respond.

The NRC staff concludes that the onsite fire brigade, onsite water supply, onsite hose lines, and

mutual aid from adequately equipped fire departments can: adequately provide defense-indepth protection from releases from all identified and credible fire scenarios, satisfy the

requirements of 10 CFR 70.64(b), and are in accordance with the guidance in

NUREG-1520 (NRC, 2002).

Consistent with the acceptance criteria from Section 7.4.3.5 of NUREG-1 520 (NRC, 2002), the

SAR (AES, 2009a) documents the fire protection systems and fire emergency response

organizations that are provided and is, therefore, acceptable. In addition, the applicant's

emergency response capability meets the requirements of 10 CFR 30.33, 40.32, 70.22, 70.61,

70.64, and 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 that the facility also meets the

requirements of 10 CFR 30.33, 10 CFR 40.32, 10 CFR 70.22, 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

(AES, 2009a) AREVA Enrichment Services, "Revision 1 to the License Application for the

Eagle Rock Enrichment Facility," April 23, 2009.

(AES, 2009b) AREVA Enrichment Services, "Eagle Rock Enrichment Facility ISA Summary,"

April 23, 2009.

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(AES, 2010) AREVA Enrichment Services, "Eagle Rock Enrichment Facility ISA Summary,"

April 23, 2009.

(NFPA, 2004) National Fire Protection Association, NFPA 70E, "Standard for Electrical Safety

in the Workplace," 2004.

(NFPA, 2005a) National Fire Protection Association, NFPA 600, "Standard on Industrial Fire

Brigades," 2005.

(NFPA, 2005b) National Fire Protection Association, NFPA 1410, "Standard on Training for

Initial Emergency Scene Operations," 2005.

(NFPA, 2006) National Fire Protection Association, NFPA 220, "Standard on Types of Building

Construction," 2006.

(NFPA, 2007a) National Fire Protection Association, NFPA 10, "Standard for Portable Fire

Extinguishers," 2007.

(NFPA, 2007b) National Fire Protection Association, NFPA 13, "Standard for the Installation of

Sprinkler Systems," 2007.

(NFPA, 2007c) National Fire Protection Association, NFPA 20, "Standard for Installation of

Stationary Pumps for Fire Protection," 2007.

(NFPA, 2007d) National Fire Protection Association, NFPA 72, "National Fire Alarm Code,"

2007.

(NFPA, 2008a) National Fire Protection Association, NFPA 22, "Standard for Water Tanks for

Private Fire Protection," 2008.

(NFPA, 2008b) National Fire Protection Association, NFPA 70, "National Electric Code," 2008.

(NFPA, 2008c) National Fire Protection Association, NFPA 801, "Standard for Fire Protection

for Facilities Handling Radioactive Materials," 2008.

(NFPA, 2008d) National Fire Protection Association, NFPA 2001, "Standard on Clean Agent

Fire Extinguishing Systems," 2008.

(NRC, 2002) U.S. Nuclear Regulatory Commission, NUREG-1 520, "Standard Review Plan for

the Review of a License Application for a Fuel Cycle Facility," March 2002.

(NRC, 2004) U.S. Nuclear Regulatory Commission, NUREG-1 805, "Fire Dynamics Tools

(FDTs) Quantitative Fire Hazard Analysis Methods for the U.S. Nuclear Regulatory Commission

Fire Protection Inspection Program," December 2004.

(NRC, 2005) U.S. Nuclear Regulatory Commission, "Confirmatory Calculations for Fire

Protection Review of National Enrichment Facility Integrated Safety Analysis (ISA) Summary,"

March 2005.

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CHAPTER 8.0 EMERGENCY MANAGEMENT

The purpose of reviewing the AREVA Enrichment Services's (AES's) Emergency Plan (EP)

(AES, 2009) is to determine if AES has established adequate emergency management facilities

and procedures to protect workers, the public, and the environment.

8.1 Regulatory Requirements

The regulations in Title 10 of the Code of Federal Regulations (CFR) 70.22(i)(1)(ii) require

applicants to submit an EP if they are requesting authorization to possess (1) enriched uranium or

plutonium for which a criticality accident alarm system is required, (2) uranium hexafluoride in

excess of 50 kilograms (kg) (110 pounds[ lb]) in a single container or 1,000 kg (2,200 lb) total or (3)

plutonium in excess of 2 curies (Ci) in unsealed form or on foils or plated sources and their

evaluation shows that the maximum dose to a member of the public offsite from a release of

radioactive materials would exceed 0.01 Sv (1 rem) effective dose equivalent or an intake of 2

milligrams of soluble uranium. The regulatory requirements for the information that must be

included in an EP are outlined in 10 CFR 70.22(i)(3). In addition, 10 CFR 70.64(a)(6) requires

applicants to address baseline design criteria, including emergency capability to maintain

control of licensed material and hazardous material produced from licensed material, evacuation

of on-site personnel, and onsite emergency facilities and services that facilitate the use of

available offsite services.

8.2 Regulatory Acceptance Criteria

The acceptance criteria for the NRC's review of the emergency management plan are outlined

in Section 8.4.3 of NUREG-1 520, "Standard Review Plan for the Review of a License

Application for a Fuel Cycle Facility" (NRC, 2002).

8.3 Staff Review and Analysis

8.3.1 Facility Description

Section 1.0 of the EP (AES, 2009) contains descriptions of the licensed activity, the facility and

site, and the area near the site. The information provided includes:

1. A description of the enrichment process;

2. A discussion of chemicals of concern, including form, physical state, location, and

quantity (uranium hexafluoride (UF 6) has been identified as the only direct chemical of

concern to be used at the facility);

3. A detailed description of the site location and layout;

4. A description of the major structures to be located at the site;

5. A description of the ventilation systems, including stack heights, maximum flow rates and

filter efficiencies;

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6. A description of the area near the site, including area land-use information;

7. Demography of the area, water use, and climate; and

8. Detailed maps of the facility and surrounding area.

As described in Section 10 of the EP (AES, 2009), the applicant will comply with the Emergency

Planning and Community Right-to-Know Act of 1986, in accordance with

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; advising or training

construction and operating personnel regarding the marking, storage, and location of all

hazardous chemicals; and reporting accidental releases of these substances. This meets the

requirements of 10 CFR 70.22(i)(3)(i).

8.3.2 Onsite and Offsite Emergency Facilities

Section 6.0 of the EP (AES, 2009) contains descriptive information regarding the emergency

response equipment and facilities. The primary Emergency Operations Center (EOC) is in the

Control Room, which is located in the Operational Support Building. The Eagle Rock

Enrichment Facility (EREF) EOC controls communications to all principal points within and

outside the facility. The EREF EOC contains current, as-built drawings; procedures; and

operational engineering information to assist in routine operations and in emergency responses.

The EREF EOC supports the following functions:

1. Assessment of abnormal conditions;

2. Determination of offsite protective action recommendations (PARs);

3. Emergency Response Organization (ERO) notification;

4. Notifications to county, State, and Federal emergency response authorities;

5. Direction and performance of accident mitigation;

6. Direction of facility operations; and

7. Implementation of onsite protective actions.

The backup EREF EOC is in the Security and Secure Administration Building, which is the

principal entry area for the facility. Depending on the nature and location of the emergency

situation, or if the Control Room becomes uninhabitable, the Emergency Director may move the

emergency response personnel to the backup EREF EOC. The same documentation that is

available in the primary EREF EOC is also available in the backup EREF EOC. The applicant

describes these two EREF EOCs as being located in different and physically separated

buildings, and it is very unlikely that both areas would be unavailable simultaneously.

Offsite emergency support and equipment are described in Section 4.3 of the EP (AES, 2009).

This section provides information regarding fire, emergency medical services, and local law

enforcement. A Memorandum of Understanding (MOU) has been established between the

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EREF and the Bonneville County Fire Protection District 1 (including the Idaho Falls Fire

Department) for fire and medical emergency services. MOUs have also been established with

the Bonneville County Sheriffs Office for law enforcement services and Eastern Idaho Regional

Medical Center for medical treatment facility services.

Section 6.4 of the EP (AES, 2009) describes the emergency monitoring equipment that is

available including: (1) personnel monitoring equipment; (2) liquid effluent monitors;

(3) gaseous effluent monitors; (4) hydrogen fluoride monitors; and (5) a meteorological

measurement system for wind speed, direction, temperature and humidity.

Emergency equipment will be inventoried and tested on a quarterly basis, as discussed in

Section 7.6 of the EP (AES, 2009). Deficiencies identified will be reported to the Safety,

Security and Emergency Preparedness Manager, who will ensure that timely corrective action is

taken.

Section 6.2 of the EP (AES, 2009) describes the communications systems, which include facility

telephones that have facsimile, call tracing or call recording capabilities; the public address

system; radios; and site alarms. The communications systems are designed with redundant

devices for emergency communications so that a failure in one system does not leave the

facility without communications capability. Cellular and satellite telephones are used to

supplement and back up the primary telephone systems.

The radio systems that support the in-facility response have multichannels and high/low band

capabilities to support offsite communications systems. These bandwidths provide the

capability to communicate with the hospital; ambulance services; fire department; State, county,

and local law enforcement agencies; and the Idaho Bureau of Homeland Security (BHS). The

radio base stations are powered by diesel-backed electrical sources and remain operative

following loss of offsite power. This meets the requirements of 10 CFR 70.64(a)(6)(iii).

8.3.3 Types of Accidents

Section 2.0 of the EP (AES, 2009) identifies postulated events that have high and intermediate

consequences. Accident sequences, as well as mitigating and preventive measures, are also

described. Since the EREF operates with only natural and low enriched (i.e., no reprocessed)

uranium in the form of UF6, there are no radiological hazards associated with the operation that

could likely result in any significant offsite radiation doses. The only significant impact to the

public safety is that associated with the potential release of UF6 to the atmosphere. The

applicant states that the possibility of a nuclear criticality incident occurring at the EREF has

been determined to be highly unlikely. To prevent or limit the impact of any chemical or

radiological release, the facility has been designed with operational safeguards appropriate to a

modern chemical plant. The licensee performed a consequence analysis 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, which is contained in AES's license application. This meets the

requirements of 10 CFR 70.22(i)(3)(ii).

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8.3.4 Classification of Accidents

AES has established Emergency Action Levels consistent with Appendix A, "Examples of

Initiating Conditions," to NRC Regulatory Guide 3.67, "Standard Format and Content for

Emergency Plans for Fuel Cycle and Materials Facilities" (NRC. 1992). Section 3.1 of the EP

(AES, 2009) explains the process used to classify an emergency as either an Alert or a Site

Area Emergency declaration, 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

declaration 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 and to protect persons offsite from reaching the exposure limits

set forth in 10 CFR 70.22(i)(1)(i). This process for classifying events is acceptable to the staff

and meets the requirements of 10 CFR 70.22(i)(3)(iii). The processes for making the

appropriate classification will be provided in the Emergency Plan Implementing Procedures.

Table 3.1-1 in the EP (AES, 2009) provides examples of site-specific incidents and the

emergency classification that will be declared for each event. The Production Supervisor is

responsible for accident classification and assumes the responsibility of the Emergency Director

until relieved by the Plant Manager or his designee. This meets the requirements of 10 CFR

70.22(i)(3)(iii).

8.3.5 Detection of Accidents

Sections 2.2 and 6.4 of the EP explain the methods and systems available to detect accidents

at the facility, including:

1. Visual observation of fire or UF6 release,

2. Liquid and gaseous effluent monitors,

3. Hydrogen Fluoride monitors,

4. Various pressure and temperature monitors,

5. Criticality event alarm systems, and

6. Automatic fire and smoke detectors.

Action taken, in response to accidents, will be outlined in detailed procedures; and

implementation will be directed by the Emergency Director. This meets the requirements of

10 CFR 70.22(i)(3)(iv).

8.3.6 Mitigation of Consequences

Section 5.3 of the EP (AES, 2009) describes actions and equipment that will be used to mitigate

the consequences of accidents at the facility. 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 due to a UF6 release or a criticality event, include automatic interruption or

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termination of specific operations, fire detection and suppression systems, operator response to

abnormal conditions/alarms, and shutdown of the ventilation system. This meets the

requirements of 10 CFR 70.22(i)(3)(v).

8.3.7 Assessment of Releases

Section 5.2 of the EP (AES, 2009) describes the actions that will be taken to assess the extent

of an accident at the facility. In case of an Alert declaration, dose projections of offsite radiation

and hazardous material exposures will be made and provided to offsite emergency response

agencies. Environmental monitoring and sampling of areas off of the EREF site will be

coordinated between Bonneville County and BHS through Annex C, "Radiological Incident

Response," of the Bonneville County Emergency Operations Plan-and the Idaho Hazardous

Materials Incident Command and Response Support Plan. Projections of offsite radiation

exposures will be based on the estimated amount 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 software. This meets the

requirements of 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 (AES, 2009). In case of

an emergency, the Production Supervisor assumes the duties of the Emergency Director until

the Plant Manager or a designee arrives. In 4.2.1 of the EP (AES, 2009), the applicant states

that the position of Production Supervisor is filled at all times during normal operations and

backshifts. As stated in Section 4.2.1 of the EP, the Emergency Director coordinates the

response effort. Specific responsibilities of the Emergency Director include:

Non-Delegable Responsibilities

1. Declaring or terminating emergency events;

2. Notifying State, county and Federal emergency response authorities (the Emergency

Director is responsible for notification accuracy and timeliness; however a communicator

may actually make the notifications); and

3. Providing PARs to authorities responsible for implementing offsite emergency measures.

Delegable Responsibilities

1. Determining onsite protective actions;

2. Directing the assessment of actual or potential consequences, both onsite and offsite

throughout the event;

3. Requesting support from offsite agencies; and

4. Approving press statements prior to their release.

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Section 4.2.2.2 of the EP summarizes the responsibilities of the remaining on-call ERO. Section

4.3 provides the local offsite assistance to the facility for fire, emergency medical services, and

local law enforcement. MOUs have been established with local agencies, and will be reviewed

annually and renewed, if necessary. This meets the requirements of 10 CFR 70.22(i)(3)(vii).

8.3.9 Notification and Coordination

As discussed in Section 8.3.4 of this Safety Evaluation Report, the classification of emergencies

is outlined in Section 3.1 of the EP (AES, 2009) and is the responsibility of the Emergency

Director. Section 3.2 of the EP (AES, 2009) provides a clear commitment to promptly notify

offsite EROs of an emergency, including the notification to the NRC Operations Center within 60

minutes. Sections 4.3 and 4.4 of the EP (AES, 2009) provide an adequate description of

provisions for assistance from offsite emergency response organizations. These sections

adequately describe the agreements held between AES and local offsite EROs and agencies,

procedures for the access to the site for these response organizations, and equipment and

services available from these organizations.

The Offsite EOC Liaison is dispatched to the Bonneville County EOC when activated and

provides event and facility response details to county authorities. The Offsite Incident

Command Liaison is dispatched to the near-site Command Post when one is established and

will provide facility layout, event status, and response details to the Incident Commander. Also,

mutual aid agreements are in place to provide additional support if other services or equipment

is not available.

As discussed previously, it is the responsibility of the Emergency Director to: (1) declare an Alert

or Site Area Emergency, (2) activate the onsite ERO, (3) notify offsite emergency response

authorities of an emergency, (4) notify the NRC Operations Center, (5) decide which onsite

protective actions to initiate, (6) decide which offsite protective actions to recommend and

provide that information to offsite emergency response authorities, (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 70.22(i)(3)(viii).

8.3.10 Information to be Communicated

Section 3.3 of the EP (AES, 2009) provides an adequate description of the type of information to

be given to offsite EROs during an emergency. AES will use the EREF Event Notification Form

as a script for initial notification of an emergency at the facility to appropriate offsite emergency

response authorities. The emergency notification form or an NRC Form 361A, "Fuel Cycle and

Materials Event Notification Worksheet," may serve as the basis for follow-up messages to the

offsite emergency response authorities. If a Site Area Emergency is declared, PARs for the

public are provided to the State and county agencies. Specific recommendations would depend

on the event in progress (e.g., amount of UF6 released, concentration of UF 6 expected offsite,

and meteorological conditions). In most cases, recommendations would involve avoiding the

area of the facility until the event concludes. This meets the requirements of 10 CFR

70.22(i)(3)(ix).

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8.3.11 Training

Section 7.2 of the EP (AES, 2009) describes the training AES 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 in 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 to provide specific information about how the ERO responds during emergency

conditions; including staffing, determining and estimating potential offsite releases of radiation

and chemicals, and interfacing with offsite assistance organizations. This training is required

before an individual is assigned to the emergency organization, and refresher training is

provided at least once every year.

The nuclear safety training program includes: (1) instructions to workers, (2) practices designed

to keep radiation exposures As Low As Reasonably Achievable, (3) contamination control

methods, (4) use of monitoring equipment, (5) emergency procedures and actions, (6) nature

and sources of radiation, (7) safe use of chemicals, (8) biological effects of radiation, (9) use of

personnel monitoring devices, (10) principles of nuclear criticality safety, (11 ) risk to pregnant

females, (12) radiation protection practices, (13) protective clothing, (14) respiratory protection,

and (15) personnel surveys.

Specific topics, performance objectives, content, training schedules, and the number of training

hours required for each position are contained in the administrative procedures. The contents

of the formal nuclear safety training programs are reviewed and updated as required at least

every 2 years by the Environmental Health, Safety, and Licensing Manager-or his or her

designee-to ensure that the programs are current and adequate. Individuals requiring

unescorted access to the Controlled Area receive annual retraining.

Facility tours and classroom training are also provided to offsite response organizations. The

training includes: (1) information concerning facility access control, (2) potential accident

scenarios, (3) emergency action levels, (4) notification procedures, (5) exposure guidelines,

(6) personnel monitoring devices, (7) communications, (8) contamination control, and (9) the

role of the offsite assistance organizations in responding to an emergency at the facility.

Physicians associated with the Eastern Idaho Regional Medical Center are offered annual

emergency training involving the transportation and treatment of radiologically or chemically

contaminated patients. This meets the requirements of 10 CFR 70.22(i)(3)(x).

8.3.12 Safe Shutdown (Recovery and Facility Restoration)

Section 9.0 of the EP (AES, 2009) states that during an emergency, immediate action is

directed towards limiting the consequences of the incident so as to afford maximum protection

to facility personnel and the general public. Once corrective measures have been taken and

effective control of the facility has been reestablished, a systematic and planned approach to

normal operations is taken. Depending on the type and extent of the event, the Emergency

Director may elect to transition to a Recovery Organization, rather than reverting to the normal

day-to-day organizational structure. The purpose of establishing a Recovery Organization is to

provide dedicated personnel and structure to the handling of repair, restoration, and recovery

planning and activities.

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Section 9.1 of the EP (AES, 2009) states that the following criteria are to be considered prior to

terminating from an emergency event:

1. Conditions requiring emergency classification no longer exist;

2. Any fire, flood, earthquake or similar initiating event no longer exists;

3. The extent of damage and condition of facility systems and equipment is understood;

4. Radioactive and/or hazardous chemical releases have been controlled, such that further

incidents will be prevented;

5. Environmental assessment activities in progress are minor and only necessary to determine

the extent of impact from the event, not the active tracking of material still being released;

6. Facility radiation, contamination, and hazardous chemical levels are stable or decreasing

and acceptable, given current conditions;

7. All required notifications have been made;

8. The EREF EOC was staffed and activated; and

9. Offsite conditions do not unreasonably limit access of outside support to the facility.

Section 9.1 further discusses that it is not necessary that all criteria listed above be met to

terminate from an emergency event; however, they all must be considered. For example, it is

possible that some conditions may remain that exceed an emergency action level, but the state

of emergency no longer exists. This meets the requirements of 10 CFR 70.22(i)(3)(xi).

8.3.13 Exercises and Drills

Section 7.3 of the EP (AES, 2009) provides adequate provisions for drills and biennial exercises

that are used to test the adequacy of procedures, emergency equipment and instrumentation,

and to ensure that all emergency response personnel are familiar with and proficient in their

duties. Section 7.4 of the EP provides that post-drill and post-exercise evaluations will be

conducted by those involved, and appropriate improvements will be implemented as deemed

necessary by the licensee. Areas evaluated include the adequacy of the EP, procedures,

equipment, facilities, personnel training, and overall response effectiveness. The Safety,

Security and Emergency Preparedness Manager is responsible for the planning, scheduling,

and conducting of emergency response drills and exercises for the facility.

Offsite EROs are invited to participate in the biennial exercise, and the NRC is invited to

participate or observe. Exercise objectives and scenarios will be submitted to the NRC for

review and comment at least 60 days before the exercise. Section 7.3 of the EP (AES, 2009)

includes an adequate provision for quarterly communications checks to verify the operability of

equipment used by the ERO to communicate with offsite agencies and response organizations.

This meets the requirements of 10 CFR 70.22(i)(3)(xii).

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8.4 Evaluation Findings

The NRC staff has evaluated AES's EP (AES, 2009) for the facility. The licensee has

established an EP for responding to the radiological hazards resulting from a release of

radioactive material or hazardous chemicals relating to the processing of licensed material in

accordance with 10 CFR 70.22(i)(1)(ii). The NRC staff reviewed AES's EP with respect to 10

CFR 70.22(i)(3), 70.64(a)(6), and the acceptance criteria in Section 8.4.3 of NUREG-1520. The

NRC staff concluded that AES's EP is adequate to demonstrate compliance with the regulatory

requirements, 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 county government agencies and to recommend

appropriate protective actions to protect members of the public; and (5) necessary recovery

actions have been established to return the facility to a safe condition after an accident.

The requirements of the EP are implemented through approved written procedures. Changes

that decrease the effectiveness of the EP may not be made without NRC approval. The NRC

will be notified of other changes that do not decrease the effectiveness of the EP within

6 months of making the changes.

8.5 References

(AES, 2009) AREVA Enrichment Services LLC, "Emergency Plan for the Eagle Rock

Enrichment Facility," 2009.

(NRC, 1992) U.S. Nuclear Regulatory Commission, Regulatory Guide 3.67, "Standard Format

and Content for Emergency Plans for Fuel Cycle and Materials Facilities," 1992.

(NRC, 2002) U.S. Nuclear Regulatory Commission, NUREG-1520, "Standard Review Plan for

the Review of a License Application for a Fuel Cycle Facility," 2002.

8-9

CHAPTER 9.0 ENVIRONMENTAL PROTECTION

The purpose of the U.S. Nuclear Regulatory Commission's (NRC's) review of the applicant's

environmental protection plan for the Eagle Rock Enrichment Facility (EREF) is to determine

whether the applicant's proposed environmental protection measures are adequate to protect

the environment and the health and safety of the public, as required by Title 10 of the Code of

Federal Regulations (10 CFR) Parts 20, 30, 40, 51, and 70.

9.1 Regulatory Requirements

For its application to be considered acceptable, the applicant, AREVA Enrichment Services LLC

(AES), 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. 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.7

7

On March 17, 2010, the NRC granted an exemption to 10 CFR 30.33(a)(5), 10 CFR 10.43(e), and

10 CFR 70.23(a)(7) (Agencywide Documents Access and Management System (ADAMS) Accession No.

ML093220446) authorizing AES to conduct certain preconstruction activities, including clearing land, grading, and

excavation, on the proposed EREF site.

9-1
5. 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

background information related to the suitability of the site or the protection of environmental

values."

6. Part 51 specifies that the applicant must submit an Environmental Report (ER) for

construction and operation of a uranium enrichment facility, as required by

10 CFR 51.60(b)(1)(vii).

7. 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.

8. 10 CFR 70.59 outlines the radiological effluent monitoring reporting requirements for a Part

70 licensee.

9. 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 guidance for NRC's review of the applicant's environmental protection

program is outlined in Section 9.4.3.2 of NUREG-1 520 (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 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 19.5 megabecquerel (MBq) (528 microcuries

[ptCi]) 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. Experience with similarly designed

European facilities has typically shown that quantities of radioactivity in air effluent were much

less than the estimated quantities, such that annual emissions are actually expected to be less

than 20 grams (0.71 ounces) which corresponds to 0.506 MBq (13.7 iCi) of uranium isotopes.

Therefore, the staff finds that a value of 19.5 MBq (528 liCi) uranium represents a reasonably

conservative upper bound on the annual quantity of air effluent from the facility.

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As noted in Table 4.12-22 of the applicant's ER, the CEDE to the maximally exposed

hypothetical member of the public (teen) located at the north-northeast side of the controlled

area boundary, resulting from the release to the atmosphere of 19.5 MBq (528 jlCi) of uranium

from gaseous release points, would be less than 0.9 microsieverts (1iSv) (0.09 [mrem])

(AES, 2009b) or 0.09 percent of the 1 mSv (100 mrem) limit on dose to the public in Part 20.

This estimated maximum public dose is also well below the 0.1 mSv (10 mrem) ALARA

constraint on air emissions described in 10 CFR 20.1101.

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. In Section 9.2 of the

license application, the applicant committed to meeting the 0.1 mSV (10 mrem) ALARA

constraint in 10 CFR 20.1101(d). In addition, the estimated dose to the maximally exposed

member of the public is a small fraction of the ALARA goals identified in the above-referenced

Regulatory Guides. The applicant's approach meets the guidance found in Section 9.4.3.2.1(1)

of NUREG-1 520 (NRC, 2002) that ALARA goals are set at a modest fraction of the values in

10 CFR Part 20 and, therefore, is acceptable to the staff.

Liquid Effluent ALARA Goal

The applicant estimated the maximum releases of uranium material from liquid effluents during

normal operations at the proposed facility. The applicant estimated that the maximum annual

quantity of radiological material in liquid effluent would be 900 becquerels (0.024 gCi) of

uranium. This effluent would be dispersed as an atmospheric release of distillate from the

evaporator in the Liquid Effluent Collection and Treatment System (AES, 2009b). The liquid

distillate is evaporated after various precipitating and filtering steps, and the final airborne

release is estimated as being less than 0.01 percent of the estimated release from gaseous

effluents. Because this adds negligible source term to the atmospheric release estimates, there

are negligible increases to the environmental or public radiological exposures resulting from

liquid effluents. The estimated maximum public dose from the liquid effluent source term (about

1/10,000 of the estimated exposure from gaseous effluents) is well below the 0.1 mSv

(10 mrem) ALARA constraint on liquid emissions described in 10 CFR 20.1101.

As noted in Section 8.7 of the applicant's ER, radioactive material may be released from the

EREF as the result of gaseous and liquid effluent discharges, including controlled releases from

the uranium enrichment process lines during decontamination and maintenance of equipment.

The CEDE to the maximally exposed member of the public (teen) located at the north side of

the controlled area boundary, resulting from the combined annual release to the atmosphere of

19.5 MBq (528 jICi) of uranium from the EREF, would be less than 0.9 p.Sv (0.09 mrem)

(AES, 2009b).

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. As stated in the previous section, the

applicant has committed to meeting the 0.1 mSv (10 mrem) ALARA constraint in 10 CFR

20.1101(d). In addition, the estimated dose to the maximally exposed member of the public is a

small fraction of the ALARA goals identified in the above-referenced Regulatory Guides. The

applicant's approach meets the guidance found in Section 9.4.3.2.1(1) of NUREG-1 520

(NRC, 2002) that ALARA goals are set at a modest fraction of the values in 10 CFR Part 20 and is,

therefore, acceptable to the staff.

9-3
9.3.1.2 Air Effluent Controls To Maintain Public Doses ALARA

In Section 4.12.2.1.5 of the applicant's ER (AES, 2009b), the applicant identified plant design

features developed to assure that radiological impacts to the environment and public are well

below regulatory limits. These include:

* Process systems that handle uranium hexafluoride (UF6) operate at sub-atmospheric

pressure to minimize outward leakage of UF6;

" UF6 cylinders are moved only when cool and when UF6 is in solid form to minimize the risk

of inadvertent release due to mishandling;

* Process off-gas from UF6 purification and other operations passes through desublimers to

solidify and reclaim as much UF6 as possible. Remaining gases pass through highefficiency filters and chemical absorbers to remove hydrogen fluoride (HF) and uranium

compounds;

" Gaseous effluent passes through pre-filters, high-efficiency particulate air (HEPA) filters, and

activated carbon filters, all of which greatly reduce the radioactive material in the final

discharged effluent to very low concentrations;

* Liquid waste is routed to collection tanks and treated through a combination of precipitation,

filtration, and evaporation to remove radioactive material prior to release of the distillate

vapors to the atmosphere; and

* Effluent paths are monitored and sampled to ensure compliance with regulatory discharge

limits.

Of primary importance in ensuring that air effluents are ALARA are the treatment systems for

the plant's ten Gaseous Effluent Ventilation Systems (GEVS) and the heating, ventilating, and

air conditioning (HVAC) systems for contaminated building exhaust. The GEVSs are identified

as: (1) the Separations Building Modules (SBMs) Safe-by-Design GEVS (one in each of the

four modules), (2) the Separations Building Modules Local Extraction GEVS (one in each of the

four modules), (3) the Technical Support Building (TSB) GEVS, and (4) the Centrifuge Test and

Post Mortem Facilities GEVS within the Centrifuge Assembly Building (CAB). The TSB; the

Blending, Sampling & Preparation Building (BSPB); and the Centrifuge Test and Post Mortem

Facilities have HVAC systems that function to maintain negative pressure and exhaust filtration

for rooms served by these systems.

GEVS

The function of each GEVS is to remove particulate matter containing uranium and HF from

potentially contaminated gas streams. Each GEVS includes ducts; prefilters; HEPA filters;

potassium carbonate impregnated activated carbon filters; fans; monitors and controls; inlet and

outlet isolation dampers; and a discharge stack or vent.

The SBM Safe-by-Design GEVS sub-atmospheric duct system transports potentially

contaminated gases to a set of redundant filters (pre-filter, HEPA filter, potassium carbonate

impregnated activated carbon filter, a final HEPA filter) and fans. The SBM Local Extraction

9-4
GEVS collects potentially contaminated gaseous effluent from local flexible hose connections

that are used during cylinder connection and disconnection and maintenance activities. The

cleaned gases are discharged via SBM rooftop exhaust vents to the atmosphere.

The TSB GEVS transports potentially contaminated gases to a set of redundant filters (pre-filter,

HEPA filter, potassium carbonate impregnated activated carbon filter, a final HEPA filter) and

fans. The cleaned gases discharge via exhaust vents on the TSB roof. The Centrifuge Test

and Post Mortem Facilities GEVS has one set of filters (pre-filter, HEPA filter, potassium

carbonate impregnated activated carbon filter, a final HEPA filter) and a single fan. It

discharges cleaned gases via exhaust vents on the roof of the CAB.

HVAC - Potentially Contaminated Areas

The TSB Contaminated Area HVAC system has two active sets of filters (roughing filter, HEPA

filter, potassium carbonate impregnated activated carbon filter, a final HEPA filter) and fans.

The Ventilated Room HVAC System in the BSPB and Centrifuge Test and Post Mortem

Facilities Exhaust Filtration HVAC System each have one set of filters (roughing filter, HEPA

filter, potassium carbonate impregnated activated carbon filter, a final HEPA filter) and one fan.

The TSB Contaminated Area HVAC System discharges cleaned gases via exhaust vents on the

roof of the TSB. The Ventilated Room HVAC System discharges cleaned gases via an exhaust

vent on the roof of the BSPB. The Centrifuge Test and Post Mortem Facilities Exhaust Filtration

System discharges cleaned gases via exhaust vents on the roof of the CAB.

Staff Evaluation of Air Effluent Controls

The staff evaluated the air effluent controls, including the GEVS and HVAC systems. 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's controls will ensure that radiation levels to the public

will remain well below regulatory limits in 10 CFR 20.1301 and ALARA air effluent goals, that

the applicant's approach to effluent controls are meets the guidance found in Section

9.4.3.2.1(2) of NUREG-1 520 (NRC, 2002) that the applicant describe and commit to use effluent

controls to maintain public doses ALARA, and, therefore, the controls are acceptable to staff.

9.3.1.3 Liquid Effluent Controls to Maintain Public Doses ALARA

The Liquid Effluent Collection and Treatment System for the EREF includes two stages of

precipitation and filtration to remove uranic material contained in liquid effluents collected from

plant processes. The final process stage of evaporation releases the resulting distillate steam

directly to the atmosphere without condensing vapor out of the air stream.

The proposed liquid effluent control systems will significantly reduce the quantity of uranium

from 114 kilograms per year (kg/yr) entering the system to less than 0.036 grams per year (g/yr)

9-5
in evaporated distillate. The uranium removed by the liquid effluent control will be disposed

offsite at a licensed disposal facility. As described earlier in Section 9.3.1.1 of this report, the

dose to the maximally exposed offsite member of the public from the evaporation of liquid

effluent distillate is about 1/10,000 of the estimate for all gaseous effluents. This dose is much

less than 1 percent of the regulatory limit found in 10 CFR 20.1301. A reasonable initial ALARA

goal for liquid effluent described in Regulatory Guide 8.34 (NRC, 1992) is 10-20 percent of the

regulatory limit.

Staff Evaluation of Liquid Effluent Controls

As stated above, the applicant does not anticipate any liquid discharges of licensed radioactive

materials from the proposed facility, and the processing of liquid waste will result in a release to

the atmosphere that is only a fraction of the gaseous effluent release. Therefore, based on the

above, the staff finds that the applicant's controls will ensure that radiation levels to the public

will remain well below the regulatory limits in 10 CFR 20.1301 and ALARA liquid effluent goals.

The applicant's approach to effluent controls meets the guidance found in Section 9.4.3.2.1(2)

of NUREG-1 520 (NRC, 2002) that the applicant describe and commit to use effluent controls to

maintain public doses ALARA, and, therefore, are acceptable to the staff.

9.3.1.4 ALARA Reviews and Reports to Management

In Section 4.2 of its Safety Analysis Report (SAR) (AES, 2009a), the applicant describes an

ALARA program for the proposed facility. The ALARA program would include annual reviews of

the content and implementation of the radiation protection program, including the effluent control

program. The Radiation Protection/Chemistry Manager is responsible for implementing the

ALARA program and ensuring that adequate resources are committed to make the program

effective. The Radiation Protection/Chemistry Manager prepares an annual ALARA program

evaluation report. As described in Section 4.2 of the SAR, 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 (AES, 2009a). Copies of the report are submitted to the AES President, Plant

Manager, Radiation Safety Committee, and the Safety Review Committee. This approach

adequately addresses the regulatory requirements in 10 CFR 20.1101(c) and is consistent with

the guidance in Section 9.4.3.2.1(3) of NUREG-1520 (NRC, 2002) and, therefore, it is

acceptable to the staff.

9.3.1.5 Waste Minimization

In Section 4.13.5 of the ER (AES 2009b), 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) a decontamination workshop designed to remove radioactive contamination from equipment

and allow some equipment to be reused rather than treated as waste, (b) closed-loop cooling

systems have been incorporated in the design to reduce water usage, (c) outer packaging

associated with consumables will be removed prior to use in a contaminated area, (d) collected

waste will be volume reduced at a centralized waste processing facility, and (e) use of glove

boxes to minimize the spread of contamination.

9-6
A contributing program to the waste minimization effort is the facility contamination control

program which is addressed in Chapter 4. Contamination levels will be reviewed by the Safety

Review Committee (SRC), which fulfills the functions of the ALARA Committee, during its

routine meetings. The SRC will also evaluate assessments made by the radiation protection

organization and make recommendations which are tracked to completion through the facility's

corrective action program.

These features and systems are consistent with the regulatory requirements in

10 CFR 20.1406; the guidance provided in NRC's Information Notice 94-23, "Guidance to

Hazardous, Radioactive, and Mixed-Waste Generators on the Elements of a Waste

Minimization Program" (NRC, 1994), and the guidance in Section 9.4.3.2.1(4) of NUREG-1 520

(NRC, 2002), and are, therefore, acceptable to the staff.

9.3.1.6 Safe Handling of Radioactive Wastes

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 Sections 9.3.1.2 and 9.3.1.3 of this

document.

In Section 3.12.2 of the applicant's ER (AES 2009b), the applicant has described a solid waste

management program at the proposed facility for industrial (non-hazardous), radioactive mixed,

and hazardous wastes. Solid waste will be grouped into one of these waste categories and, in

addition, radioactive and mixed waste will be further segregated according to the quantity of

liquid that is not readily separable from the solid material. The EREF may send wastes that are

candidates for volume reduction, recycling, or treatment to licensed treatment facilities that have

the ability to volume reduce most Class A low-level wastes (LLWs); and to process

contaminated oils and some mixed wastes. The applicant will also operate several systems for

reprocessing and recovery of uranic materials when practicable. The applicant does not

propose to process or treat onsite solid waste that does not contain economically recoverable

uranic materials.

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 Solid Waste Collection Room for inspection. Suitable waste will be volume

reduced and all radioactive waste will be 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 be collected at the point of

generation, transferred to the Solid Waste Collection Room, 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.

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As noted in Section 3.12.2.1.2.9 of the applicant's ER (AES, 2009b), the operation of the facility

would yield an annual production of 1,222 cylinders of depleted UF6 per year, or approximately

15,270 metric tonnes (16,832 tons). The Full Tails Cylinder (FTC) Storage Pad would have a

capacity of 33,638 cylinders. The NRC has evaluated the environmental impacts of alternatives

for disposition of depleted UF6 in its "Environmental Impact Statement for the Proposed Eagle

Rock Enrichment Facility in Bonneville County, Idaho" (NRC, 2010). During temporary storage

of this material on the FTC Storage Pad, the applicant will have 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 FTC Storage Pad; to

monitor cylinder integrity by conducting routine inspections for breaches; and to perform cylinder

maintenance, as needed. The cylinders will be stored on saddles made of materials that will not

cause corrosion of the cylinders. The storage array of the cylinders will permit easy visual

inspection. The cylinders will be surveyed for external contamination before being placed on

the FTC Storage Pad and before being transported offsite. In addition, the cylinders will be reinspected 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 and a root cause assessment of

the deterioration will be conducted.

On the basis of this analysis, the staff finds that the applicant's implementation of its program for

management of solid radiological and non-radiological wastes related to facility operation

reduces unnecessary exposures as per ALARA requirements in 10 CFR 20.1101(b) is

consistent with waste minimization guidance in Section 9.4.3.2.1(4) of NUREG-1520

(NRC, 2002) and, therefore, is acceptable to the staff.

9.3.2 Effluent and Environmental Monitoring

9.3.2.1 Air Effluent Monitoring

As described earlier in Section 9.3.1.1, the staff finds that the radioactive materials in airborne

effluents would result in exposures well below the ALARA constraint 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 in

Sections 4.12 and 8.7 of the ER (AES, 2009b) and determined that they are reasonable as

emissions were estimated from emissions from similarly designed plants. The staffs evaluation

of radiation exposures from normal operations is found in Chapter 4 (Section 4.2.10.2, "Facility

Operations") of the Draft Environmental Impact Statement (EIS) (NRC, 2010).

The applicant has identified all airborne effluent discharge locations, and will monitor discharges

from potentially contaminated processes and areas in accordance with 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" (NRC, 1985) and NUREG-1302, "Offsite Dose Calculation

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Manual Guidance: Standard Radiological Effluent Controls for Boiling Water Reactors"

(NRC 1991). The applicant will perform continuous monitoring at the exhaust vents for

alpha/beta emitting radioactivity and HF. The applicant will also perform monitoring of work

areas that process dispersible uranium, but 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, GEVSs, and exhaust filtration systems that discharge building

ventilation exhaust is provided in Table 9.3-1 of this report.

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 5x1 0-14 ýLCi/mL for weekly

gross alpha measurements and 5x10 1 4 ýtCi/mL for uranium isotopes. This value is less than

2 percent of the effluent concentration limit for class "D" solubility uranium materials in

10 CFR 20, Appendix B. Because the uranium materials in gaseous effluent from the facility are

expected to be class "D," this detection limit is acceptable to staff.

As noted in the applicant's ER (AES, 2009b), the applicant will develop 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:

1. The sample parameter's lowest alarm point will be selected based on conditions of

service, e.g., three times the normal background level.

2. The sample parameter exceeds any of the existing administrative limits.

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 ER (AES, 2009b), in addition to meeting NRC

requirements, the applicant will also obtain required Federal and State permits for hazardous air

pollutants.

As described in Section 6.1.2 of the applicant's ER (AES, 2009b), the applicant's reporting

procedures comply with the requirements of 10 CFR 70.59 and the specific guidance in

Regulatory Guide 4.16 (NRC, 1985).

On the basis of this analysis, including a review of the information provided by the applicant as

noted above, the staff finds that the applicant's air effluent monitoring during operation of the

9-9
facility will be adequate to ensure that radioactive materials in air effluent meet the regulatory

limits in 10 CFR.1301; is consistent with the guidance in Section 9.4.3.2.2(1) of NUREG-1 520

(NRC, 2002); and, therefore, is acceptable to the staff.

Table 9.3-1 Overview of Gaseous Effluent Monitoring Program

Sample Location Sample Type Analysis I Frequency

Separations Building GEVS Continuous air particulate filter Gross alpha/beta- Weekly

exhaust vents

Isotopic analysisd- Quarterly

TSB GEVS exhaust vent composite

TSB Contaminated Area

HVAC System exhaust vent

Centrifuge Test and Post

Mortem Facilities GEVS

exhaust venta

Centrifuge Test and Post

Mortem Facilities Exhaust

Filtration System exhaust

venta

Ventilated Room HVAC

System exhaust vent

Evaporator Continuous liquid condensate Gross alpha/beta - Weekly

sample from exhaust vent

Isotopic analysisd - Quarterly

composite

Process Areasb Local area continuous air Gross alpha/beta - Weekly

particulate filterc

Isotopic analysisd- Quarterly

composite

Non-Process Areasb Local area continuous air Gross alpha/beta -

particulate filterc

Quarterly composite

Notes:

a The continuous sampling system is operated only when the Centrifuge Test Facility or Post Mortem Facility

is in operation.

b A "Process Area" is any area of the facility where UF6 process flow between feed, product, or tails cylinders

occurs-including areas where cylinders containing UF6 are opened for testing, inspection, or sampling. A

"Non-Process Area" is any other area where uranic material is present in an open form.

c These will generally be collected with mobile continuous air monitors, as required to complement the effluent

monitoring program.

d Isotopic analysis for uranium if gross alpha and gross beta activities indicate that an individual radionuclide

could be present in a concentration greater than 10 percent of the concentrations specified in Table 2 of

Appendix B to 10 CFR Part 20.

9-10
9.3.2.2 Liquid Effluent Monitoring

As described in Section 9.3.1.2 of this SER, there are no expected liquid effluents to surface

water bodies. Instead, evaporated distillate will be released to the atmosphere while residual

evaporator liquids and precipitants are treated and disposed of offsite at a licensed facility. The

applicant will demonstrate compliance with gaseous effluent limits as discussed in

Section 9.3.2.1. This section also contains a discussion of the proposed action levels.

The staff reviewed the applicant's assumptions and conclusions used in its calculations (ER

Sections 4.12 and 8.7 [AES 2009b]) and determined that they are reasonable as the emission

estimates are based on similarly designed facilities. The staffs evaluation of radiation

exposures from normal operations is found in Section 4.2.10.2 of the Draft EIS (NRC, 2010).

General site stormwater runoff is routed to the Site Stormwater Detention Basin. The two

Cylinder Storage Pads Stormwater Retention Basins collects stormwater runoff from the

Cylinder Storage Pads (i.e., Full Feed Cylinder Storage Pads, Full Tails Cylinder Storage Pads,

Full Product Cylinder Storage Pad, and Empty Cylinder Storage Pads), as well as treated water

from the Domestic Sanitary Sewage Treatment Plant. The applicant proposes to include

sampling and analysis of water and sediment from each of the retention-detention basins in the

Radiological Effluent Monitoring Program.

As described in Section 1.3 of the applicant's ER (AES, 2009b), in addition to meeting the

NRC's regulatory requirements, the applicant will obtain required Federal and State permits

relating to liquid discharges (primarily related to groundwater protection). In addition to having

direct permitting authority, the State of Idaho also has oversight responsibility for the U.S.

Environmental Protection Agency's (EPA's) water permits granted through the Idaho

Environmental Protection and Health Act (Idaho Code Chapter 1, Title 39).

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 radioactive materials in liquid effluent will meet the limits in

10 CFR 20.1301; is consistent with guidance in Section 9.4.3.2.2(1) of NUREG-1520 (NRC,

2002); and the monitoring is, therefore, acceptable to the staff.

9.3.2.3 Laboratory Quality Control

The applicant has a Radiological Environmental Monitoring Program (REMP) with associated

quality assurance. The REMP provides data to confirm the effectiveness of effluent controls

and the effluent monitoring program. All environmental samples will be analyzed onsite for

facility-related radiological constituents. Samples may also be shipped to a qualified

independent laboratory for analyses.

The Quality Control (QC) procedures used by the laboratories performing the plant's REMP will

be used to validate the analytical results and will conform to guidance in Regulatory

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Guide 4.15, "Quality Assurance for Radiological Monitoring Programs (Inception through Normal

Operations to License Termination)-Effluent Streams and the Environment" (NRC, 1979).

These QC procedures include the use of established standards such as those provided by the

National Institute of Standards and Technology (NIST), as well as standard analytical

procedures such as those established by the National Environmental Laboratory Accreditation

Conference (NELAC).

Monitoring procedures will employ established analytical methods and instrumentation. The

instrument maintenance and calibration program will be in accordance with manufacturers'

recommendations. The onsite laboratory and any contractor laboratory used to analyze

samples will participate in third-party laboratory intercomparison programs appropriate to the

media and analytes being measured. All radiological and non-radiological laboratory vendors

will be certified by the National Environmental Laboratory Accreditation Program (NELAP) or an

equivalent state laboratory accreditation agency for the analytes being tested.

For the physiochemical monitoring program (described in the applicant's ER, Section 6.2.8), the

applicant has a quality assurance program that will use a set of formalized and controlled

procedures for sample collection, laboratory analysis, chain of custody, reporting of results, and

corrective actions. Samples sent to laboratories will include blanks and duplicates at specified

frequencies to provide data for identifying routine reporting or analytical errors as part of quality

assurance checks on the data. Analyses will only be performed at laboratories with appropriate

EPA and State of Idaho certifications. The laboratory analyses will be conducted using the best

available standard techniques at State or EPA-certified laboratories.

The staff finds that these procedures are adequate to validate the analytical results produced by

the REMP, which is described in the applicant's ER Section 6.1.2 (AES, 2009b). These

procedures are consistent with regulatory the guidance in Section 9.4.3.2.2(1 )(h) of

NUREG-1 520 (NRC, 2002); adequate to validate the analytical results from the laboratory; and

therefore, acceptable to the staff.

9.3.2.4 Environmental Monitoring

The applicant has established 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 for the environment and the public.

Compliance is demonstrated primarily through effluent monitoring (see applicant's ER

Section 6.1.2 [AES, 2009b]).

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 to develop a sufficient environmental baseline.

The types of samples to be collected under the REMP are listed in Table 9.3-3 of this report.

The staff's evaluation of the proposed corrective action levels, which apply to both effluent

monitoring and the REMP, is described in Section 9.3.2.1 of this report. The scope of the

applicant's REMP meets the environmental monitoring criteria found in NUREG-1520

(NRC, 2002), Section 9.4.3.3.2(2). As noted in the applicant's ER (Section 6.1.2 [AES, 2009b]),

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the REMP sampling locations are based on NRC guidance found in NUREG-1 302, "Offsite

Dose Calculation Manual Guidance: Standard Radiological Effluent Controls for Boiling Water

Reactors," (NRC, 1991).

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 the regulatory requirements in 10 CFR 20.1406, the

guidance found in Section 9.4.3.3.2(2) of NUREG-1 520 (NRC, 2002) and the program is,

therefore, acceptable to the staff.

Table 9.3-2 Types of Samples Collected under the REMP

Sample Minimum Number Sampling and Collection Type of Analysis

Type/Location of Sample Frequency

Locations

Continuous Airborne 5 Continuous operation of air Gross beta/gross

Particulate sampler with sample alpha analysis each

collection as required by dust filter change.

loading but at least biweekly. Quarterly isotopic

Quarterly composite samples analysis on

by location, composite sample.

Vegetation 9 1 to 2-kg (2.2 to 4.4-1b) Isotopic analysisa

samples collected

semiannually

Groundwater 10 4-L (1.06-gal) samples Isotopic analysisa

collected semiannually

Basins 1 from each of 3 4-L (1.06-gal) water sample/1 Isotopic analysisa

basinsb to 2-kg (2.2 to 4.4-1b)

sediment sample collected

quarterly

Soil 9 1 to 2-kg (2.2 to 4.4-1b) Isotopic analysisa

samples collected

semiannually

Domestic Sanitary 1 1 to 2-kg (2.2 to 4.4 Ib) solid Isotopic analysisa

Sewage Treatment fraction sample semiannually

Plant II

TLD 18 Quarterly Gamma and neutron

IIose equivalent

Notes:

a Isotopic analysis for Uranium.

b Site Stormwater Detention Basin and Cylinder Storage Pads Stormwater Retention Basins.

9.3.3 ISA Summary

In Chapter 3 of this report, the staff provides its evaluation of the ISA Summary (AES, 2009c)

and documents its conclusion that the ISA Summary is complete, provides reasonable

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estimates of the likelihood and consequences of each accident sequence, and contains

sufficient information to determine whether adequate engineering or administrative controls are

identified for each accident sequence. Chapter 11 of this report contains the staffs 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. 10 CFR 70.61(c)(3) identifies the environmental

performance requirement that the applicant apply controls such that a credible intermediate

consequence 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

concentrations 5,000 times the values in Table 2 of Appendix B to 10 CFR Part 20. The

restricted areas are defined by the applicant in Section 4.7.1.2 of the SAR (AES, 2009a). The

restricted areas include the storage areas for UF 6 in the Cylinder Receipt and Shipping Building

and the potentially contaminated areas in the TSB. Restricted areas will also include "radiation

areas," "airborne radioactivity areas," "high-radiation areas," "contaminated areas," and areas

where inhalation of 1 milligram (mg) soluble uranium in a week is possible without respiratory

protection. These areas will be posted accordingly.

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., U1F6),

the applicable value of Table 2 of Appendix B to 10 CFR Part 20 is 3 x 10-12 PiCi/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 ýtCi/mL total radioactivity. AES assumed a bounding enrichment of

6 percent. Using footnote 3 of Part 20, Appendix B, the specific activity is approximately 2.8 jiCi

per gm 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 5.47 mg U/m3, was

evaluated at the restricted area boundary. In Sections 3.7 and 3.8 of the ISA Summary

(AES, 2009c), the applicant showed 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 ensure 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 ensured that the

environmental performance requirements of 10 CFR 70.61(c)(3) will be met. The staff

determined that environmental consequences may occur only if uncontrolled, intermediate or

high consequences to workers are also present. The staff did not identify any accident

sequence that would fail to meet the environmental performance requirements of

10 CFR 70.61(c)(3).

The applicant's approach to risk reduction will 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 this report, which addresses accident sequences

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for high and intermediate consequences. It also addresses preventive and mitigative measures.

The staff finds that the applicant's ISA adequately addresses radiological and non-radiological

risks to the environment consistent with 10 CFR 70.61, is consistent with guidance in

Section 9.4.3.2.3 of NUREG-1 520 (NRC, 2002), and the ISA is, therefore, acceptable to the

staff.

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

and ER, is adequate to protect the environment and the health and safety of the public and

complies with regulatory requirements in 10 CFR Parts 20, 30, 40, 51, and 70.

The NRC staff expects to issue a Final EIS, as required by 10 CFR 51.20, in February 2011 for

this license application. The Final EIS will consider the environmental impacts of the proposed

construction, operation, and decommissioning of the proposed facility and compare alternatives,

to inform the NRC staff recommendation concerning the proposed license application.

9.5 References

(AES, 2009a) AREVA Enrichment Services LLC, "Eagle Rock Enrichment Facility Safety

Analysis Report," Revision 1, 2009.

(AES, 2009b) AREVA Enrichment Services LLC, "Eagle Rock Enrichment Facility

Environmental Report," Revision 1, 2009.

(AES, 2009c) AREVA Enrichment Services LLC, "Eagle Rock Enrichment Facility ISA

Summary," Revision 1, 2009.

(NRC, 2010) U.S. Nuclear Regulatory Commission, NUREG-1 945, "Draft Environmental Impact

Statement for the Proposed Eagle Rock Enrichment Facility in Bonneville County, Idaho," 2010.

(NRC, 1979) U.S. Nuclear Regulatory Commission, Regulatory Guide 4.15, "Quality Assurance

for Radiological Monitoring Programs (Normal Operations) - Effluent Streams and the

Environment," 1979.

(NRC, 1985) U.S. Nuclear Regulatory Commission, 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, NUREG-1 302, "Offsite Dose Calculation

Manual Guidance: Standard Radiological Effluent Controls for Boiling Water Reactors," 1991.

(NRC, 1992) U.S. Nuclear Regulatory Commission, Regulatory Guide 8.34, "Monitoring Criteria

and Methods to Calculate Occupational Radiation Doses," 1992.

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(NRC, 1993) U.S. Nuclear Regulatory Commission, Regulatory Guide 8.37, ALARA Levels for

Effluents From Materials Facilities," 1993.

(NRC, 1994) U.S. Nuclear Regulatory Commission, 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, NUREG-1 520, "Standard Review Plan for

the Review of a License Application for a Fuel Cycle Facility," 2002.

(PNNL, 2007) Pacific Northwest National Laboratory, GENII 2.06, 2007.

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CHAPTER 10.0 DECOMMISSIONING

The purpose of the U.S. Nuclear Regulatory Commission's (NRC's) review of AREVA

Enrichment Services's (AES's) decommissioning plans is to evaluate whether the applicant will

be able to decommission the facility safely and in accordance with NRC requirements.

At the time of the initial License Application for a uranium enrichment facility, an 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 (FA) to cover the cost of those activities in the

future. The DFP, therefore, should contain an overview of the applicant's proposed

decommissioning activities, the methods used to determine the cost estimate, and the FA

mechanism. This overview should contain sufficient detail to enable the reviewer to determine

whether the decommissioning cost estimate is reasonably accurate.

10.1 Regulatory Requirements

The following NRC regulations require planning, FA, and recordkeeping for decommissioning,

as well as procedures and activities to minimize waste and contamination:

* Title 10 of the Code of Federal Regulations (10 CFR) 20.1401-1406 "Radiological

Criteria for License Termination"

* 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 "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) "Contents of Applications" (DFP)

* 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 Guidance and Acceptance Criteria

The guidance applicable to NRC's review of the decommissioning section of the Safety Analysis

Report (SAR) is contained in Chapter 10 of "Standard Review Plan for the Review of a License

Application for a Fuel Cycle Facility," NUREG-1 520 (NRC, 2002a) and in Volume 3 of

"Consolidated NMSS Decommissioning Guidance," NUREG-1 757 (NRC, 2003). Chapter 10 of

NUREG-1520 (NRC, 2002a) is applicable to the Eagle Rock Enrichment Facility (EREF), except

that the review used NUREG-1 757 (NRC, 2003), which is the updated version of NUREG-1727,

"NMSS Decommissioning Standard Review Plan" (NRC, 2000), which is referenced in

NUREG-1520 (NRC, 2002a).

NUREG-1757 (NRC, 2003) provides guidance for developing final decommissioning plans

required under 10 CFR 30.36(g), 10 CFR 40.42(g), and 10 CFR 70.38(g). A final

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decommissioning plan will be provided at the time of decommissioning. At the time of initial

licensing and for license renewals, Section 10.1 of NUREG-1 520 (NRC, 2002a) describes an

overview of the proposed decommissioning activities needed to develop the DFP. This

overview is a more generalized discussion of the detailed information that would be needed for

the final decommissioning plan described in NUREG-1 757 (NRC, 2003).

With regard to the acceptance criteria, because depleted uranium (DU) deconversion services

are not currently available in the U.S., DU generated in the operation of the EREF is considered

as a potential decommissioning obligation in the DFP. Thus, in addition to providing FA for the

decommissioning of the facility and site, AES intends to provide financial assurance for the

expected costs associated with the disposition of depleted uranium tails.

The requirements in 10 CFR 40.36(d) and 10 CFR 70.25(e) specify that the DFP must contain a

certification by the licensee that FA for decommissioning has been provided in the amount of

the cost estimate for decommissioning. AES has requested an exemption, under 10 CFR 40.14

and 70.17, to the decommissioning funding requirements in order to incrementally fund its FA

obligation as depleted uranium is generated and as each Separations Building Module (SBM)

becomes operational. Section 1.2.4.2 of this SER describes the exemption request and the

NRC staff's evaluation, which concluded that the exemption could be granted.

10.3 Staff Review and Analysis

The NRC's staff review of the planned decommissioning activities as described in Chapter 10 of

the SAR (AES, 2010) focused on the applicant's conceptual decommissioning activities for the

EREF, the decommissioning cost estimates, and the FA for decommissioning activities. The

applicant identified the decommissioning activities that may be needed in the future for

decommissioning and presented site-specific estimates of decommissioning costs for those

activities. The applicant intends to decommission the site such that any residual radioactivity

that is distinguishable from background radiation results in a total effective dose equivalent to an

average member of the critical group that does not exceed 25 mrem (0.25 mSv) per year.

Using the cost data as a basis, the applicant stated that it has presented FA to cover the costs

required to release the EREF for unrestricted use.

The following subsections contain the NRC staffs assessment of the applicant's proposed

decommissioning schedule and activities, cost estimate, and funding plan. Before license

termination, the applicant will provide a detailed decommissioning plan that will include specific

activities which will be used to protect workers, the public, and the environment.

10.3.1 Request for Alternate Schedule for Decommissioning

The requirements in 10 CFR 70.38(h) state that the decommissioning of the site or a separate

building or outdoor area must be completed within 24 months of initiating decommissioning

activities. However, 10 CFR 70.38(i) states the Commission may approve requests for alternate

decommissioning schedules. In its September 28, 2009, response to NRC's August 26, 2009,

request for additional information (RAI) (NRC, 2009b), the applicant requested approval of an

alternate schedule for decommissioning and provided justification for the longer schedule,

stating that decommissioning of each SBM would take approximately 4.5 years and completion

of site decommissioning would take approximately 8 years (AES, 2009b).

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The NRC staff reviewed the applicant's request for an alternate schedule for decommissioning

in light of the five considerations in 10 CFR 70.38(i). These are: (1) whether it is technically

feasible to complete decommissioning within the allotted 24-month period; (2) whether sufficient

waste disposal capacity is available to allow completion of decommissioning within the allotted

24-month period; (3) whether a significant volume reduction in wastes requiring disposal will be

achieved by allowing short-lived radionuclides to decay; (4) whether a significant reduction in

radiation exposure to workers can be achieved by allowing short-lived radionuclides to decay;

and (5) other site-specific factors which the Commission may consider appropriate on a caseby-case basis. The applicant addressed all five considerations in the RAI response (AES,

2009b). In regards to technical feasibility, the applicant asserted in the RAI response dated

September 28, 2009, that the decommissioning of a single SBM is assumed to take 4.5 years, 3

years for decommissioning of the centrifuges and associated equipment and 1.5 years for

decontamination of the structure (AES, 2009b). Dismantling and decontamination in the four

SBMs will be performed in a phased approach such that the decommissioning of all four SBMs

is completed within an 8-year timeframe. The primary reason for the 3-year timeframe for

decommissioning the centrifuges in each SBM is the quantity of the centrifuges and associated

equipment that must be decommissioned and the fact that this equipment is classified. An

additional consideration is the size of the SBM itself.

Based on the applicant's justification that it is not technically feasible to complete

decommissioning within a 24-month period, the staff determined that the alternate schedule is

warranted and, therefore, the applicant's request is approved.

10.3.2 Conceptual Decontamination and Decommissioning Activities

Section 10.1 of NUREG-1520 (NRC, 2002a) states that the DFP is required to contain an

overview of the proposed decommissioning activities. This section of the Safety Evaluation

Report (SER) describes the staff's review of the overview of the proposed decommissioning

activities. A detailed decommissioning plan will be provided at the time of decommissioning in

accordance with 10 CFR 30.36(g), 10 CFR 40.42(g), and 10 CFR 70.38(g).

Section 10.1.4, "Decommissioning Strategy" of the SAR (AES, 2009a) notes that the plan for

decommissioning the EREF is to promptly decontaminate or remove all materials from the site

which prevent release of the facility and site for unrestricted use. At the end of useful plant life,

the EREF will be decommissioned such that the site and remaining facilities may be released

for unrestricted use as defined in 10 CFR 20.1402 (AES, 2009a). The applicant performed a

site characterization survey that included collection and analysis of soil and groundwater

samples to determine background levels of uranium and thorium, both of which exist in nature

and will be used at the facility. The initial site characterization data is provided in sections 3.3

and 3.11 of AES's Environmental Report (AES, 2009b). In the RAI (NRC, 2009b), the staff

indicated that the number of samples in the initial site characterization were not sufficient to use

for demonstration of compliance with the 10 CFR Part 20 Subpart E decommissioning criteria.

In an RAI response (AES, 2009b), the applicant provided the following commitment to perform

sampling based on the guidance provided by Section A.3.4 of Appendix A to NUREG-1757

(NRC, 2000):

Prior to the commencement of construction, AES will collect additional surface soil

samples and analyze them for radiological constituents. The site property will be divided

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into four survey units, and 15 surface soil samples will be taken per survey unit (i.e., 60

additional soil samples). The sample collections will be taken from areas that include

(1) the detention and retention basins, (2) Full Tails, Full Feed, and Empty Cylinder

Storage Pads north of the main facilities, (3) the Technical Services Building (TSB),

Blending, Sampling and Preparation Building, SBMs, Uranium Hexafluoride (UF6)

Handling Areas, and Full Product Cylinder Storage Pad, and (4) areas on-site, but

outside those that are scheduled to be disturbed during plant construction. During

construction of the main plant facilities, additional soil samples from disturbed areas next

to facility foundations will be taken to characterize foundation soils prior to UF6 cylinders

arriving on-site.

NRC staff reviewed the site characterization data and commitments to collect and analyze

additional samples and determined that the data and the commitment to conduct additional

sampling and analysis will be adequate to determine a background value for the site, which can

be used at decommissioning time to determine whether the site meets the unrestricted release

criteria.

To formalize AES's commitment, the staff will impose the following license condition:

Prior to the commencement of construction, AES shall collect additional surface soil

samples and analyze them for radiological constituents. The site property will be divided

into four survey units, and 15 surface soil samples shall be taken per survey unit (i.e., 60

additional soil samples). The sample collections shall be taken from areas that include

(1) the detention and retention basins, (2) Full Tails, Full Feed, and Empty Cylinder

Storage Pads north of the main facilities, (3) the Technical Services Building, Blending,

Sampling and Preparation Building, SBMs, UF6 Handling Areas, and Full Product

Cylinder Storage Pad, and (4) areas on-site, but outside those that are scheduled to be

disturbed during plant construction.

Section 10.1.5, "Decommissioning Design Features," of the SAR states that decommissioning

planning begins with ensuring design features are incorporated into the plant's initial design to

simplify eventual dismantling and decontamination. The plans are implemented through proper

management and health and safety programs. Decommissioning policies address radioactive

waste management, physical security, and material control and accounting. Major features

incorporated into the facility design that facilitate decontamination and decommissioning include

radioactive contamination control, worker exposure and waste volume control, management

organization, health and safety, waste management, security and material control, and

recordkeeping.

The NRC staff reviewed the applicant's decommissioning design features and design feature

implementation plans and determined that they provide reasonable assurance that the plant will

be designed to facilitate eventual decommissioning in accordance with NRC regulations and are

therefore adequate.

Section 10.1.6, "Decommissioning Process," of the SAR describes the decommissioning

methodology to be employed at the EREF. The four SBMs will be shutdown in sequence

starting with SBM 1. Termination of SBM 4 operations will mark the end of uranium enrichment

operations at the facility. Decommissioning of the remaining plant systems and buildings will

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begin after SBM 4 operations have been permanently terminated. A conceptual

decommissioning schedule is provided in Figure 10.1-1 of the SAR, "Eagle Rock Enrichment

Facility - Conceptual Decommissioning Schedule."

The decommissioning process will include performing a radiological survey of the facility

immediately prior to the start of decommissioning and preparation of a detailed

Decommissioning Plan in accordance with 10 CFR 70.38 and the applicable guidance provided

in NUREG-1757. 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 will not be 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.

Decommissioning, using the decontamination and dismantlement (DECON) approach, requires

residual radioactivity to be reduced below specified levels so the facilities may be released for

unrestricted use. The unrestricted release criteria in 10 CFR Part 20.1402 serve as the basis

for decontamination costs estimated herein. The intent of decommissioning the facility is to

remove all enrichment-related equipment from the buildings such that only the building shells

and site infrastructure remain.

The standard decontamination methodology to be used during EREF decommissioning will

employ conventional decontamination techniques as described in Section 10.1.6 of the SAR.

Decontaminated components may be reused or sold as scrap. All equipment that is to be

reused or sold as scrap will be decontaminated to a level at which further use is unrestricted.

Materials that cannot be decontaminated will be disposed of in a licensed radioactive waste

disposal facility. Contaminated portions of the buildings will be decontaminated as required.

When decontamination is complete, all areas and facilities on the site will be surveyed to verify

that further decontamination is not required. Decontamination activities will continue until the

entire site is demonstrated to be suitable for unrestricted use.

Section 10.1.6.9 of the SAR describes the final radiation survey that will be used to verify proper

decontamination to allow the site to be released for unrestricted use. The evaluation of the final

radiation survey will be based in part on an initial radiation survey performed prior to initial

operation, discussed above, which provides a datum for measurements which determine any

increase in levels of radioactivity. The survey procedures and results will be analyzed and

shown to be below allowable residual radioactivity limits; otherwise, further decontamination will

be performed.

Section 10.1.7, "Decontamination Facilities," of the SAR describes a specialized facility that will

be required to accommodate decommissioning and describes a general procedure for

decontaminating contaminated plant components, including centrifuges. This 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 required for

handling the remainder of the various plant components. These facilities are assumed to be

installed in existing plant buildings [such as the Centrifuge Assembly Building (CAB)]. The

decontamination facility will have four functional areas that include: (1) a disassembly area,

(2) a buffer stock area, (3) a decontamination area, and (4) a scrap storage area for cleaned

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stock. Contaminated plant components will be cut up or dismantled, then processed through

the decontamination facilities. Contamination of site structures will be limited to areas in the

SBMs, the Blending, Sampling and Preparation Building, the CAB, and TSB, and will be

maintained at low levels throughout plant operation by regular cleaning. Although

decommissioning operations are planned to be underway while all the activities considered in

the integrated safety analysis (ISA) continue to occur in the other portions of the plant, the

current ISA has not considered these decommissioning risks. An updated ISA will be performed

at a later date, but prior to decommissioning, to incorporate the risks from decommissioning

operations on concurrent enrichment operations.

The NRC staff has reviewed the applicant's conceptual decommissioning process and

determined that the applicant has considered site-specific decommissioning activities that may

be needed in the future, and, therefore, that the applicant's proposed process provides

reasonable assurance that decommissioning can be performed in accordance with NRC

regulations.

10.3.3 Financial Assurance for Decommissioning

Applicants for licenses under 10 CFR Part 30, 40 and 70 are subject to FA requirements for

decommissioning, decontamination and reclamation pursuant to 10 CFR 30.35, 40.36 and

70.25. As part of the application, AES submitted various documents to demonstrate that

adequate FA will be provided to decontaminate and decommission the facility and site to

unrestricted release criteria. AES intends to provide financial assurance for site and facility

decommissioning as well as for the disposition of DU. AES is responsible for the disposal of

DU. Section 3113(a) of the USEC Privatization Act, 42 USC § 2297h-1 1(a), requires the

Department of Energy (DOE), upon request of the operator of a uranium enrichment facility

licensed by the NRC, to accept DU for disposal for a fee if it is determined that DU is a low level

radioactive waste. In 2005, the Commission made that determination. Louisiana Energy

Services, L.P. (National Enrichment Facility), CLI-05-5, 61 NRC 22, 36 (2005). Therefore, the

applicant has assumed that DOE will take title to and possession of DU for disposal. The

applicant intends to provide FA for both site and facility decommissioning and the disposition of

DU incrementally over time, and thereby accrete the total amount FA provided proportionally to

the then current decommissioning cost estimate.

As described in section 10.2.1 of the SAR, the applicant intends to rely on a Letter of Credit

(LOC) to provide such FA as set forth in 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 LOC and standby trust agreement and

supporting documentation. The applicant will finalize the financial instruments to be relied on

and, prior to receiving any licensed material at the EREF, will provide executed originals of the

reviewed financial instruments to the NRC for final review and confirmation. As required by

10 CFR 70.38(e), AES will provide adequate FA continuously until license termination.

Unless otherwise noted, all updates to the DFP and cost estimate will be based on the costs of

a third party contractor, will not take credit for any salvage value that might be realized from the

sale of potential assets during or after decommissioning, shall be updated to current year

dollars -- taking into account applicable changes in costs such as, but not limited to, changes in

labor, disposal, and shipping costs; inflation; currency exchange rates; and other site and facility

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factors -- and shall contain a 25 percent contingency factor. These requirements will be

included in a license condition that will be imposed to ensure compliance with AES's

incremental funding mechanism, discussed in Section 10.3.3.1.1.

10.3.3.1 Description of Incremental Funding Approach and Approach to Updating the DFP,

Cost Estimate, and Financial Instruments

In Sections 1.2.5, 10.2.1 and 10.2.2 of the SAR (AES, 2009a), AES describes its proposed

approach to providing FA incrementally over time by updating the decommissioning cost

estimate and FA instruments. Specifically, in Section 1.2.5 of the SAR (AES, 2009a), the

applicant requested an exemption to 10 CFR 40.36(d) and 10 CFR 70.25(e), which require a

certification of FA to be provided in the amount of the cost estimate. As described in Section

10.2.1 of the SAR (AES, 2009a), the applicant intends to build, install and operate the SBMs

sequentially over time and thus requests to provide FA on an incremental and forward looking

basis, proportional to the then current decontamination and decommissioning liability.

These updates will be made annually on a forward-looking, incremental basis until the facility is

at full capacity operation, which would occur in approximately March 2022, according to the

AES's schedule. Until such time, AES will provide annual updates to the DFP, cost estimate

and FA mechanism for site and facility decommissioning, and the DU disposition cost estimate

and its associated FA mechanism. As described in AES's supplemental response to RAI SE-1,

after the facility has reached full capacity operation, the cost estimate and FA mechanisms for

site and facility decommissioning will be updated at least triennially, and the cost estimate and

FA instrument for DU disposal will continue to be updated annually (AES, 201 Oa).

10.3.3.1.1 Annual, Forward-Looking Updates until Full Capacity Operation of EREF

As stated in AES's response to RAI SE-1 (AES, 2009b) and as further clarified in AES's

supplemental response to RAI SE-1 (AES, 2010a), the first facility decommissioning cost

estimate will be provided six months prior to the delivery of natural UF6, [less than 20 kilograms

(kg) Uranium (U)] as testing material for the CAB. The facility decommissioning cost estimate

will then be updated six months in advance of the first delivery of natural UF6 (more than 50

kgU) as feed material for the first SBM. Thereafter, it will be updated annually (i.e., six months

in advance of each anniversary of the first delivery of natural UF6 (more than 50 kgU) as feed

material for the first SBM).

According to AES's milestones described in Chapter 10 of the SAR (AES, 2009a) and in the

response to RAI SE-1 (AES, 2009b), AES intends to install four SBMs and introduce initial feed

material in four separate Phases.

* Phase One - The receipt of Natural UF6 as feed material for the first SBM

* Phase Two - The receipt of Natural UF6 as feed material for the second SBM

* Phase Three - The receipt of Natural UF6 as feed material for the third SBM

* Phase Four - The receipt of Natural UF6 as feed material for the fourth SBM

AES expects Phase One to commence shortly after completing testing at the CAB. Thereafter,

AES expects two years to elapse between subsequent Phases. Until the facility is at full

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capacity operations, as stated in its response to RAI SE-1 (AES, 2009a) and supplemental

response to RAI SE-1 (AES, 2010b), AES commits to providing NRC with an updated, forward

looking cost estimate for both site and facility decommissioning, and DU disposition on an

annual basis and in accordance with the phasing of the first delivery of natural UF6 as feed

material for each SBM, at least six months prior to the described above milestones. The NRC

will review each cost estimate. After NRC review and approval, at least 21 days prior to any of

the above phases, AES will provide an executed version of the reviewed financial mechanism(s)

whose aggregate amount will be equal to at least the approved, updated cost estimate, in

accordance with 10 CFR 70.25(e). The specific scope of each update is described later in this

section.

At each of the above phases, AES will not receive any additional material as initial feed material

for each subsequent SBM not previously in operation until the decommissioning cost estimate is

approved and full funding for decommissioning the site and facility and disposition of DU, is

provided to and confirmed by NRC. Providing an executed FA mechanism will not permit AES to

bring additional material prior to the approval of the DFP and cost estimate. This requirement

will be included in a license condition that will be imposed to ensure compliance with AES's

proposed incremental funding.

Initial Financial Assurance

AES's initial FA is intended to cover the testing phase of facility. During the testing phase, the

facility is not in operations, with respect to enriching uranium. At least six months prior to the

receipt of natural UF6 (less than 20 kgU) as test material, AES will provide an updated

decommissioning cost estimate and final copies of the proposed financial instrument(s) for NRC

review. The updated cost estimate and associated FA instrument(s) shall be sufficient to cover

the decommissioning costs of the: (1) CAB; and (2) any part of the facility not fully

decontaminated as approved by NRC (i.e., NRC has determined can be released for

unrestricted use). According to Section 1.2.3 of the SAR (AES, 2009a), AES expects to utilize

"...[o]ther source materials and by-product materials.. .for instrument calibration purposes....,"

and "[t]hese materials will be identified during the design phase, and AES will submit a request

to amend the Materials License to incorporate the proposed quantities and types for the sealed

and unsealed instrument calibration sources to its possession limits." As stated by AES in its

responses to NRC's RAI SE-1 and GI-1 (AES, 2009b), the above referenced update to the

decommissioning cost estimate and funding instrument will also provide adequate funding for

the decommissioning activities related to these additional materials. After NRC review, and at

least 21 days prior to the receipt of test material, AES will provide an executed version of the

reviewed financial mechanism(s) whose aggregate amount will be equal to at least the

approved, updated decommissioning cost estimate.

Financial Assurance Updates

Following the initial FA submittal and until the facility is at full capacity, AES will provide an

updated DFP, cost estimate and financial assurance instrument(s) at least six months in

advance of the planned date for receipt of natural UF 6 (greater than 50 kgU) as initial feed

material for SBMs One through Four. In addition, after SBM One begins operation, AES will

provide annual updates to the DFP, cost estimate, and FA instruments. Each update will be

forward-looking and will cover: (1) any part of the facility that would currently be in operation;

(2) any part of the facility that has been in operation, or any part of the facility reasonably

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believed to be contaminated and not fully decontaminated and decommissioned as approved by

NRC; (3) all plant areas where licensed material is stored or used; (4) any part of the facility that

would be in operation within the next 12 months; and (5) an update to the DU disposal cost

estimate and financial assurance mechanism (described in the next paragraph). After NRC

review and approval, at least 21 days prior to the receipt of initial feed material to each SBM,

AES will provide an executed version of the reviewed financial mechanism(s) -- whose

aggregate amount will be equal to at least the approved, updated decommissioning cost

estimate - for NRC review and confirmation. According to AES's supplemental response to RAI

SE-1 (AES, 201 Oa), the applicant will also provide an executed version of the reviewed financial

mechanism(s) with each annual update to the DFP and cost estimate.

The DU disposal cost estimate and financial mechanism, submitted in connection with the

update that encompasses the receipt of initial feed material for SBM One, will cover the

disposition of the accumulated DU expected onsite stemming from the first three years of

operations. According to AES's supplemental response to RAI SE-1 (AES, 201 Oa), annually for

the next two years, AES will provide an update to its DU disposition cost estimate and financial

mechanism in conjunction with the annual updates described in the above paragraph, sufficient

to cover the disposition of the accumulated DU expected onsite stemming from the first three

years of operations. According to AES's supplemental response to RAI SE-1 (AES, 201 Oa),

after the first two years of operations, AES will provide annual updates to the DU disposition

cost estimate, sufficient to cover the accumulated DU onsite and a projection of the amount of

DU that would be onsite within the subsequent 12 months. All updates for DU disposal will be

based on an updated DOE cost estimate and the total amount provided for DU disposition will

not be less than the updated DOE cost estimate. In its supplemental response to RAI SE-1

(AES, 2010a), AES states the accumulated DU onsite quantity will not include any quantity of

DU that would have been already delivered offsite for disposition. NRC finds this acceptable,

provided that DOE has taken title to and possession of such DU and assumed responsibility for

its decommissioning and disposal.

This method of providing incremental and annual updates automatically captures changes in

construction schedules. Since each update is forward looking for the next 12 months, if SBM

construction schedules are accelerated, the timing and/or scope of the updates would

automatically address such changes. Since AES committed to providing annual updates to the

cost estimate and financial mechanism(s), should construction schedules be delayed, the cost

estimate for site and facility decommissioning and DU disposal will continue to be updated

annually.

To ensure compliance with AES's proposed methodology of providing FA on an incremental,

forward-looking basis, the staff will impose the following license condition:

The licensee shall provide FA on the following schedule:

a. The licensee shall provide an updated DFP, updated facility decommissioning

cost estimate, and final copies of proposed financial assurance instruments to

the NRC for review at least six months prior to the following dates:

(1) planned date for obtaining test material (5 20 kg U) for the CAB

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(2) planned date for obtaining feed material (> 50 kgU) for initial production in

the first SBM

(3) planned date for obtaining feed material (> 50 kgU) for initial production in

the second SBM

(4) planned date for obtaining feed material (> 50 kgU) for initial production in

the third SBM

(5) planned date for obtaining feed material (> 50 kgU) for initial production in

the fourth SBM

The updates shall be forward-looking through the 12-month period beginning on

the applicable date listed above. For each update, the licensee shall provide

final executed copies of the NRC-reviewed financial assurance instruments to

NRC at least 21 days prior to receipt of test material or receipt of feed material

for initial production in an SBM.

b. After the first SBM begins operations, and until the plant reaches full capacity,

the licensee shall, on an annual basis, provide an updated DFP, an updated

facility decommissioning cost estimate, and final copies of proposed financial

assurance instruments to NRC for review. These annual updates shall be

provided six months prior to the anniversary date of obtaining feed material for

initial production in the first SBM, and shall be forward-looking through the 12-

month period beginning on the anniversary date. For each annual update, the

licensee shall provide final executed copies of the NRC-reviewed financial

assurance instruments to NRC at least 21 days prior to the anniversary date.

If the licensee provides an annual update at least six months prior to the planned

date for obtaining feed material for initial production in the second, third, or fourth

SBM, that annual update may also serve as the update required in paragraph (a)

for that date.

c. The updated DFPs, updated cost estimates, and financial assurance instruments

described in paragraphs (a) and (b) shall include full funding for decontamination

and decommissioning of." (1) any part of the facility currently in operation; (2) any

part of the facility that has been in operation, or any other part of the site or

facility reasonably believed to be contaminated, that has not been fully

decontaminated and decommissioned as approved by NRC (including the CAB);

(3) all plant areas where licensed material is stored or used; and (4) any part of

the facility (including SBMs) expected to be in operation by the end of the

applicable forward-looking 12-month period in paragraph (a) or (b).

d. The licensee shall provide an initial depleted uranium (DU) disposition cost

estimate and final copies of proposed financial assurance instruments for DU

disposition in conjunction with the updated DFP, updated facility

decommissioning cost estimate, and financial assurance instruments that will be

submitted at least six months prior to obtaining feed material for initial production

in the first SBM. The DU disposition cost estimate and proposed financial

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assurance instruments shall include full funding to cover disposition of the first

three years of DU tails generation. The DU disposition cost estimate shall

include an update to the U.S. Department of Energy (DOE) DU disposition cost

estimate. The total amount funded for DU disposition shall not be less than the

updated DOE cost estimate.

For the initial DU disposition cost estimate, the licensee shall provide final

executed copies of the NRC-reviewed financial assurance instruments for DU

disposition to NRC at least 21 days prior to the receipt of feed material for the

first SBM.

e. The licensee shall provide updates to the DU disposition cost estimate and

financial assurance instruments for DU disposition as described below:

(1) During the first two years of operation, the licensee shall provide updated

DU disposition cost estimates and final copies of proposed financial

assurance instruments for DU disposition in conjunction with the updates

required in paragraphs (a) and (b). The updated cost estimates shall

provide full funding to cover disposition of the first three years of DU tails

generation.

(2) After the first two years of operation and until the facility reaches full

capacity, the licensee shall provide updated DU disposition cost

estimates and final copies of proposed financial assurance instruments

for DU disposition in conjunction with the updates required in paragraphs

(a) and (b). The updated DU disposition cost estimates shall provide full

funding to cover disposition of all DU stored onsite and all DU expected to

be generated by the end of the applicable forward-looking 12-month

period in paragraph (a) or (b).

(3) After the plant reaches full capacity, the licensee shall continue to provide

annual updates to the DU disposition cost estimate, along with revised

financial assurance instruments. These annual updates shall include full

funding to cover disposition of all DU stored onsite and all DU expected to

be generated by the end of the 12-month period beginning on the

anniversary date of obtaining feed material for initial production in the first

SBM. The annual updates to the DU disposition cost estimate and final

copies of proposed financial assurance instruments shall be provided to

NRC for review six months prior to the anniversary date.

The licensee may exclude from the updated DU disposition cost estimates any

DU that the DOE has taken title to and possession of pursuant to Section 3113 of

the USEC Privatization Act. All updates to the DU disposition cost estimates

shall include an update to the DOE cost estimate for DU disposition. The total

amount funded for DU disposition shall not be less than the updated DOE cost

estimate.

For DU disposition cost estimate updates, the licensee shall provide final

executed copies of the NRC-reviewed financial assurance instruments for DU

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disposition to NRC at least 21 days prior to the receipt of feed material for an

SBM, or the anniversary date of obtaining feed material for initial production in

the first SBM, as applicable.

fr If the construction and/or operation of any SBM is delayed or cancelled, the

licensee is not relieved of its commitment to provide updated DFP, facility

decommissioning cost estimates, DU disposition cost estimates, and final copies

of proposed financial assurance instruments to NRC as described in paragraphs

(a)-(e).

g. When an update to the DFP, cost estimates for facility decommissioning and DU

disposition, and financial assurance instruments encompasses the first delivery

of natural uranium hexafluoride (> 50 kgU) as feed material to an SBM not

previously in operation, the licensee shall not receive such initial feed material

until the NRC reviews the updated DFP and cost estimates and confirms the

executed financial assurance instrument(s).

h. All updates to the DFP, cost estimates for facility decommissioning and DU

disposition, and financial assurance instruments, shall be updated to current year

United States dollars and shall encompass all current cost data, taking into

account changes in inflation, foreign currency exchange rates, possession limits,

licensed material, labor rates, disposal and shipping rates, and site and facility

factors. All costs shall be based on the costs of a third party contractor and shall

not take credit for any salvage value that might be realized from the sale of

potential assets during or after decommissioning. All costs (including those for

DU disposition) shall include a contingency factor of at least 25 percent.

10.3.3.1.2 Updates for Full Capacity Operation of EREF

After the facility has reached full production output, updates to the cost estimate for DU disposal

follow a different schedule than updates to the cost estimate for site and facility

decommissioning. Subsequent updates to the facility decommissioning cost estimate and

revised financial instruments for facility decommissioning will be provided at least every three

years pursuant to 10 CFR 30.35(e), 40.36(d), and 70.25(e). The updated facility

decommissioning cost estimate updates will be accompanied with final copies of proposed

financial instruments whose aggregate amount is sufficient to cover the updated

decommissioning cost estimate and cost estimate for DU disposal.

Updates to the cost estimate and revised financial instruments for the disposition of DU will

continue to be provided on an annual, forward-looking basis. This annual update to the DU

disposition cost estimate will be based on an updated DOE cost estimate and will be sufficient

to cover the disposition of accumulated DU onsite and the projected amount of DU that would

be generated onsite within the subsequent 12 months. The updated cost estimate for the

disposition of DU will be accompanied with final copies of proposed financial instruments whose

aggregate amount is sufficient to cover the approved decommissioning cost estimate and

updated cost estimate for DU disposal.

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Under AES's requested approach to provide financial assurance on an incremental and forward

looking basis, the staff finds that AES will maintain adequate financial assurance at least equal

to its most recently approved cost estimate for decommissioning. The staff finds that the

described schedule and methodology for updating the DFP, cost estimate and FA mechanisms

is satisfactory, will provide updates to the cost estimate at least triennially consistent with

10 CFR 30.35(e), 40.36(d), and 70.25(e), and therefore the staff finds that AES's approach will

not endanger life or property.

10.3.3.2 Discussion on the Adequacy of the Cost Estimate

A detailed review was conducted of AES's decommissioning cost estimate. The

decommissioning cost estimate was developed by the centrifuge supplier, Enrichment

Technology Company Limited (ETC). In its supplemental response to RAI D-4 (AES, 2010a),

AES states that ETC has experience in Europe with regard to dismantling, decontaminating,

and disposing of parts of earlier generations of centrifuges. Additionally, AES stated that ETC

"undertook studies to look at how the modern centrifuges as using in the EREF would be

decommissioned and declassified at the end of their lifetime" (AES, 201 Oa). AES further states

that this "process flow.. .was fully tested [by ETC] using both intact and crashed centrifuges of

the same type as used in EREF" (AES, 201 Oa). AES stated that "a layout and process flow for

a declassification facility has been developed and costed.. .[which] included the required

manning levels and throughputs; these have been built into the decommissioning model"

(AES, 2010a). Furthermore, AES stated that ETC updates its "model...on a regular basis for

each plant and [is] updated to account for any changes in disposal costs or additional

experience gained" (AES, 2010a). Based on AES's description, the staff finds that ETC's

experience provides an adequate basis for the cost estimate.

AES's decommissioning cost estimate does not include provisions for restoration of

contaminated areas on the facility grounds. In its response to RAI D-8 (AES, 2009b) and

supplemental response to RAI D-8 (AES, 201 Ob), AES stated that based on European

experience, "enrichment facilities are expected to operate cleanly because of plant

contamination controls and environmental monitoring and should result in no ground

contamination." Additionally, if surface contamination levels exceed the levels described in

section 4.8.5.1 of the SAR, clean-up of the contamination is initiated within 24 hours

(AES, 2009a). Given the basis supplied by AES, the staff finds AES's basis acceptable.

However, the staff is including a provision in the license condition described in Section

10.3.3.1.1 requiring AES to provide FA for contamination that is not cleaned-up.

Relying on its application, the initial financial assurance provided by AES is estimated to be

$12,782,500. Since the aforementioned dollar amount may change due to changes in inflation,

foreign currency exchange rates, labor rates, disposal rates, possession limits, and other cost

factors, and because AES may decide to utilize other source and/or byproduct materials for

testing purposes, the decommissioning cost estimate and its associated funding instrument will

be updated as described in Section 10.3.3.1.1 of the SAR (AES, 2009a), Initial Financial

Assurance. At the time of CAB operation, no SBM will be in operation.

After providing initial FA, AES intends to commence Phase One of operations. According to

section 10.2.2, and Table 10.1-14 of the SAR (AES, 2009a), as well as the responses to RAI

SE-1 (AES, 2009b), the estimated cost to decommission all applicable site and facility areas

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covered by Phase One, excluding DU disposal, is $131,098,750, assuming a third party

contractor completes the decommissioning and including a 25 percent contingency factor. AES

estimates that the cost to decommission the entire site and facility (all four phases), excluding

DU disposal, is $445,427,500. Therefore, the amount of FA provided in connection with Phase

One is approximately 29% of the total cost of site and facility decommissioning. According to

sections 10.2.2 and 10.3 and Table 10.3-1 of the SAR (AES, 2009a), AES estimates that it will

generate 11,452 MT DU during the first three years of operations, and the estimated cost of DU

disposal is $7.66 per kg. AES states that an amount of $109.6 million, based on a DOE cost

estimate and including a 25 percent contingency, is sufficient to cover the cost of disposing the

DU generation stemming from the first three years of operations. Therefore, the Phase One's

total estimated cost of site and facility decommissioning, and DU disposition is $240,751,000.

As stated previously, AES will be required to appropriately update the cost estimate to current

year United States Dollars and include all changes to relevant cost factors.

The staff finds the cost estimate consistent with NUREG-1 757, Volume 3, based on costs of a

third party contractor, does not take credit for any salvage value that might be realized from the

sale of potential assets during or after decommissioning, and includes a 25 percent contingency

factor. Additional key assumptions of AES's cost estimate are that: (a) decommissioning

activities will be performed in accordance with current day regulatory requirements; (b)

decommissioning costs and DU disposition costs are presented in 2007 dollars; and (c) some

costs were provided in Euros (E), and a rate of 0.714E to $1 was used to convert 2007 E to 2007

$. As indicated previously, AES's estimated cost to decommission the entire facility, excluding

DU disposition, is $445,427,500 (2007 U.S. dollars). The CAB, and subsequently the first SBM,

are not expected to require FA until 2012, therefore these figure are not updated to current year

United States Dollars, since AES will be required to update them appropriately prior to receipt of

any licensed material.

According to Table 10.1-14 and Section 10.3 of the SAR (AES, 2009a), based on a DOE cost

estimate, AES estimates total DU disposition costs, excluding a 25% contingency, to be $2.461

billion. Thus, AES estimates the total site and facility decommissioning and DU disposition cost,

including a 25% contingency factor, to be $3,523,436,000. Relying on AES's proposed

methodology of providing FA on an incremental basis, this total estimated cost of DU disposal,

would not be incurred immediately, but over the life of the facility. Therefore, the cost to

decommission the facility would not be $3,523,436,000 immediately at the commencement of

operations.

Based on its review, the staff finds the cost estimate to decommission the site and facility, and

the estimated cost for DU disposition, satisfies the requirements of 10 CFR 30.35(e) 40.36(d),

70.25(e), is consistent with NUREG-1757, Volume 3 and therefore is acceptable.

10.3.3.3 Review of Financial Instruments

AES provided NRC with a draft LOC and Standby Trust Agreement for review. The purpose of

these financial instruments is to provide reasonable assurance that funds will be available when

needed, to decommission and decontaminate the licensed site and facility to unrestricted

release criteria, as set forth in 10 CFR 20.1402. The submitted draft instruments are not the

proposed final versions. As stated previously, at least six months prior to the receipt of licensed

material for the CAB, AES will provide an updated facility decommissioning cost estimate and

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final copies of the proposed financial instruments. At least 21 days prior to the receipt of

licensed material, AES will provide executed copies of the NRC-reviewed financial instruments

for final confirmation before the applicant receives licensed material at the site and facility. The

specific regulatory requirements with regard to the LOC are set forth in 10 CFR 30.35(f)(2)(i) to

(iii), 40.36(e)(2)(i) to (iii) and 70.25(f)(2)(i) to (iii), which state:

(i) The [LOC] must be open-ended or, if written for a specified term, such as five years,

must be renewed automatically unless 90 days or more prior to the renewal date, the

issurer notifies the [NRC], the beneficiary, and the licensee of its intention not to

renew. The [LOC] must also provide that the full face amount be paid to the

beneficiary automatically prior to the expiration without proof of forfeiture if the

licensee fails to provide a replacement acceptable to the [NRC] within 30 days after

receipt of notification of cancellation.

(ii) The LOC must be payable to a trust established for decommissioning costs. The

trustee and trust must be acceptable to the [NRC]. An acceptable trustee includes

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.

(iii) The [LOC] must remain in effect until the [NRC] has terminated the license.

The staff has conducted a review of AES's draft LOC. The proposed LOC language is

consistent with the guidance in NUREG-1757, Volume 3, Appendix A.10. However, at this time,

the language does not satisfy the NRC regulations in 10 CFR 30.35(f)(2)(i), 40.36(e)(2)(i) and

70.25(f)(2)(i), because, among other items, no financial institution is specified at this time and

the aggregate dollar amount of the instrument is not indicated. As stated above, AES will

finalize this instrument and financial instruments are not required at this time, consistent with

10 CFR 70.25(b)(2). Once AES finalizes the specific instrument and submits it for review with

an updated cost estimate, the staff will review it for compliance with regulatory requirements at

that time.

The staff has conducted a review of AES's draft Standby Trust Agreement. The specific

regulatory requirements for standby trust agreements are set forth in 10 CFR 30.35(f)(2)(ii),

40.36(e)(2)(ii) and 70.25(f)(2)(ii). The language of the Standby Trust Agreement is consistent

with guidance contained in NUREG-1757 Volume 3, Appendix A.17. However, at this time, the

language does not satisfy the regulations because, among other items, no trustee is specified.

As stated above, AES will finalize this instrument and financial instruments are not required at

this time, consistent with 10 CFR 70.25(b)(2). Once AES finalizes the specific instrument and

submits it for review with an updated cost estimate, the staff will review it for compliance with te

regulatory requirements at that time.

On the basis of the NRC staffs above review, the applicant has demonstrated that (1) the cost

estimate for both the decommissioning of the site and facility, and disposal of DU is adequate;

(2) the language of the draft FA instruments is acceptable; (3) and the proposed incremental

funding of decommissioning FA is adequate to maintain appropriate and reasonable levels of

FA for the operational phases of the project. Under the above incremental funding approach,

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AES will maintain an appropriate amount of decommissioning financial assurance proportional

to its decommissioning and disposal obligations at any point in time. Therefore, the proposed

approach will not endanger human health and safety or property.

10.4 Evaluation Findings

The NRC staff has evaluated the applicant's plans and FA for decommissioning in accordance

with NUREG-1757, "Consolidated NMSS Decommissioning Guidance," September 2003. On

the basis of this evaluation, the NRC staff finds the applicant's plans and FA for

decommissioning comply with the NRC's regulations set forth in 10 CFR 30.35, 40.36, and

70.25, and provide reasonable assurance of protection for workers, the public, and the

environment.

To ensure compliance with the site characterization commitments, the staff will impose the

following license condition:

Prior to the commencement of construction, AES shall collect additional surface soil

samples and analyze them for radiological constituents. The site property will be divided

into four survey units, and 15 surface soil samples shall be taken per survey unit (i.e., 60

additional soil samples). The sample collections shall be taken from areas that include

(1) the detention and retention basins, (2) Full Tails, Full Feed, and Empty Cylinder

Storage Pads north of the main facilities, (3) the Technical Services Building, Blending,

Sampling and Preparation Building, SBMs, UF6 Handling Areas, and Full Product

Cylinder Storage Pad, and (4) areas on-site, but outside those that are scheduled to be

disturbed during plant construction. During construction of the main plant facilities,

additional soil samples from disturbed areas next to facility foundations shall be taken to

characterize foundation soils prior to UF6 cylinders arriving on-site.

To ensure compliance with the proposed methodology of providing FA on an incremental,

forward-looking basis, the staff will also impose the license condition described in

Section 10.3.3.1.1 of this SER.

10.5 References

(NRC, 2002a) U.S. Nuclear Regulatory Commission, NUREG-1520, "Standard

Review Plan for the Review of a License Application for a Fuel Cycle Facility," March 2002.

(NRC, 2003) U.S. Nuclear Regulatory Commission, NUREG-1 757, "Consolidated

NMSS Decommissioning Guidance," September 2003.

(NRC, 2009) U.S. Nuclear Regulatory Commission, "Request for Additional Information," August

26, 2009.

(AES, 2009a) AREVA Enrichment Services LLC, Eagle Rock Enrichment Facility, "Revision 1

to License Application for the Eagle Rock Enrichment Facility"; April 23, 2009.

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(AES, 2009b) AREVA Enrichment Services LLC, Eagle Rock Enrichment Facility; Response to

Requests for Additional Information - AREVA Enrichment Services LLC License Application for

the Eagle Rock Enrichment Facility; September 28, 2009.

(AES, 2010a) AREVA Enrichment Services LLC, Enclosure to email from Jim Kay, AES, to

Breeda Reilly, NRC, "Decommissioning Funding Plan and Financial Assurance RAI SE-1, D-4,

and D-8," May 25, 2010.

(AES, 2010a) AREVA Enrichment Services LLC, Enclosure to email from Jim Kay, AES, to

Breeda Reilly, NRC, "Response to RAI D-8," June 18, 2010.

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CHAPTER 11.0 MANAGEMENT MEASURES

Management measures are functions performed by the licensee, generally on a continuing

basis, that are applied to items relied on for safety, to ensure the items are available and reliable

to perform their functions when needed. Management measures are required to assure

compliance with performance requirements of 10 CFR 70.61; 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 Title 10 of the Code of Federal

Regulations (CFR) Part 70 definition of management measures, including: (1) configuration

management (CM), (2) maintenance, (3) training and qualifications, (4) procedures, (5) audits

and assessments, (6) incident investigations, (7) records management, and (8) other quality

assurance (QA) elements.

The purpose of this review is to verify whether AREVA Enrichment Services, LLC (AES) Eagle

Rock Enrichment Facility's (EREF's) application provided conclusive information to ensure that

the management measures applied to IROFS, as documented in the ISA Summary, ensure that

the IROFS will be available and reliable to perform their functions when needed, to comply with

the performance requirements of 10 CFR 70.61. This review also determines whether the

measures are applied to the IROFS in a graded manner commensurate with the IROFS

importance to safety.

In implementing the QA Program Description (QAPD), AES will apply management measures

commensurate to the reduction of the risk attributable to the items. The applicant uses QA

Level designations -- QA Level 1, 2 and 3 -- to describe their graded approach. As described in

the license application, management measures will be applied to QA Level 1 and QA Level 2

items (AES, 2009a). All IROFS will be designated as QA Level 1 or QA Level 2 items. As

described in the QAPD, QA Level 1 items and activities include those items and activities whose

failure or malfunction could directly result in a condition that adversely affects public, worker and

the environment as described in 10 CFR 70.61. In addition, the failure of a single QA Level 1

item could result in a high or intermediate consequence. QA Level 2 items and activities are

described in the QAPD as those items and activities whose failure or malfunction could

indirectly result in a condition that adversely affects public, worker and the environment as

described in 10 CFR 70.61 (CFR, 2008a). The failure of a QA Level 2 item, in conjunction with

the failure of an additional item, could result in a high or intermediate consequence. All building

and structure IROFS associated with credible external events are QA Level 2. QA Level 2 items

and activities also include those attributes of items and activities that could interact with IROFS

due to a seismic event, and result in high or intermediate consequences as described in

10 CFR 70.61.

The QAPD was reviewed by the staff and accepted on April 8, 2010 (NRC, 2010). Based on

that review, documented in the accompanying Staff Evaluation Report, the staff found the

program acceptable for application to the design, construction, operation, including maintenance

and modification, and decommissioning of the proposed EREF (NRC, 2010). The Staff

Evaluation Report discusses the staff review of the QAPD which was based on NUREG-1520

(NRC, 2002) and documents the staffs conclusion that the QAPD adequately describes the

application of other QA elements and has adequately established other QA elements as part of

Management Measures as required by 10 CFR Part 70.62(d) (NRC, 2010).

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11.1 Requlatory 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: (1) CM, (2) maintenance,

(3) training and qualifications, (4) procedures, (5) audits and assessments, (6) incident

investigations, (7) records management, and (8) other Quality Assurance (QA) elements.

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 or licensee to establish management measures

to ensure that engineered or administrative controls and control systems that are

designated as IROFS are designed, implemented, and maintained, as necessary, to

ensure they are available and reliable to perform their function when needed, to comply

with the performance requirements of 10 CFR 70.61.

4. 10 CFR 19.12 states requirements for instructions to workers that are applicable to

personnel training and qualifications.

5. 10 CFR 70.64(a)(1) requires the design of new applications or new processes at existing

facilities to be developed and implemented in accordance with management measures,

to provide adequate assurance that items relied on for safety will be available and

reliable to perform their function when needed. Appropriate records of these items must

be maintained by or under the control of the licensee throughout the life of the facility.

6. 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.

7. 10 CFR 70.72 requires a licensee to establish a CM program, documented in written

procedures, to evaluate, implement, and track each change to the site, structures,

processes, systems, equipment, components, computer programs, and activities of

personnel.

8. 10 CFR 70.74 states 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

AES EREF management measures program are contained in Section 11.4.3 of NUREG-1520

(NRC, 2002).

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11.3 Staff Review And Analysis

11.3.1 Configuration Management (CM)

NRC staff has reviewed AES's CM function in accordance with the regulatory acceptance

criteria of Section 11.4.3.1 of NUREG-1520 (NRC, 2002).

AES's CM program is implemented through the QA program requirements as described in

EREF QAPD (see Section 11.4.8 of this SER for a discussion of the QAPD) and associated

procedures. The AES President has the overall responsibility for the establishment and

implementation of the QA program. The CM program is described in Section 11.1 of the Safety

Analysis Report (SAR) (AES, 2009a).

11.3.1.1 CM Policy

As described in Section 11.1.1 of the SAR (AES, 2009a), through the CM system, AES will

control any documentation that can create changes to onsite, structures, processes, systems,

equipment, components, computer programs, and activities of personnel-including the ISA.

The Engineering Organization will be responsible to maintain the CM program as the project

goes from design and construction to the operational phase. The Engineering Organization will

have lead engineers in charge of each discipline that will also be responsible for the

interdisciplinary reviews for design changes. After issuance of the license, the Engineering

Manager will be responsible for design of and modification to facility SSCs.

Section 11.1.1 of the SAR (AES, 2009a) also explains that design changes will be reviewed in

accordance with formal procedures including, as a minimum, a review for ISA impacts.

Changes to the facility or to activities of personnel will be evaluated in accordance with the

requirements of 10 CFR 70.72, as applicable. In addition, for changes that involve or could

affect the uranium onsite, a Nuclear Criticality Safety (NCS) evaluation or analysis, as

applicable, will be prepared and approved. Each change or modification will also be evaluated

and documented for radiation exposure to minimize worker exposures and to maintain the

facility within the as low as reasonably achievable (ALARA) program. After completion of any

modification, the Engineering Manager-or designee- ensures that the applicable testing has

been completed to the affected system, that the system operates correctly, and that the

modification and documentation are complete.

The scope of the CM program will include all the QA Level 1 and QA Level 2 items and

activities. The scope of QA Level 1 and QA Level 2, as described in Section 11.1.1.1 of the

SAR (AES, 2009a), includes documents that expand through design, construction, initial startup

and operations, including the appropriate documentation for the processes. During design,

documents such as calculations, safety analyses, design criteria, engineering drawings, system

descriptions, technical documents, and specifications will be maintained in CM when initially

approved. During construction, documents such as vendor data, test data, inspection data, and

applicable procedures will be included. Drawings and specifications prepared and issued for

procurement, fabrication or construction of QA Level 1 and QL Level 2 items and activities will

be included in CM. The implementing documents that support the CM will be controlled within

the document control system to ensure that only reviewed and approved documentation is used.

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The applicant describes the functional interfaces of the CM program with other management

measures in Section 11.1.1.2 of the SAR (AES, 2009a); those management measures include

QA, records management, maintenance, training and qualifications, incident investigations,

audits and assessments, and procedures. Three areas of special note are QA, maintenance,

and records management. The QA program specifically establishes the framework for the CM

and other management measures. Maintenance will be established as part of the design basis

and controlled under the CM program. Records management will provide evidence on the

conduct of the activities associated with CM and other management measures.

In Section 11.1.1.3 of the SAR (AES, 2009a), the application states that the objectives of CM

will be to ensure design and operation within the design basis of QA Level 1 and QA Level 2

items and activities. Furthermore, the application states that the objectives will be met by

identifying and controlling the preparation and review of documentation and changes of QA

Level 1 and QA Level 2 items. The physical configuration of the facility will be maintained

consistent with the approved design.

The applicant provides a description of the CM activities in Section 11.1.1.4 of the SAR

(AES, 2009a). Design activities included in the CM program will be conducted through a

systematic process of preparation, review and approval. The process will ensure consistency

between the design and design bases of the QA Level 1 and QA Level 2 items and activities

during the design and construction phases. In addition, it will include the activities for

QA Level 1 and QA Level 2 items during operations to ensure the activities are within the limits

and constraints established in the ISA-and that changes to the facility are controlled in

accordance with 10 CFR 70.72. Finally, the CM program will include training records of

personnel conducting activities associated with QA Level 1 and QA Level 2 items and activities

to ensure that only qualified personnel are performing the work.

As described in Section 11.1.1.5 of the SAR (AES, 2009a), the interfaces between AES and

contractors or among contractors performing quality-related activities will be documented. AES

and contractors will have the responsibility to identify quality problems and elevate their

concerns to the appropriate management if there is disagreement among the parties.

11.3.1.2 Design Requirements

The design requirements will be established and maintained by the Engineering Organization

during design, construction and operations. Section 11.1.2 of the SAR (AES, 2009a) states that

design requirements will be documented in a design requirements document that provides a

hierarchical distribution of the requirements through design basis documents. Design

documents associated with QA Level 1 and QA Level 2 will be based on design control

provisions and associated procedural controls over design to establish and maintain the

technical baseline (AES, 2009a). The design documents will be subject to interdisciplinary

review and design verification, as applicable.

Qualified individuals will prepare design documents ensuring that the appropriate codes,

standards, and licensing commitments are included (AES, 2009a). Any deviations or changes

will be identified in the design documentation package. The design documentation package will

be reviewed by an additional qualified individual for concept and conformity to design inputs

emphasizing conformance with the applicable codes, standards and licensing commitments

(AES, 2009a). The individual performing the review will be independent and will have the

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authority to delay the approval process until resolution of any issues. The Engineering Manager

will document the review, and the manager having the responsibility for the design function will

approve the document. The QA Manager will audit the design control process through

augmented audit teams.

The design verification will be performed by qualified individuals other that those who performed

the design (AES, 2009a). Any verification performed by supervisors of the individuals doing the

design will be documented and approved by the supervisor's management. Any changes to QA

Level 1 and QA Level 2 design documents will be reviewed, checked, and approved

commensurate with the original approval requirements to ensure consistency with the design

bases (AES, 2009a).

The responsible engineer will send approved design documents to the document control center

for distribution (AES, 2009a). When required, the document control center will maintain the

required receipt verification in the file.

The design interface will be maintained by communication among the appropriate personnel

(AES, 2009a). During the operational phase, any design changes will be provided to pertinent

personnel to ensure the correct performance of their duties.

As described in Section 11.1.2.1 of the SAR (AES, 2009a), EREF QAPD requires procedures to

give specifications for the work performed. The procedures, instructions and drawings will

incorporate the acceptance criteria established by design. Procedures will also be reviewed to

ensure that they are maintained up-to-date with facility configuration and regulatory

requirements (AES, 2009a).

11.3.1.3 Document Control

As described in Section 11.1.3 of the SAR (AES, 2009a), procedures will be established to

control the preparation and issuance of documents. Measures will ensure that documents,

including revisions, are adequately reviewed, approved and released for use by authorized

personnel (AES, 2009a). The document control procedures will establish the distribution

requirements and controls to ensure that documents are transmitted and received in a timely

manner at the appropriate locations. The controlled copies will be distributed to the persons

performing the activity for their use.

The applicant will use an electronic document management system to file project records and to

make available the official (controlled) copy of the current documents. The system will index the

controlled documents through unique numbering of the documents, including revision numbers.

Personnel will be trained to use the system to retrieve controlled documents. Superseded or

cancelled documents will be appropriately labeled and maintained as records.

Documents that are relevant and relied on for safety through the document control and records

management procedures will be captured as described in Section 11.1.3 of the application

(AES, 2009a). Some examples of these documents include design requirements, design bases,

ISA of IROFS, NCS analyses, NCS evaluations, as-built drawings, specifications, procedures

that are IROFS, procedures involving training, QA/Quality Control (QC) documentation,

maintenance, audits and assessments reports, emergency operating procedures, emergency

response plans, assessment reports and engineering documents.

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11.3.1.4 Change Control

Section 11.1.4 of the SAR (AES, 2009a) describes the change control process and states that

procedures will be used to control changes to the technical baseline including an appropriate

level of technical, management, and safety review and approval prior to implementation.

Changes will be controlled throughout the design, construction, and operations phases to

maintain consistency among the design requirements, physical configuration and facility

documentation. The ISA and other documents affected by design bases of QA Level 1 and QA

Level 2 items and activities including the design requirement documents and basis of design

documents, will be systematically reviewed and modified to reflect design and operational

changes from an established safety basis prior to implementation.

As described in Section 11.1.4.1 of the SAR (AES, 2009a) during the design phase, a

systematic interdisciplinary review will ensure that: (1) design changes do not impact the ISA;

(2) if any changes impact the ISA, they are accounted for in subsequent ISA changes; or (3)

changes are not approved or implemented before review. Potential changes that reduce the

level of commitment or margin of safety in the design bases of QA Level 1 and QA Level 2 items

and activities will be submitted to the NRC for review and approval prior to implementation of

the change.

As described in Section 11.1.4.2, of the SAR (AES, 2009a) during the construction phase, any

changes to documents issued for construction, fabrication and procurement will be documented,

reviewed, approved, and posted against each affected design document (AES, 2009a). The

changes will continue to be evaluated against the approved design bases of QA Level 1 and QA

Level 2 items and the ISA. Upon issuance of the license, the applicant is mandated by the

10 CFR 70.72 to implement a change control process that includes the reporting of changes

made without prior NRC approval and submit a license amendment for changes that require the

NRC's approval before implementation (AES, 2009a).

Section 11.1.4.3 of the SAR (AES, 2009a) states that during the operations phase, changes will

be documented reviewed and approved prior to implementation. For changes made in

accordance with 10 CFR 70.72, the applicant will have measures in place to ensure that

changes to the onsite documentation are made promptly to avoid inadvertent access by

responsible facility personnel to outdated information that may affect the performance of their

duties. The applicant will evaluate each change or modification to the facility for required

changes to procedures, personnel training, testing program, or regulatory documents such as

the ISA, ISA Summary, SAR, QA Program, Environmental Report, Physical Security Plan,

Emergency Plan, Fundamental Nuclear Material Control Plan, and Standard Practice and

Procedure Plan for the Protection of Classified Matter. The applicant states that measures will

be established to ensure that the quality of the facility's SSCs is not compromised by planned

changes. The modification process will be described in administrative procedures that are

approved by the Engineering and QA Managers. The procedures will include the requirements

that must be met to implement the modification and the requirements for initiating, approving,

monitoring, designing, verifying and documenting the modification. The modifications will be

evaluated to ensure consistency among the facility's procedure, personnel training, testing

program, and regulatory documents. Other areas that the applicant will consider during the

evaluation of changes or modifications are: radiation exposure, modification cost, lessons

learned, QA aspects, potential operability or maintainability concerns, constructability concerns,

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post-modification testing requirements, environmental considerations, and human factors. After

the completion of the modification or change, the system will be tested and personnel are

trained to ensure correct operations. When the system becomes operational, all the required

documentation will be distributed to operations and maintenance including formal notice to all

appropriate managers.

11.3.1.5 Assessments

As described in Section 11.1.5, of the SAR (AES, 2009a) periodic assessments of the CM

program will be conducted to determine the program effectiveness and to correct deficiencies.

Document assessments and system walkdowns will be planned, conducted, and documented in

accordance with the facility's audit and assessment program as described in Section 18 of

EREF QAPD.

11.3.1.6 Evaluation Findings

As described in the preceding sections of this SER, the staff reviewed the applicant's CM

function following the guidance in Section 11.4.3.1 of NUREG-1520 (NRC, 2002). Based on

this review, the staff found that the applicant described in Section 11.1 of the SAR (AES, 2009a)

the following elements of the CM function: the CM program (Section 11.1.1 of the SAR), design

requirements (Section 11.1.2 of the SAR), document control (Section 11.1.3 of the SAR),

change control (Section 11.1.4 of the SAR), and assessments (Section 11.1.5 of the SAR).

Also, the CM program is described in Section 11.1.1 of the SAR in terms of its policy, objectives,

activities, scope of structures systems and components, organizational structure and its

interfaces. The application described how design requirements and associated design bases

are established and are maintained through control of the design process. Technical

management review is also described including approval functions. The applicant provided a

description of the functional interfaces of CM with other management measures that addresses

each management measure individually.

In Section 11.1.2 of the SAR (AES, 2009a), the application described a design process leading

to drawings and other statements of requirements that proceeds logically from the design basis.

The applicant also described how design requirements and associated design bases are

established and are maintained through design control provisions. The technical management

review and approval functions are adequately described.

In Section 11.1.3 of the SAR (AES, 2009a), the applicant described an acceptable method to

create and control documents that are relied on for safety. These documents will include design

requirements, design basis, ISAs, NCS analyses, NCS evaluations, as-built drawings,

specifications, all procedures that are IROFS, procedures involving training, QAIQC

documentation, maintenance, audits and assessments, emergency operating procedures,

emergency response plans, system modification documents, assessment reports, and others

that the applicant deems part of CM.

In Section 11.1.4 of the SAR (AES, 2009a), the applicant described how the CM function will

maintain strict consistency among the design requirements, the physical configuration, and the

facility documentation during design, construction and operations. This section described how

changes to the technical baseline are controlled through procedures to ensure that each change

is evaluated, implemented, and tracked. Sections 11.1.4.2 and 11.1.4.3 of the SAR (AES,

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2009a) described an acceptable process for providing reasonable assurance that the ISA is

systematically reviewed and modified to reflect design or operational changes from an

established safety basis and that all documents outside the ISA that are affected by safety-basis

changes are properly modified, authoritatively approved, and made available to personnel. The

applicant also described the documentation process following changes made to ensure that the

change process fully implements the provisions of 10 CFR 70.72.

The applicant described in Section 11.1.5 of the SAR (AES, 2009a) the assessments of the CM

program. The section confirms that periodic assessments of the CM function will be conducted

to determine the program's effectiveness and to correct deficiencies. The applicant indicated in

Section 11.1.5 of the SAR that such assessments will be planned and conducted in accordance

with Section 18 of the EREF QAPD.

Thus, the CM program described by the applicant in Section 11.1 and its subsections is

consistent with the guidance provided in Section 11.4.3.1 of NUREG-1 520 (NRC, 2002) and is,

therefore, acceptable.

11.3.2 Maintenance

The NRC staff has reviewed the maintenance program following the regulatory acceptance

criteria in Section 11.4.3.2 of NUREG-1520 (NRC, 2002).

The applicant describes the maintenance and functional testing programs for the operations

phase of the facility. The maintenance organization will be responsible for planning, scheduling,

tracking, and maintaining the records for maintenance activities.

As described in Section 11.2 of the SAR (AES, 2009a), the methods and practices that will be

applied, to the corrective, preventive, and functional tests maintenance elements will include, as

applicable: (a) part lists; (b) as-built or redlined drawings; (c) a notification to the operations

function before conducting repairs and removing an IROFS form 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 removal of components from service for maintenance and for

return to service; (h) ensuring safe operations during the removal of IROFS from service; and

(i) notification to operations personnel that repairs have been completed (AES, 2009a).

Written procedures for the performance of maintenance will be prepared and will include steps

(a) through (i) as listed above. Maintenance procedures will include reviews by the various

safety disciplines, including criticality, fire, radiation, industrial, and chemical process safety.

AES states that the procedures describe, as a minimum, the following: (a) qualifications of

personnel authorized to perform the maintenance, functional testing, or surveillance/monitoring;

(b) controls on and specification on any replacement components or materials to be used;

(c) post-maintenance testing to verify operability of the equipment; (d) tracking and records

management of maintenance activities; and (e) safe work practices.

As applicable, contractors that work on or near IROFS identified in the ISA Summary will be

required by the applicant to follow the same maintenance guidelines described for the AES

maintenance function.

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Section 11.2 of the SAR (AES, 2009a) also states that maintenance procedures for QA Level 1

and QA Level 2 items and activities will commit, as applicable to: (a) reviews of the work to be

performed for pre-maintenance activities; (b) NCS evaluation and analysis, if required, for new

procedures or work activities that involve or could affect uranium onsite; (c) steps that require

notification of affected parties before performing work and completion of maintenance work; and

(d) control of work by comprehensive procedures to be followed by maintenance technicians.

11.3.2.1 Surveillance/Monitoring

In Section 11.2.1 of the SAR (AES, 2009a) the surveillance/monitoring is described as the

process that will be used to detect degradation and adverse trends of IROFS so that action may

be taken prior to component failure. The parameters to be monitored will be selected based

upon their ability to detect the predominant failure modes of the critical components.

Surveillance of QA Level 1 and QA Level 2 items will be performed at specified intervals. The

surveillance activity supports the determination of performance trends for IROFS indicating

when potential performance degradation exists, adjusting the preventive maintenance

frequencies or taking any other corrective action. Moreover, surveillance/monitoring results will

be evaluated to determine any impact on the ISA or any updates needed. For surveillance tests

that can only be done while the equipment is out of service, proper compensatory measures will

be prescribed in maintenance procedures.

Section 11.2.1 also describes how incident investigations may identify root causes of failures

that are related to the type or frequency and maintenance. These incident investigation results

will be used as lessons learned in the surveillance/monitoring and preventive maintenance

program as appropriate. Records showing the current surveillance schedule, performance

criteria, and test results for IROFS will be maintained in accordance with the Records

Management System.

11.3.2.2 Corrective Maintenance

As described in Section 11.2.2 of the SAR (AES, 2009a), the corrective maintenance will involve

repairs or replacement of equipment that has unexpectedly degraded or failed. The program

will provide for restoration of a QA Level 1 or QA Level 2 item to acceptable performance

through a planned, systematic, controlled, and documented approach for the repair and

replacement activities. After conducting a corrective maintenance and before returning QA

Level 1 or QA Level 2 items to operational status, the applicant will conduct a functional test, if

necessary, to ensure that the QA Level 1 or QA Level 2 items will perform their intended safety

functions. Corrective maintenance results will be evaluated to determine any impact on the ISA

or any updates needed.

11.3.2.3 Preventive Maintenance

Section 11.2.3 of the SAR (AES, 2009a) states that preventive maintenance (PM) will include

pre-planned and scheduled periodic refurbishment, partial or complete overhaul, or replacement

of QA Level 1 and QA Level 2 items, if necessary, to ensure their continued safety function.

The applicant will use the results of surveillance and monitoring, including failure history, to plan

PM activities.

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Calibration and testing are also described in Section 11.2.3 of the SAR (AES, 2009a) within the

PM function. Calibration standards will be used by facility personnel to calibrate equipment and

monitor devices to plant safety and safeguards. The applicant will provide compensatory

measures during testing of non-redundant QA Level 1 and QA Level 2 items to ensure that their

function is performed until the item is back in service.

The applicant will use industry experience, operating data, surveillance data, and plant

equipment operating experience to determine initial PM frequency and procedures. The

applicant states that in determining the frequency of PM, consideration will be given to

appropriately balancing the objective of preventing failures through maintenance against the

objective of minimizing the unavailability of IROFS because of PM. Feedback from PM,

corrective maintenance, and incident investigation will be used, as appropriate, to modify the

frequency or scope of PM activities. The applicant will document the rationale for any PM

deviations from industry standards or vendor recommendations.

After conducting PM and before returning QA Level 1 or QA Level 2 item to operational status,

the applicant will conduct a functional test, if necessary, to ensure that the QA Level 1 or

QA Level 2 item will perform its intended safety function. Records pertaining to PM will be

maintained in accordance with the Records Management System. PM activities results will be

evaluated to determine if there are impacts on the ISA or if there are updates needed.

11.3.2.4 Functional Testing

Functional testing for QA Level 1 and QA Level 2 items as described in Section 11.2.4 of the

SAR (AES, 2009a) will be performed, as appropriate, following initial installation as part of

periodic surveillance testing, and after corrective or preventive maintenance or calibration to

ensure that the item is capable of performing its safety function, when required.

The applicant's overall testing program is divided into two major testing programs: preoperational testing and operational testing program. The objectives of these programs as

described in Section 11.2.4.1 of the SAR (AES, 2009a) are to ensure that IROFS: (1) have

been adequately designed and constructed; (2) meet contractual, regulatory, and licensing

requirements; (3) do not adversely affect worker or the public's health and safety; and (4) can

be operated in a dependable manner so as to perform their intended function. The programs

also ensure that operating and emergency procedures are correct and that personnel have

acquired the correct level of expertise.

As described in Section 11.2.4.2 of the SAR (AES, 2009a) test requirements will be specified in

written procedures, except that in lieu of written procedures, appropriate sections of related

documents (e.g., American Society for Testing and Materials methods, external manuals, and

maintenance instructions) may be used. The applicant describes the content of test

procedures. Procedures will be detailed enough that qualified personnel can perform the

required functions without direct supervision.

The pre-operational testing program is described in Section 11.2.4.3 of the SAR (AES, 2009a)

and will consist of three parts: constructor turnover, pre-operational functional testing, and initial

start-up testing. During the constructor turnover, the constructor will be responsible for

completion of as-built drawing verification, purging, cleaning, vacuum testing, system turnover,

and initial calibration of instrumentation in accordance with the design and installation

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specifications that were provided by the vendors and architect engineers. The applicant will

conduct pre-operational functional testing following constructor turnover to initially determine

various facility parameters and to initially verify the capability of SSCs to meet performance

requirements. The applicant will perform initial start-up testing beginning with the introduction of

uranium hexafluoride (UF6) and ending with the start of commercial operation.

The pre-operational testing program objective will be to perform all necessary tests to verify that

the QA Level 1 and QA Level 2 items that are essential to the safe operation of the plant are

capable of performing their intended safety function. The applicant will perform initial start-up

testing to ensure safe and orderly UF6 feeding, and to verify parameters assumed in the ISA.

The overall program will be reviewed and approved by the Plant Manager prior to the initial UF6

introduction. Results of each preoperational test will be reviewed and approved by the

responsible department manager or designee. The results for start-up testing will be reviewed

and approved by the Start-Up Manager.

The applicant describes the operational testing program in Section 11.2.4.4 of the SAR (AES,

2009a). The program will consist of periodic testing and special testing. The applicant will

conduct periodic testing to monitor various facility parameters, and to verify continuing integrity

and capability of QA Level 1 and QA Level 2 items. The applicant will conduct special testing

which consists of testing that does not fall under any other testing program and is of a nonrecurring nature. The procedures for periodic and special testing will be sufficiently detailed that

qualified personnel can perform their functions without supervision.

The overall responsibility for the operational program will be shared among the Maintenance

Manager; the Operations Manager; and the Environmental, Health, Safety and Licensing

(EHS&L) Manager. The applicant also describes the responsibilities for Test Coordinators

during operational testing in Section 11.2.4.4 of the SAR (AES, 2009a).

As described in Section 11.2.4.4.1 of the SAR (AES, 2009a) periodic testing will consist of

testing conducted to verify the continuing capability of QA Level 1 and QA Level 2 items to meet

performance requirements. The testing will begin during the pre-operational stage and continue

to be scheduled through the facility's life. The testing schedule will be revised periodically to

ensure that the changes in the testing requirements are reflected. The Maintenance

Department will maintain a testing status index to assist groups in assuring that surveillances

are being completed within the specified time interval. The applicant will process any

occurrence of missed tests through their Corrective Action Program (CAP).

Special testing, as described in Section 11.2.4.4.2, of the SAR (AES, 2009a) will consist of

testing conducted at the facility that is not a facility pre-operational test, periodic test, postmodification test or post-maintenance test. Some of the purposes for special testing will be: (1)

acquisition of particular data for special analysis; (2) determination of information relating to

facility incidents; (3) verification that required corrective actions reasonably produce expected

results and do not adversely affect the safety of operations; and (4) confirmation that facility

modifications reasonably produce expected results and do not adversely affect systems,

equipment, and/or personnel by causing them to function outside established design conditions

(applicable to testing performed outside of a post-modification test). The applicant will consider

special tests as facility changes or changes in process safety information that will require to be

reviewed by the ISA method and in accordance with 10 CFR 70.72(a) and (c). The applicant

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does not intend to include routine surveillances, normal operation evaluations, and activities

where which there is previous experience in the conduct and performance of the activity within

special testing.

11.3.2.5 Evaluation Findings

As described in the preceding sections of this SER, the staff reviewed the applicant's

maintenance program following the guidance in Section 11.4.3.2 of NUREG-1520 (NRC, 2002).

In Section 11.2 of the SAR (AES, 2009a) the applicant described the maintenance program

which includes surveillance and monitoring (Section 11.2.1 of the SAR), corrective maintenance

(Section 11.2.2 of the SAR), PM (Section 11.2.3 of the SAR), and functional testing (Section

11.2.4 of the SAR). The applicant described how the maintenance function will be designed

and will have measures in place to ensure that the objective of preventing failures through

maintenance is appropriately balanced against the objective of minimizing unavailability of

IROFS because of monitoring or PM.

In Section 11.2.1 of the SAR (AES, 2009a), the applicant described the surveillance/monitoring

function. The description included the statement that surveillance activities will be conducted at

specified intervals. The applicant also described how the surveillance activity supports the

determination of performance trends for IROFS, thus providing data useful in determining PM

frequencies. The applicant also provides an adequate description of the records retention for

surveillance activities. Furthermore, for surveillance tests that can be done only while IROFS

are out of service, the applicant stated that proper compensatory measures will be prescribed

for the continued normal operation of a process. The applicant provided a description of how

the results of incident investigations are used to modify the affected maintenance function and

eliminate or minimize the root cause.

In Section 11.2.2 of the SAR (AES, 2009a), the applicant described the approach for corrective

maintenance. The applicant described how the maintenance function provides a planned,

systematic, integrated, and controlled approach for the repair and replacement activities

associated with identified unacceptable performance deficiencies of IROFS.

In Section 11.2.3 of the SAR (AES, 2009a), the applicant described PM which includes

preplanned and scheduled periodic refurbishing, or partial or complete overhaul, of IROFS to

minimize occurrences of their unanticipated losses. The applicant also described how the

retention of records showing the PM schedule and results for all IROFS subject to PM will be

handled. In addition the applicant addressed calibration as part of the maintenance function.

The applicant provided the methodology that will be used to determine the PM frequency and

also provided how the PM function will be designed to assure that the objective of preventing

failures through maintenance will be appropriately balanced against the objective of minimizing

unavailability of IROFS because of monitoring or PM.

In Section 11.2.4 of the SAR (AES, 2009a), the applicant described the methods and approved

procedures used to perform functional testing, as needed, of IROFS after PM or corrective

maintenance. The applicant also described how functional tests are designed to include all

operational aspects of the IROFS during startup of new processes. The applicant described the

handling and maintenance of records showing the functional test schedule and results for all

IROFS subject to functional testing.

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Based on the above, the staff found that the maintenance program description provided by the

applicant in Section 11.2 and its subsections is consistent with the guidance provided in

Section 11.4.3.2 of NUREG-1520 (NRC, 2002) and is, therefore, acceptable.

11.3.3 Training and Qualifications

The NRC staff has reviewed the applicant's training and qualification program following the

regulatory acceptance criteria in Section 11.4.3.3 of NUREG-1 520 (NRC, 2002).

The applicant's QA program provides training and qualification requirements during design,

construction, and operations phases for QA training of personnel performing QA Level 1 and QA

Level 2 work activities; for non-destructive examination, inspection and test personnel; and for

QA auditors. As described in Section 11.3 of the SAR (AES, 2009a), the principal objective of

the training program system is to ensure job proficiency of all of all facility personnel involved in

QA Level 1 and QA Level 2 activities through effective training and qualification.

11.3.3.1 Organization and Management of the Training Function

As described in Section 11.2.1 of the SAR (AES, 2009a), line management will be responsible

for the content and effective conduct of training. The position description for line managers

includes their training responsibilities; they are given the authority to implement training for their

personnel supported by the training organization. The Training Manager will be responsible for

the facility training programs. The applicant will use performance-based training as the primary

management tool for analyzing, designing, developing, conducting, and evaluating training.

The applicant will have administrative procedures establishing requirements for indoctrination

and training for personnel performing activities relied on for safety and to ensure that all phases

of training are conducted reliably and consistently. The applicant will grant exception from

training requirements when justified, documented, and approved by appropriate management.

The applicant will include affected lesson plans in the change control process when any design

changes or facility modifications are implemented. The applicant will maintain programmatic

and individual training records to support management information needs associated with

personnel training, job performance, and qualifications. Individual records will include general

employee training, technical training, and employee development training conducted at the

facility.

11.3.3.2 Analysis and Identification of Functional Areas Requiring Training

As described in Section 11.3.2 of the SAR (AES, 2009a), a needs/job analysis will be performed

and will identify tasks to ensure that appropriate training is provided to personnel working on

tasks related to QA Level 1 and QA Level 2 items and activities. On-the-job training (OJT) will

include Job Hazards Analysis to provide the employees with the skills necessary to perform

their jobs safely.

Relevant technical and management personnel will be consulted as necessary to identify

activities where training is appropriate. The applicant will compare and review the activities

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selected for training with training materials as part of a training effectiveness evaluation. The

applicant will update the list of activities selected for training as necessitated by changes in

procedures, processes, plant systems, equipment, or job scope.

11.3.3.3 Position Training Requirements

As stated in Section 11.3.3 of the SAR (AES, 2009a), minimum training requirements will be

developed for positions whose activities are relied on for safety. The initial identification of jobtraining requirements will be based on experience; and the level at which an employee will

initially enter the training program will be determined by an evaluation of the employee's past

experience, level of ability and qualifications. The applicant's position descriptions will describe

the entry level criteria for positions whose activities are relied on for safety. The applicant also

states that facility personnel may be trained through participation in general employee training,

technical training, or employee development/management-supervisory training.

Training requirements will be applicable to personnel within the plant organization who have

direct relationship to the operation, maintenance, testing, or other technical aspects of the

facility's IROFS. To maintain personnel proficiency, the applicant will conduct continuing and

periodic retraining.

General Employee Training

Section 11.3.3.1 of the SAR (AES, 2009a) describes the general employee training as those

including QA, radiation protection, safety, emergency, and administrative procedures

established by facility management and applicable regulations. It states that persons that are

under the supervision of facility management, including contractors, must participate in general

employee training; however, certain facility support personnel may not participate in all topics

depending on their normal work assignment. The general employee training will include the

following topics: general administrative controls and procedure use, QA policy and procedures,

facility systems and equipment, nuclear safety, industrial safety, health and first aid, emergency

plan and implementing procedures, facility security programs, chemical safety, fire protection

and fire brigade, and new employee orientation.

Nuclear Safety Training

As described in Section 11.3.3.1.1 of the SAR (AES, 2009a), training programs will be

established for various types of job functions commensurate with criticality and radiation safety

responsibilities associated with each function. The nuclear safety training program will include

topics such as notices, reports and instruction to workers; practices designed to keep radiation

exposures ALARA; methods of controlling radiation exposures; contamination control methods;

use of monitoring equipment; emergency procedures and actions; nature and sources of

radiation; safe use of chemicals; biological effects of radiation; use of personnel monitoring

devices; principles of nuclear criticality safety; risk to pregnant females; radiation protection

practices; protective clothing; respiratory protection; and personnel surveys. Also Section

11.3.3.1.1 states that the criticality safety program will be in accordance with ANSI/ANS 8.19 -

1996 and ANSI/ANS 8.20 - 1991.

The training sessions will be conducted by instructors assigned by the EHS&L Manager. The

trainings will have an initial examination that must be passed to demonstrate understanding and

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effectiveness of the training. Employees will be required to complete the nuclear safety training

before having unescorted access into controlled areas. Unescorted access to controlled areas

requires annual retraining.

The EHS&L Manager or designee will periodically revise and update the nuclear safety training

programs to ensure that they are still current and adequate. The radiation protection sections

will be reviewed at least annually or by the request of the Radiation Protection/Chemistry

manager or designee. Also, supervisors will review radiation safety topics at least annually with

their workers.

Fire Brigade Training

Section 11.3.3.1.2 of the SAR (AES, 2009a) describes the primary purpose of the fire brigade

training program as to develop a group of employees that will be skilled in fire prevention, firefighting techniques, first-aid procedures, and emergency response. Its intention is not to

replace the local fire fighters. They will be the first-response and will supplement the local fire

fighters. The program will provide initial training, semi-annual classroom training and drills,

annual practical training, and leadership training for the fire brigade leaders.

Technical Training

Technical Training, as described by the applicant in Section 11.3.3.2 of the SAR (AES, 2009a),

will be designed, developed and implemented for assisting employees to understand applicable

fundamentals, procedures, and practices common to a gas centrifuge uranium facility. In

addition, the technical training will be used to develop the skill to perform the assigned work.

The technical training will consist of initial training, OTJ training and qualifications, continuing

training, and special training.

Initial Training

As described in Section 11.3.3.2.1 of the SAR (AES, 2009a), initial training will be used to

provide understanding of the fundamentals, basic principles, and procedures related to the

employees' work activities. Initial training will consist of-but not limited to-live lectures, taped

and filmed lectures, self-guided study, demonstrations, laboratories, workshops, and OJT

training.

The training requirements for certain new or transferred employees that are partially qualified

from previous, applicable training or experience will be determined by applicable regulations;

performance in review sessions; comprehensive examinations; or other techniques that can

identify the employee's level of ability. Allowances will be made to suit specific situations,

depending on regulatory requirements and individual needs. The trainee progress will be

evaluated by written examinations and oral and written tests.

The training program will be arranged in logical blocks or modules and provides brief

descriptions of the modules. The description of operations, mechanical maintenance,

instrumentation and electrical maintenance, health physics and chemistry, and

engineering/professional will be included as part of initial training modules.

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ODerations Initial Trainina

The applicant describes four modules for the operation's initial training: general systems,

specific systems, nuclear preparatory, and plant familiarization. The general systems module

will provide the trainees with basic concepts and fundamentals in mathematics, physics,

chemistry, heat transfer, and electrical theory. The applicant may conduct OJT orientation.

The specific training module will provide basic instruction in system and component

identification, and basic system operating characteristics. The training module will provide a

general overview of enrichment plant equipment and acquaints the trainees with enrichment

plant terminology and nomenclature. It also provides instructions describing basic systems

operations.

The applicant states that the nuclear preparatory training module will be presented to operations

personnel following the specific systems module. The module develops the necessary concepts

in basic nuclear physics, plant chemistry, basic thermodynamics, radiation protection, and

enrichment theory.

The plant familiarization module will be provided by the applicant as orientation for employees to

become familiar with plant layouts, plant systems, and practicable laboratory and equipment

work at the facility.

Mechanical Maintenance Initial Training

The applicant describes three modules for the mechanical maintenance initial training: general

systems, fundamental shop skills, and plant familiarization. The general systems module will be

the same as the previously described course in the Operations Initial Training.

The fundamental shop skills module, as described by the applicant, will provide instruction in

fundamentals of mechanical engineering performance. The applicant will combine academic

instruction and hands-on training in this module to familiarize the trainees with design

operational and physical characteristics of enrichment facility components and basic skills and

procedures used to perform mechanical repairs and equipment replacement. The applicant will

use task training lists to assure the trainees level of performance. Trainees will be able to use

work procedures to guide them through an assigned task. All the phases of this module will

stress the radiological and industrial safety.

The plant familiarization module will be the same course as noted previously.

Instrumentation and Electrical Maintenance Traininq

The applicant describes four modules for the instrumentation and electrical maintenance initial

training: general systems, basic instrument and electrical, basic performance and plant

familiarization. The general systems module will be the same as the previously described

course in the Operations Initial Training.

The basic instrument and electrical module, as described by the applicant, will provide trainees

with refresher training in electrical and electronic fundamentals, digital techniques and

application, instrumentation and control theory and application, and an introduction to the types

and proper use of measuring and test equipment commonly used in enrichment facilities. Also,

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the applicant will provide working knowledge of nuclear and non-nuclear instrumentation

systems; overall integrated plant operation and control; and, in particular, the hazards of

calibration during plant operation.

The applicant explained the fundamental performance modules as the performance

familiarization with plant test procedures, test equipment, and testing-as well as plant records,

reports, and data collection. As part of this module the applicant will provide training to trainees

for basic understanding of thermodynamics used in testing plant heat transfer.

The plant familiarization module will be the same course as noted previously.

Health Physics and Chemistry Initial Training

The applicant describes four modules for the health physics and chemistry initial training:

general systems, fundamental health physics, fundamental chemistry, and plant familiarization.

The general systems module will be the same as the previously described course in the

Operations Initial Training.

The applicant will use the fundamental health physics training to present trainees with a

comprehensive and technical understanding of the nuclear processes. The module will utilize

non-automated counting and spectrographic equipment, and portable survey instruments. The

applicant describes the training as providing detailed overview of administrative material.

The fundamental chemistry module, as described by the applicant, will provide familiarization

with chemistry theory, techniques, and procedures to ensure that chemistry technicians can

work safely and competently.

The plant familiarization module will be the same course as noted previously.

Engineer/Professional Initial Training

The applicant's engineer/professional training program will be for technical staff and managers.

It is described in four modules: facility orientation; basic engineer/professional training;

enrichment/chemical engineer/professional training; and engineer/professional systems training.

The facility's orientation training module, as described by the applicant, will provide an

orientation to each section of the facility-including an OJT task list with objectives that must be

completed while working on the section (AES, 2009a).

The basic engineer/professional training module, as described by the applicant, will provide a

basic understanding of uranium enrichment, systems and components required for final product,

and the interrelationship that is required among facility organizations to achieve the overall

objective (AES, 2009a).

The enrichment/chemical engineer/professional training, as described by the applicant, will

provide specific theoretical information related to the operations of the enrichment plant (AES,

2009a). Some of the topics, such as thermal science and nuclear physics, will address the

applications in the enrichment facility (AES, 2009a).

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The engineer/professional systems training module, as described by the applicant, will provide

an overview of plant systems, components, and procedures that are required to operate the

facility safely and efficiently (AES, 2009a).

OJT Training and Qualifications

The applicant describes the OJT as a systematic method of providing the required job-related

skills and knowledge for a position. OJT will be performed in the work area, and it will

supplement the formal classroom training. As described in Section 11.3.3.2.2 of the SAR

(AES, 2009a), the objective of the program is to assure the trainee's ability to perform job tasks

as described in task descriptions and the Training and Qualification Guides.

Continuing Training

Continuing training is defined by the applicant in Section 11.3.3.2.3 of the SAR (AES, 2009a) as

any training not provided as initial qualification and basic training which improves job-related

knowledge and skills. The categorization of continuing training will be as follows:

" Facility systems and component changes;

" OJT/Qualifications program retraining;

* Policy and procedure changes;

* Operating experience program documents review to include industry and in-house operating

experiences;

* Continuing training required by regulation (e.g., emergency plan training);

* General employee, special, administrative, vendor, and/or advanced training topics

supporting tasks that are elective in nature;

* Training identified to resolve deficiencies (task-based) or to reinforce seldom-used

knowledge skills;

* Refresher training on initial training topics;

" Structured pre-job instruction, mock-up training, and walkthroughs; and

* Quality awareness

The applicant's continuing training will consist of formal and informal components performed at a

certain frequency, ensuring proficiency on the job. A systematic approach will be used to

determine the content of continuing training that can be offered, as needed, in any of the topics

previously mentioned.

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Special Traininq

Special training, as described in Section 11.3.3.2.4 of the SAR (AES, 2009a), involves those

subjects of a unique nature required for a particular area of work. The applicant will provide

special training to selected personnel based on specific needs. Special training will not be

directly related to disciplinary lines.

11.3.3.4 Basis and Objectives for Training

The applicant's learning objectives are described in Section 11.3.4 of the SAR (AES, 2009a)

and will identify the training content based on needs/jobs analyses and position-specific

requirements. The objectives will include the knowledge, skills and abilities to be demonstrated;

the conditions under which these actions are going to take place; and the standards of

performance that is expected by the applicant.

11.3.3.5 Organization of Instruction, Using Lesson Plans and Other Training Guides

As described in Section 11.3.5 of the SAR (AES, 2009a), lessons plans will be developed under

the guidance of the training function from the learning objectives. The lesson plans will be

reviewed and approved by the training function and the cognizant organization in the subject

matter prior to use.

11.3.3.6 Evaluation of Trainee Learning

Section 11.3.6 of the SAR (AES, 2009a) states that observation, demonstration, and oral or

-written tests will be used by the applicant to evaluate the trainee's understanding and command

of learning activities. The evaluations are going to be performed by individuals qualified in the

subject matter.

11.3.3.7 Conduct of OJT Training

As described in Section 11.3.7 of the SAR (AES, 2009a), OJT training will be used in

combination with classroom activities that are QA Level 1 and QA Level 2. OJT training will be

conducted by competent, designated personnel using the current performance-based training

materials. The applicant states that the completion of OJT training will be demonstrated by

actual task performance or performance of a simulation of the task with the trainee explaining

task actions using conditions encountered during the performance of the task-including

references, tools, and equipment reflecting the actual task to the extent practical.

11.3.3.8 Evaluation of Training Effectiveness

As described in Section 11.3.8 of the SAR (AES, 2009a), the applicant will evaluate the training

program periodically to measure its effectiveness. The evaluation will consider feedback

provided from trainees after completion of the classroom training sessions and identifies the

program's strengths and weaknesses. The evaluation, as described by the applicant, will also

determine whether the program content matches current job needs and if corrective actions are

needed to improve the program's effectiveness.

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The training function will be leading any training program evaluations and implementing the

corrective actions. Some elements that are addressed by the applicant in the training objectives

are:

* Management and administration of training and qualification programs;

* Development and qualification of the training staff;

* Position training requirements;

" Determination of training program content, including its facility change control interface with

the configuration management system;

" Design and development of training programs, including lesson plans;

" Conduct of training;

" Trainee examinations and evaluations;

* Training program assessments and evaluations.

The applicant will document the evaluation results and will highlight the program's strength and

weaknesses. The applicant will review identified weaknesses; recommend improvements; and

make any changes to the affected procedures, practices, or training materials.

The facility system and training program will be audited by the QA Department. Trainees and

vendors may provide inputs related to the facility's training program.

11.3.3.9 Personnel Qualification

Section 11.3.9 of the SAR (AES, 2009a) describes the personnel qualification as qualification

and training requirements that will be established and implemented for process operator

candidates in plant procedures. Qualification requirements for key management positions are

described by the applicant in Chapter 2, "Organization and Administration" of the SAR. The

training and qualification requirements for QA personnel are provided in the QAPD for design,

construction, operation, and decommissioning of the EREF.

11.3.3.10 Periodic Personnel Evaluations

As described in Section 11.3.10 of the SAR (AES, 2009a), personnel who perform activities

relied on for safety will be evaluated at least biennially to determine whether they are capable of

continuing to perform these activities. Personnel will be evaluated by written test, oral test, or

on-the-job performance evaluation. The applicant will provide retraining or other appropriate

action when results of the evaluations dictate the need. Retraining will be provided to personnel

if any new or revised information results in plant modifications, procedure changes, and QA

program changes.

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11.3.3.11 Evaluation Findings

As described in the preceding sections of this SER, the staff reviewed the applicant's training

and qualifications program following the guidance in Section 11.4.3.3 of NUREG-1 520 (NRC,

2002).

The applicant described in Section 11.3 of the SAR (AES, 2009a) the training programs for the

operational phase of the facility, including the preoperational functional testing and initial startup

testing.

In Section 11.3.1 of the SAR (AES, 2009a), the applicant described the organization and

management of the training function. The applicant also described: (1) the line management

responsibilities for the content and effective conduct of the training, (2) the job function,

responsibility, authority, and accountability of personnel involved in managing, supervising, and

implementing training; (3) the performance-based training as the primary management tool for

analyzing, designing, developing, conducting, and evaluating training and (4) maintenance of

programmatic and individual training records that support management information needs and

provide required data on each individual's training and qualification. In addition the applicant

provided an adequate method to ensure that: (1) documentation and implementation of

procedures to provide reasonable assurance that all phases of training are conducted reliably

and consistently; (2) training documents are linked to the CM system to provide reasonable

assurance that the training reflects design changes and modifications; and (3) exemptions are

from training to trainees and incumbents are will be granted only when justified, documented,

and approved by management.

The applicant described in Section 11.3.2 of the SAR (AES, 2009a) the analysis and

identification of functional areas required training to ensure that formal training is provided for

each position or activity that is relied on for safety either in a classroom or OTJ or both. Section

11.3.3 of the SAR (AES, 2009a) described the different parts of the training program -General

Employee Training, Technical Training and Employee Development/Management Supervisory

Training - that will be conducted and which personnel will be required to complete them. The

applicant also described how each activity selected for training (initial or continuing) from the

facility-specific activities is correlated with supporting procedures and training materials.

In Section 11.3.3 of the SAR (AES, 2009a), the applicant described the position training

requirements as appropriate for facility personnel and other staff who perform regulated

activities commensurate with the assigned functional responsibility and authority of the

respective personnel.

Section 11.3.4 of the SAR (AES, 2009a) described the training basis and objectives which

includes knowledge, skills, and abilities that the trainee should acquire; the conditions under

which required actions will take place; and the standards of performance the trainee should

achieve upon completion of the training activity.

The applicant described in Section 11.3.5 of the SAR (AES, 2009a) the lesson plans and other

training guides that will ensure that guidance is provided consistent to the conduct of training

activities and based on required learning objectives derived from specific job performance

requirements. The applicant will have requirements in place for the review and approval of all

lesson plans or guides and other training materials before their issue and use.

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The applicant described in Section 11.3.6 of the SAR (AES, 2009a) the evaluation of trainee

accomplishment which will evaluate the trainee understanding through

observation/demonstration or oral or written test as appropriate to measure the training skills

and job performance.

The applicant described in Section 11.3.7 of the SAR (AES, 2009a) the OTJ training element

that will be used for some training activities of activities relied on for safety. The applicant

described OTJ training as a well-organized training activity that uses current training materials

and is conducted by designated personnel who are competent in the program standards and

training methods. Completion of OTJ training will be by actual task performance or "walked

down," the conditions of task performance, references, tools, and equipment reflecting the

actual task to the extent possible, when the actual task cannot be performed.

The applicant described in Section 11.3.8 of the SAR (AES, 2009a) the process by which the

training effectiveness is evaluated and its relation to job performance. The process described

by the applicant will ensure that its activities convey all required skills and knowledge and are

used to revise the training, where necessary, based on the performance of trained personnel in

the job setting. In addition, the applicant will (1) periodically conduct a comprehensive

evaluation of individual training to identify strengths and weaknesses, (2) use feedback from

trainee performance during training and from former trainees and their supervisors to evaluate

and refine the training, and (3) initiate, evaluate, track, and incorporate improvements and

changes to initial and continuing training to correct training deficiencies and performance

problems.

The applicant described in Section 11.3.10 of the SAR (AES, 2009a) the periodic personnel

evaluations that provide for continuing assurance of personnel training and qualification. The

applicant provided a method that addresses periodic requalification of personnel by training or

testing or both, as necessary, at least biennially, to provide reasonable assurance that

personnel continue to understand, recognize the importance of, and be qualified to perform

activities that are relied on for safety.

Based on the above, the staff found that the training and qualifications program description

provided by the applicant in Section 11.3 and its subsections is consistent with the guidance

provided in Section 11.4.3.3 of NUREG-1 520 (NRC, 2002) and is, therefore, acceptable.

11.3.4 Procedures Development and Implementation

The NRC staff has reviewed the applicant's description of its procedures development and

implementation following the regulatory acceptance criteria in Section 11.4.3.4 of NUREG-1 520

(NRC, 2002).

As described in Section 11.4 of the SAR (AES, 2009a), the applicant will conduct activities

involving licensed materials or QA Level 1 and QA Level 2 items and activities in accordance

with approved procedures. Procedures will be made available to the NRC for inspection prior to

the initial enrichment activities at the facility.

The applicant describes the four types of plant procedures that will be used to control activities:

(1) operating procedures, (2) administrative procedures, (3) maintenance procedures, and

(4) emergency procedures.

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Operating procedures, as described by the applicant, will be developed for workstation and

Control Room operators. The applicant will implement a methodology for identifying,

developing, approving, implementing, and controlling operating procedures. The operating

procedure will be used to directly control process operations. The operating procedures will

include, among other things, the: (a) purpose of the activity; (b) governing regulations, polices,

and guidelines; (c) type of procedure; (d) steps for each operating process phase; (e) hazards

and safety considerations; (f) operating limits; (g) measures to prevent exposure; (h) associated

IROFS and their functions, and (i) the timeframe for which the procedure is valid. Applicable

safety limits and IROFS will be clearly identified in the procedures.

The applicant will consider the ISA results when identifying needed procedures. The method

will ensure, as a minimum, that: (a) operating limits and IROFS are specified in the procedure;

(b) procedures include required actions for off-normal conditions of operation; as well as normal

operations; (c) if needed, safety checkpoints are identified at appropriate steps in the procedure;

(d) procedures are validated through field tests; (e) procedures are approved by management

personnel responsible and accountable for the operation; (f) a mechanism is specified for

revising and reissuing procedures in a controlled manner; (g) the QA elements and CM Program

at the facility provide reasonable assurance that current procedures are available and used at

all work locations; and (h) the facility's training program trains the required persons in the use of

the latest procedures available (AES, 2009a).

The applicant will use administrative procedures to perform activities that support the process

operations. The activities include management measures such as: (a) configuration

management; (b) nuclear criticality, radiation, chemical, and fire safety; (c) QA; (d) design

control; (e) plant personnel training and qualification; (f) audits and assessments; (g) incident

investigations; (h) recordkeeping and document control; (i) reporting; and (j) procurement.

Administrative procedures will also be used for the implementation of the emergency and

security plans including the Emergency Plan, the Physical Security Plan, the Fundamental

Nuclear Material Control Plan, and the Standard Practice Procedures Plan for the Protection of

Classified Matter.

Maintenance procedures will address the preventive and corrective maintenance, surveillance

functional testing, and the requirements for pre-maintenance activity involving reviews of the

work to be performed and reviews of procedures of QA Level 1 and QA Level 2 activities or

items. The emergency procedures will address the pre-planned actions of operators and other

plant personnel in the event of an emergency.

The applicant will establish and implement procedures for the NCS in accordance with

ANSI/ANS-8.19-1996 (ANSI, 1996). The applicant states that NCS procedures will be written

such that no single, inadvertent departure from a procedure could cause an inadvertent

criticality. The applicant will establish NCS postings identifying the administrative controls

applicable to the activity or area in question.

The applicant will conduct periodic reviews of procedures to assure their continued accuracy

and usefulness. The applicant, at a minimum, will review all operating procedures every five

years and emergency procedures every year. In addition, the applicant will review applicable

procedures after unusual incidents, such as an accident, unexpected transient, significant

operator error, equipment malfunction, or after any modification to a system. Procedures will be

revised as needed.

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11.3.4.1 Preparation of Procedures

As described in Section 11.4.1 of the SAR (AES, 2009a), the procedures will be assigned to a

member of the facility staff or contractor for development. The applicant will verify the technical

accuracy of procedures by interdisciplinary reviews. The applicant will perform a walkdown of

the procedure in the field or a tabletop walkthrough to verify that the procedure can be

performed as written. The procedures for the operation of QA Level 1 and QA Level 2 items will

be subjected to an independent review that will be performed by personnel not having direct

responsibility for the work function under review. The applicant states that the designated

approver will determine if any additional cross-disciplinary review is required. The Plant

Manager or designee will have the responsibility of approving all procedures. However, any

procedures that involve QA directly will require the QA Manager's approval.

11.3.4.2 Administrative Procedures

The applicant describes in Section 11.4.2 of the SAR (AES, 2009a) administrative procedures

as those written by each department as necessary to control activities that support process

operations, including management measures. The applicant will write administrative

procedures for several areas, such as:

A. The operator's authority and responsibility,

B. Activities affecting facility operation or operating indications,

C. Manipulation of facility control,

D. Relief of duties,

E. Equipment control,

F. Master surveillance testing schedule,

G. A Control Room Operations Logbook, and

H. Fire Protection Procedures.

The applicant describes the process for administrative control of maintenance. The applicant

will establish a maintenance program for QA Level 1 and QA Level 2 items. The personnel

performing maintenance activities will be qualified in accordance with applicable codes and

standards and procedures. The work will be conducted in accordance with written procedures

that conform to applicable codes, standards, specifications, and other appropriate criteria. The

applicant will schedule maintenance appropriately and without jeopardizing the facility operation

or the safety of facility personnel. The applicant will maintain maintenance records for

QA Level 1 and QA Level 2 items.

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11.3.4.3 Procedures

The applicant states in Section 11.4.3 of the SAR (AES, 2009a) that activities involving licensed

materials or QA Level 1 and QA Level 2 items and activities will be conducted in accordance

with approved procedures. These procedures' intent is to provide a pre-planned method of

conducting operations of systems to eliminate errors due to on-the-spot analysis and judgments.

Procedures, as described by the applicant, will be sufficiently detailed so that qualified

individuals can perform the required functions without direct supervision and will contain a

degree of flexibility appropriate to the activities being performed (AES, 2009a). In addition, the

applicant has procedural guidance to identify the manner in which procedures are to be

implemented.

The applicant will write plant-specific procedures for abnormal events for the facility. These

procedures will be based on a sequence of observations and actions, with emphasis placed on

operator responses to indications in the Control Room. The actions outlined by the applicant in

abnormal event procedures will be based on a conservative course of action to be followed by

the operating crew.

The applicant will issued temporary changes to procedures for operating activities that are of a

non-recurring nature. The applicant will use the temporary changes to procedures when

revision of an operating or other permanent procedure is not practical. The applicant states that

temporary changes to procedures shall not involve a change to the ISA and shall not alter the

intent of the original procedure. The temporary changes to procedures will be approved by two

members of the facility management staff, at least one of whom is a Production Manager. The

applicant will document, review and approve temporary changes to procedures in accordance

with the process described in Section 11.4.4 of the SAR, "Changes to Procedures," and will

identify the approved duration for the temporary change in the temporary procedure change in

accordance with 10 CFR 70.72(a)(5).

The applicant will have written procedures, documented instructions, checklists, or drawings for

the maintenance of the facility's SSCs-including testing and calibration. The procedures will

be written in accordance to the appropriate circumstances and will conform to applicable codes,

standards, specifications, and other appropriate criteria. The Maintenance Department, under

the Maintenance Manager, will be responsible for the preparation and implementation of

maintenance procedures. In addition, the Maintenance Manager will have the overall

responsibility for assuring that the periodic testing is in compliance with the requirements.

The applicant will provide for compensatory measures to be put into place when testing IROFS

that are not redundant to ensure that the IROFS will perform until it is put back into service. The

periodic test procedures will be performed by the Operations and Maintenance departments.

Chemical and radiochemical activities associated with facility IROFS will be performed in

accordance with approved, written procedures. The Radiation Protection/Chemistry Manager

will have the responsibility for preparation and implementation of chemistry procedures. The

radioactive waste management activities associated with the facility's liquid, gaseous, and solid

waste systems will also be performed in accordance with approved, written procedures. In this

case, the facility's operations and radiation protection/chemistry departments will have the

responsibility for preparation and implementation of the radioactive waste management

procedures.

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Procedures, as stated by the applicant, will include provisions for operations to stop and place

the process in a safe condition if a step of a procedure cannot be performed as written.

11.3.4.4 Changes to Procedures

As described in Section 11.4.4 of the SAR (AES, 2009a), the process for the changes to

procedures will have the preparer of the change document the change and the reason for the

change. The applicant will perform an evaluation in accordance with 10 CFR 70.72, as

appropriate, to determine if a change to the license is needed to implement the proposed

changes. If a change to the license is needed, the change is not going to be implemented until

prior approval is received from the NRC. The procedure will then be reviewed by a qualified

reviewer. The Plant Manager, a functional area 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: and by which department(s). The applicant will perform

interdisciplinary reviews, as a minimum, for changes involving chemical safety, radiation safety,

criticality safety, and changes involving nuclear material control and accounting. The applicant

will maintain the records for cross-functional reviews for all changes to procedures involving

licensed materials or QA Level 1 and QA Level 2 items and activities in accordance with

Section 11.7 of the SAR, Records Management.

11.3.4.5 Distribution of Procedures

As described in Section 11.4.5 of the SAR (AES, 2009a), the applicant will distribute originally

issued, approved procedures and approved procedure revisions in a controlled manner by

Document Control. Document Control, as stated by the applicant, will establish and maintain an

index of the distribution of copies of facility procedures. Revisions will be controlled and

distributed in accordance with this index. Indexes will be reviewed and updated on a periodic

basis or as required. The Department Managers or their designees will be responsible for

ensuring that personnel doing work which require the use of the procedures have ready access

to controlled copies of the procedures.

11.3.4.6 Evaluation Findings

As described in the preceding sections of this SER, the staff reviewed the applicant's

procedures development and implementation strategy following the guidance in

Section 11.4.3.4 of NUREG-1520 (NRC, 2002).

The applicant provided in Section 11.4 of the SAR (AES, 2009a) information regarding the

procedure categories used at the facility including operating, administrative, maintenance, and

emergency procedures for activities involving licensed materials or QA Level 1 and QA Level 2

items and activities. The applicant provided a listing of the types of activities that are covered,

or are planned to be covered, by written procedures including topics of administrative

procedures; system procedures that address startup, operation, and shutdown; abnormal

operation or alarm response; maintenance activities that address system repair, calibration,

inspection, and testing; and emergency procedures. The applicant also described methods for

identifying, developing, approving, implementing, and controlling operating procedures. This

method will consider the ISA in identifying needed procedures; it requires the procedures to

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(1) specify operating limits and IROFS, (2) include required actions for off-normal conditions of

operation, as well as normal operations, and (3) identify safety checkpoints, as required; and

use field tests to validate procedures. In addition, the method will also ensure that (1)

management personnel who are responsible and accountable for the operation will approve the

procedures; (2) mechanism for revising and reissuing procedures in a controlled manner is

specified; (3) QA elements and CM functions at the facility provide reasonable assurance that

current procedures are available and used at all work locations; and (4) training program

instructs the required personnel in the use of the latest procedures.

In addition, in Section 11.4 of the SAR (AES, 2009a) the applicant (1) included provisions for

periodic reviews of the procedures to ensure their continued accuracy and usefulness and

established the timeframe for reviews of the various types of procedures; (2) described the use

and control of procedures; and (3) included provisions for the review of procedures after

unusual incidents. The applicant provided in Section 11.4.1 of the SAR (AES, 2009a) the

method for preparation of procedures. The applicant described the verification of procedures

and those steps that ensure that the procedures are technically accurate and can be performed

as written.

The applicant described in Section 11.4.3 of the SAR (AES, 2009a) the requirements governing

the use of temporary procedures. The requirements include provisions for the issuance of

temporary procedures only when permanent procedures do not exist to (1) direct operations

during testing, maintenance, and modifications, (2) provide guidance in unusual situations not

within the scope of permanent procedures, and (3) provide assurance of orderly and uniform

operations for short periods when the facility, system, or component is performing in a manner

not covered by permanent procedures. The applicant also described the approval and

timeframe for use of the temporary procedure. Section 11.4.3 included provisions that allow for

operations to stop and place the process in a safe condition if a step of a procedure cannot be

performed as written.

Based on the above, the staff found that the description provided by the applicant for the

procedures that will be used at the facility in Section 11.4 and its subsections is consistent with

the guidance provided in Section 11.4.3.4 of NUREG-1 520 (NRC, 2002) and is, therefore,

acceptable.

11.3.5 Audits and Assessments

The NRC staff has reviewed the applicant's program for audits and assessments following the

regulatory acceptance criteria in Section 11.4.3.5 of NUREG-1 520 (NRC, 2002).

Section 11.5 of the SAR (AES, 2009a) describes audits and assessments and states that audits

will be focused on verifying compliance with regulatory and procedural requirements, and

licensing commitments. Furthermore, the applicant states that assessments will be focused on

effectiveness of activities and ensuring that QA Level 1 and QA Level 2 items are reliable and

are available to perform their intended safety functions. The applicant will perform audits and

assessments on critical work activities associated with facility safety, environmental protection

and other areas that they identify via trends. The QA Department will be responsible for audits

and that these are conducted in accordance with a written plan that identifies and schedules

audits to be performed. The applicant provides a description of the audit team. The audit team

members will not have direct responsibility for the function and area being audited, and will have

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technical expertise or experience in the area being audited. In addition, team members will be

indoctrinated in audit techniques. Audits will be conducted on an annual basis.

Assessments will be divided by the applicant into two categories: management assessments

that will be conducted by the line organizations responsible for the work activity and

independent assessments will be conducted by individuals not involved in the area being

assessed. Audits of work activities associated with QA Level 1 and QA Level 2 items and

activities will be the responsibility of the QA Department.

Audits and assessments will be performed to assure that facility activities are conducted in

accordance with the written procedures and that the processes reviewed are effective. The

applicant will conduct audits and assessments in the areas of: radiation safety, nuclear

criticality safety, chemical safety, industrial safety including fire protection, environmental

protection, emergency management, QA, CM, maintenance, training and qualification,

procedures, CAP/Incident investigation and records management.

The applicant will perform audits and assessments routinely by qualified staff personnel that are

not directly responsible for production activities. Any deficiencies identified during the audit or

assessment requiring corrective action will be forwarded to the responsible manager of the

applicable area or function for action in accordance with the CAP procedure. In addition, future

audits and assessments will include a review to evaluate if corrective actions regarding any

deficiencies found have been effective.

As discussed in the applicant's September 30, 2009 response to the staffs Request for

Additional Information (RAI) (NRC, 2009), the activities of the QA Program which the QA

Organization has direct responsibility for implementing will be performed by an independent

third-party, Certified Lead Auditor that will be contracted or obtained from affiliate companies to

an appropriate level of independence from AES (AES, 2009b). In addition, the QA Organization

performs self-assessments of their areas of responsibility.

The results of the audits will be provided in a written report to the AES President, Plant

Manager, the Safety Review Committee (SRC), and the Managers responsible for the activities

audited. The applicant will maintain records of the instructions and procedures, persons

conducting the audits or assessments, and any identified violations of license conditions and

corrective actions taken.

11.3.5.1 Scheduling of Audits and Assessments

As described in Section 11.5.2 of the SAR (AES, 2009a), a schedule will be established

i dentifying the audits and assessments to be performed and the responsible organization

assigned to conduct the activity. The frequency of audits and assessments will be based on the

status and safety importance of the activities being performed and upon work history. The

system of audits and assessments will be designed by the applicant to ensure program

oversight every three years. The schedule will be reviewed periodically and will be revised as

necessary to ensure coverage commensurate with current and planned activities. NCS audits

will be conducted and documented quarterly such that all aspects of the Nuclear Criticality

Safety Program will be audited at least every two years. The Operations Department will be

assessed to ensure that NCS procedures are being followed and that the process conditions

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have not been altered to adversely affect NCS. Assessments will be conducted at least semiannually. In addition, the applicant will conduct and document weekly NCS walkthroughs of UF6

process areas.

11.3.5.2 Procedures for Audits and Assessments

Section 11.5.3 of the SAR (AES, 2009a) states that internal and external audits and

assessments will be conducted using approved procedures that meet the QA Program

requirements. The procedures will provide requirements for audit and assessment activities,

such as scheduling and planning of the audit and assessment; certification requirements of

audit personnel; development of audit plans and audit and assessment checklists as applicable;

performance of the audit and assessment; reporting and tracking of findings to closure; and

closure of the audit and assessment. The applicable procedures will emphasize reporting and

correction of findings to prevent recurrence.

The applicant will conduct audits and assessments by using approved audit and assessment

checklists, interviewing responsible personnel, performing plant area walkdowns, reviewing

controlling plans and procedures, observing work in progress, and reviewing completed QA

documentation. The audit and assessment results data will be tracked in the CAP and

periodically analyzed for potential trends.

The audits and assessments team leader will be required to develop the audit and/or

assessment report to management with documented verification of performance against

established performance criteria for QA Level 1 and QA Level 2 items and activities

documenting the findings, observations, and recommendations for program improvement. The

applicant will require that responsible managers review the reports and provide any required

responses to reported findings. Audit reports will require an effectiveness evaluation and

statement for each of the applicable QA program elements reviewed during the audit. The

audit/assessment will be considered closed with the proper documentation as required by the

applicable audit and assessment procedure.

11.3.5.3 Qualification and Responsibilities for Audits and Assessments

As described in Section 11.5.4 of the SAR (AES, 2009a), the QA Manager will initiate the audits;

and in coordination with the responsible Lead Auditor, they will determine the scope of each

audit. The QA Manager may initiate special audits or expand the scope of audits.

The Lead Auditor, as described by the applicant, will direct the audit team in developing

checklists, instructions, or plans and how to perform the audit. Audits will be conducted in

accordance with the checklists, but the scope may be expanded by the audit team during the

audit. The audit team will consist of one or more auditors that are responsible for performing

audits in accordance with the applicable QA procedures. The auditors and lead auditors will

have to hold certifications as required by the QA Program. Prior to certification under the AES's

QA Program, auditors will have complete training on: AES's QA Program; audit fundamentals,

including audit scheduling, planning, performance, reporting, and follow-up action involved in

conducting audits; objectives and techniques of performing audits; and OJT training. The

certification of auditors and lead auditors will be based on the QA Manager's evaluation of

education, experience, professional qualifications, leadership, sound judgment, maturity,

analytical ability, tenacity, and past performance and completion of QA training courses.

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Lead auditors will have to participate in a minimum of five QA audits, or audit equivalent within a

period of time, not to exceed three years prior to the date of certification. Audit equivalents

include assessments, pre-award evaluations or comprehensive surveillances (provided the

prospective Lead Auditor took part in the planning, checklist development, performance, and

reporting of the audit equivalent activities). One of the audits must be a nuclear-related QA

audit or audit equivalent within the year prior to certification.

Personnel performing assessments will not require certification; but they will be required to

complete QA orientation training, as well as training on the assessment process. The NCS

assessments will be performed under the direction of the criticality safety staff.

11.3.5.4 Evaluation Findings

As described in the preceding sections of this SER, the staff reviewed the applicant's audits and

assessments program following the guidance in Section 11.4.3.5 of NUREG-1520 (NRC, 2002).

In Section 11.5 of the SAR (AES, 2009a) the applicant described the program directives

covering the audit and assessment function including the activities to be audited, audit

frequency, guidance in conducting the audit or assessment, assigned responsibilities for each

phase of the work, and procedures for recording the results and recommending actions to be

taken. It provided for the conduct of internal audits and independent assessments of activities

significant to facility safety and environmental protection. The applicant described the audits as

those performed to verify that operations are being conducted in accordance with regulatory

requirements and license commitments and independent assessments as those conducted by

offsite groups or individuals not involved in the licensed activity to verify that the health, safety,

and environmental compliance functions are effectively achieving their designed purposes.

In Section 11.5.1 the applicant provided a list of the areas where audits and assessments will be

conducted. The areas listed include: radiation safety, nuclear criticality safety, chemical safety,

fire safety, environmental protection, emergency management, QA, CM, maintenance, training

and qualification, procedures, incident investigation, and records management.

The applicant also included provisions for qualified personnel without direct responsibility for the

function and area being audited or assessed to perform the audits and assessments. The

subsections on Section 11.5 included specifications with regards to staff positions and

committees responsible for audits and assessments and describe the levels of management to

which results are reported.

Based on the above, the staff found that the audits and assessment program description

provided by the applicant in Section 11.5 and its subsections is consistent with the guidance

provided in Section 11.4.3.5 of NUREG-1520 (NRC, 2002) and is, therefore, acceptable.

11.3.6 Incident Investigations and Corrective Action Process

The NRC staff has reviewed the applicant's process for conducting incident investigations and

its corrective action process following the regulatory acceptance criteria in Section 11.4.3.6 of

NUREG-1520 (NRC, 2002).

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11.3.6.1 Incident Investigations

The applicant describes in Section 11.6.1 of the SAR (AES, 2009a) the incident investigation

process as a simple mechanism available for use by any person at the facility to report

deficiencies, abnormal events, and potentially unsafe conditions or activities. Any abnormal

events that potentially threaten or lessen the effectiveness of health, safety or environmental

protection will be identified, and reported to, and investigated by the EHS&L Manager. The

applicant will consider each event in terms of its requirements for reporting in accordance with

regulations and will evaluate the event to determine the level of investigation required. The

applicant's process of incident identification, investigation, root cause analysis, environmental

protection analysis, recording, reporting, and follow-up will be addressed in and performed by

written CAP procedures. CAP procedures will include guidance for classifying occurrences,

including examples of threshold off-normal occurrences. The depth of the investigation will

depend, as explained by the applicant, upon the severity of the classified incident in terms of the

levels of uranium released and/or the degree of potential for exposure of workers, the public, or

the environment.

The EHS&L Manager will be responsible for maintaining a list of agencies to be notified in case

of the event; determining if a report to any agencies is required and notifying the appropriate

agencies when required. In addition, the EHS&L Manager or designee will maintain a record

and track to completion corrective actions to be implemented as a result of off-normal

occurrence investigations in accordance with CAP procedures. The licensing organization will

be responsible for all appropriate communications with government agencies.

The incident investigation process, as described by the applicant, will establish a process to

investigate abnormal events that may occur during operation of the facility to determine their

specific or generic root cause(s) and generic implications; to recommend corrective actions; and

to report to the NRC as required by 10 CFR 70.50 (CFR, 2008d) and 70.74 (CFR, 2008e). The

investigation process will include a prompt, risk-based evaluation and, depending on the

complexity and severity of the event, one individual may suffice to conduct the evaluation. The

investigator(s) will be independent from the line function(s) involved with the incident under

investigation-and are assured of no retaliation for participating in investigations. Investigations

will begin within 48 hours of the abnormal event, or sooner, depending on safety significance of

the event. The record of IROFS failures required by 10 CFR 70.62(a)(3) (CFR, 2008f) will be

reviewed as part of the investigation. Record revisions necessitated by post-failure investigation

conclusions will be made within five working days of the completion of the investigation.

Qualified internal or external investigators will be appointed to serve on investigating teams

when required. The teams will include at least one process expert and at least one team

member trained in root cause analysis.

The applicant will monitor and document corrective actions through completion. The applicant

will maintain auditable records and documentation related to abnormal events, investigations,

and root cause analyses so that "lessons learned" may be applied to future operations of the

facility. For each abnormal event, the incident report will include a description, contributing

factors, a root cause analysis, findings, and recommendations. Relevant findings will be

reviewed with all affected personnel. Details of the event sequence will be compared with

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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 CAP procedures for conducting an incident investigation, and the

procedures will contain elements such as: (a) a documented plan for investigating an abnormal

event; (b) a description of the functions, qualifications, and/or responsibilities of the manager

who would lead the investigative team and those of the other team members; the scope of the

team's authority and responsibilities; and assurance of cooperation of management;

(c) assurance of the team's authority to obtain all the information considered necessary and its

independence from responsibility for or to the functional area involved in the incident under

investigation; (d) retention of documentation relating to abnormal events for two years or for the

life of the operation, whichever is longer; (e) guidance for personnel conducting the investigation

on how to apply a reasonable systematic, structured approach to determine the specific or

generic root cause(s) and generic implications of the problem; (f) requirements to make

available original investigation reports to the NRC on request; and (g) a system for monitoring

the completion of appropriate corrective actions.

11.3.6.2 Corrective Action Process

Section 11.6.2 of the SAR (AES, 2009a) states that AES QA Program will identify the

responsibilities and provide authority for those individuals involved in quality activities to identify

any condition adverse to quality, such as failures, malfunctions, deficiencies, deviations,

defective materials and equipment, and non-conformances. These individuals will identify and

document conditions adverse to quality, analyze and determine how the conditions can be

corrected or resolved, and take such steps as necessary to implement corrective actions in

accordance with documented procedures.

The QA Program will require regularly scheduled audits and assessments to ensure that

needed corrective actions are identified and that employees have the authority and

responsibility to initiate the corrective action process if they discover deficiencies. The QA

Program will contain procedures for identifying, reporting, resolving, documenting, and

analyzing conditions adverse to quality. The reports of conditions adverse to quality will be

analyzed to identify trends in quality performance. Significant conditions adverse to quality and

significant trends will be reported to senior management in accordance with CAP procedures,

and follow-up actions will be taken by the QA Manager to verify proper and timely

implementation of the corrective actions.

11.3.6.3 Evaluation Findings

As described in the preceding sections of this SER, the staff reviewed the applicant's incident

investigations and corrective action processes following the guidance in Section 11.4.3.6 of

NUREG-1520 (NRC, 2002).

In Section 11.6.1 of the SAR (AES, 2009a) the applicant described the incident investigation

program as a mechanism available for use by any person to report abnormal events that may

occur during operation of the facility to be investigated to determine their specific or generic root

cause(s), generic implications, and risk significance; to recommend corrective actions; and to

report to the NRC as required by 10 CFR 70.50, "Reporting Requirements," and 10 CFR 70.74,

"Additional Reporting Requirements." In addition, the applicant include provision within this

section to monitor and document corrective actions through completion and ensure that

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corrective actions are taken within a reasonable period to resolve findings from abnormal event

investigations. The applicant also provided for the maintenance of documentation related to

abnormal events for the life of the operation so that "lessons learned" may be applied to future

operations of the facility. The applicant specified that 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.

Based on the above, the staff found that the incident investigation program description provided

by the applicant in Section 11.6 and its subsections is consistent with the guidance provided in

Section 11.4.3.6 of NUREG-1520 (NRC, 2002) and is, therefore, acceptable.

11.3.7 Records Management

11.3.7.1 Records Management program

The NRC staff has reviewed the applicant's records management program following the

regulatory acceptance criteria in Section 11.4.3.7 of NUREG-1 520 (NRC, 2002).

As described in Section 11.7 of the SAR (AES, 2009a), records management will be performed

in a controlled and systematic manner to provide identifiable and retrievable documentation.

The applicant will identify the QA Records to be generated, supplied or held in applicable design

specifications, procurement documents, or other documents in accordance with approved

procedures. The applicant will consider QA records valid after they are stamped, initialed,

signed, or otherwise authenticated and dated by authorized personnel. For computer codes

and computerized data used for activities relied on for safety, the applicant will establish

procedures for maintaining readability and usability of older codes and data as computing

technology changes.

The AES QA Program will require procedures for reviewing, approving, handling, identifying,

retention, retrieval and maintenance of QA records. The applicant includes records such as

results of tests and inspections required by applicable codes and standards, construction,

procurement, and receiving records; personnel certification records; design calculation;

purchase orders; specifications and amendments; procedures; incident investigation results and

approvals or corrective action taken; various certification forms; source surveillance and audit

reports; component data packages; and any other QA documentation required by specifications

or procedures. The applicant will maintain records at locations where they can be reviewed and

audited to establish that the required quality has been assured.

The applicant will maintain a Master File within the facility. The applicant will control the access

and use of the Master File. Documents in the Master File will be legible and identifiable as to

the subject to which they pertain. The applicant states that documents in the Master File may

be originals or reproduced copies. The applicant may use computer storage of data in the

Master File. The Master File as described by the applicant will provide for accurate retrieval of

information without undue delays and will preclude deterioration of records. The applicant will

provide written instructions for the storage of records in the Master File and will designate a

supervisor to be responsible of the implementation of these requirements. The applicant will

include in the instructions: (a) a description of the location(s) of the Master File and an

identification of the location(s) of the various record types within the Master File; (b) the filing

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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. This is not required for documents

generated within a section for use and storage in the same sections' satellite files; (d) a method

for maintaining a record of the records 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.

If the applicant uses a single records storage facility, the applicant will review the storage facility

for adequacy of protecting the records by a person competent in the technical field of fire

protection and fire extinguishing. The applicant states that dual records storage facilities will not

be subject to the same review. Records related to health and safety will be maintained in

accordance with the applicable requirements of Title 10, Code of Federal Regulations. The

applicant will retain the records in accordance with the Records Management procedures.

The applicant provides a list of examples of records that will be retained.

11.3.7.2 Evaluation Findings

As described in the preceding sections of this SER, the staff reviewed the applicant's records

management program following the guidance in Section 11.4.3.7 of NUREG-1 520 (NRC, 2002).

In Section 11.7 of the SAR (AES, 2009a) the applicant described the records management

program which prepares, verifies, characterizes, and maintains records and ensures that

records are legible, identifiable, and retrievable for their designated lifetimes. The applicant

included provisions to (1) categorize records by relative safety importance to identify record

protection and storage needs and to designate the retention period for individual kinds of

records; (2) protect records against tampering, theft, loss, unauthorized access, damage, or

deterioration while in storage; (3) establish and document procedures specifying the

requirements and responsibilities for record selection, verification, protection, transmittal,

distribution, retention, maintenance, and disposition; (4) implement procedures that assign

responsibilities for records management, specify the authority needed for records retention or

disposal, specify which records must have controlled access and provide the controls needed,

provide for the protection of records from loss, damage, tampering, and theft or during an

emergency, and specify procedures for ensuring that the records management system remains

effective; and (5) have procedures in place to promptly detect and correct any deficiencies in the

records management system or its implementation.

For computer codes and computerized data used for activities relied on for safety, as specified

in the ISA Summary, the applicant described that procedure(s) will be established for

maintaining readability and usability of older codes and data as computing technology changes.

Based on the above, the staff found that the records management program description provided

by the applicant in Section 11.7 is consistent with the guidance provided in Section 11.4.3.7 of

NUREG-1 520 (NRC, 2002) and is, therefore, acceptable.

11-34
11.3.8 Other QA Elements

11.3.8.1 Description of other QA Elements

The applicant described other QA elements in Section 11.8 of the SAR (AES, 2009a). The

applicant's QA Program and its supporting manuals, procedures and instructions are applicable

to items and activities designated as QA Level 1 and QA Level 2. The QA Program specifies

mandatory requirements for performing activities affecting quality and is set forth in procedures

and its revisions which are distributed on a controlled basis to organizations and individuals

responsible for quality.

Prior to undertaking an activity, the applicant will document, approve and implement the

applicable portions of the QA Program. The applicant states that a management assessment of

the QA program is performed at least six months prior to scheduled receipt of licensed material

on the site. Prior to this initial management assessment, it is monitored through project review

meetings and annual assessments. Any items identified by the applicant as needing completion

or modification will be entered into the CAP and corrective action completed before scheduled

receipt of licensed material.

While construction activities are in progress, the applicant will simultaneously use the QA

Program for design, construction, and pre-operational testing-and the QA Program for

operations. The applicant will plan and schedule system turnover as construction is completed.

Prior to system turnover, written procedures are developed for control of the transfer of SSCs

and associated documentation. The procedures include checklists, marked drawings,

documentation lists, system status, and receipt control.

The applicant states that anyone may propose changes to the QA Program supporting manuals

and procedures. The changes will be implemented after the QA Manager reviews them and

found acceptable and compatible with applicable requirements, guidelines and AES policy. The

QA Program and supporting manuals and procedures will be reviewed periodically to ensure

compliance with applicable regulations, codes, and standards.

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 will establish formal training programs for QA policies, requirements, procedures,

and methods. The applicant will provide ongoing training to ensure continuing proficiency as

procedural requirements change. New employees will be required to attend a QA indoctrination

class on authority, organization, policies, manuals, and procedures. The applicant will provide

additional formal training in specific topics such as NRC regulations and guidance, procedures,

auditing, and applicable codes and standards. The applicant will perform supplemental training

as required. OJT training will be performed by the employee's supervisor in QA area-specific

procedures and requirements. Training records will be maintained by the applicant for each

person performing quality-related job functions.

Any activities that will be contracted will be identified and controlled by the applicant. The

principal contractors will be required to comply with the portions of QA Program applicable to

the scope of their work and their performance of contracted activities will be formally evaluated

by the applicant commensurate with the importance of the activities to safety.

11-35
11.3.8.2 Evaluation Findings

In a letter dated October 30, 2009, AES submitted a request to the NRC for the expedited

review and approval of the QAPD for the EREF (AES, 2009c). AES requested the expedited

approval in order to be able to apply the QAPD language during its procurement of services and

material. The staff completed a technical review of the QAPD on April 8, 2010 (NRC, 2010).

Based on the review, documented in the accompanying Staff Evaluation Report, the staff found

the program acceptable for application to the design, construction, operation, including

maintenance and modification, and decommissioning of the proposed EREF (NRC, 2010). The

Staff Evaluation Report discusses the staff review of the QAPD which was based on NUREG1520 (NRC, 2002) and documents the staff's conclusion that the QAPD adequately describes

the application of other QA elements and has adequately established other QA elements as part

of Management Measures as required by 10 CFR Part 70.62(d) (NRC, 2010).

11.4 Evaluation Findings

This section is a summary of the staffs evaluation findings concerning management measures.

11.4.1 Configuration Management

As described in Section 11.3 of this SER, the staff has reviewed the CM function for the EREF

following the criteria in Section 11.4.3.1 of the NUREG-1 520. 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. The applicant has suitably

and acceptably described its commitment to a proposed CM system, including the method for

managing changes in procedures, facilities, activities, and equipment for IROFS. Managementlevel policies and procedures, including an analysis and independent safety review of any

proposed activity involving IROFS, are described and will provide reasonable assurance that

consistency among design requirements, physical configuration, and facility documentation is

maintained as part of a new activity or change in an existing activity involving licensed material.

The management measures will include the following elements of CM:

CM Management

The applicant will develop the organizational structure, procedures, and responsibilities

necessary to effectively implement CM.

Design Requirements

The applicant's design requirements and bases will be documented and supported by

analyses. All design documentation will be maintained current.

Document Control

The applicant's documents, including drawings, will be appropriately stored and

accessible. Drawings and related documents captured by the system will include those

necessary and sufficient to adequately describe IROFS.

11-36
Changqe Control

Responsibilities and procedures will adequately describe how the applicant will achieve

and maintain strict consistency among the design requirements, the physical

configuration, and the facility documentation. Methods will be in place for suitable

analysis, review, approval, and implementation of identified changes to IROFS, including

appropriate CM controls.

Assessments

The applicant has committed to an adequate function that includes both initial and

periodic assessments. The assessments will verify and ensure the adequacy of the CM

function.

The CM program description provided by the applicant in Section 11.1 and its

subsections is consistent with the guidance provided in Section 11.4.3.1 of NUREG1520 (NRC, 2002) requiring the applicant to describe the overall CM function, how

design requirements are established and maintained, document control, change control,

and assessments of the CM function, and is therefore acceptable.

11.4.2 Maintenance

As described in Section 11.3.2 of this SER, the staff reviewed the applicant's description of its

program to maintain the availability and reliability of IROFS identified in the ISA Summary and

compared it to the guidance in Section 11.4.3.2 of NUREG-1520 (NRC, 2002). The

maintenance program will include: surveillance/monitoring, corrective maintenance, PM,

functional testing, equipment calibration, and work control for maintenance of IROFS. The

applicant's maintenance function is proactive and uses maintenance records, PM records, and

surveillance tests to analyze equipment performance and to seek the root causes of repetitive

failures.

The surveillance and monitoring, PM, and functional testing activities described in the license

application provide reasonable assurance that the IROFS identified in the ISA Summary will be

available and reliable to prevent or mitigate accident consequences.

The maintenance function will: (1) be based on approved procedures; (2) employ work control

methods that properly consider personnel safety, awareness of facility operating groups, QA,

and the rules of CM; (3) use the ISA Summary to identify IROFS that require maintenance and

determine the level of maintenance needed; (4) justify the PM intervals in terms of the

equipment reliability goals; (5) provide for training that emphasizes the importance of IROFS

identified in the ISA Summary, regulations, codes, and personnel safety; and (6) will create

documentation that includes records of all surveillance, inspections, equipment failures, repairs,

and replacements of IROFS.

The staff concludes that the applicant's description of the maintenance program meets

the requirements of 10 CFR Part 70 and provides reasonable assurance of public health

and safety and the protection of the environment.

11-37
11.4.3 Training and Qualification

As described in Section 11.3.3, the staff reviewed the applicant's strategy for development and

implementation of the training and qualification program and compared it to the acceptance

criteria guidance in Section 11.4.3.3 of NUREG-1520 (NRC, 2002). Based on this review, the

NRC staff has concluded that the applicant has adequately described and assessed its personal

training and qualification in a manner that: (1) satisfies regulatory requirements and (2) is

consistent with the guidance in Section 11.4.3.3 of NUREG-1 520 (NRC, 2002) which requires

the applicant cover its training program appropriately covered: (a) organization and

management of the training function, (b) analysis and identification of functional areas requiring

training, (c) position training requirements, (d) training basis and objectives, (e) organization of

instruction, (f) evaluation of trainee learning, (g) conduct OJT training, (h) evaluation of training

effectiveness, (i) personnel qualification, and (j) periodic personnel evaluation. There is

reasonable assurance that implementation of the described training and qualification will result

in personnel who are qualified and competent to design, construct, start up, operate, maintain,

modify, and decommission the facility safely. The staff concludes that the applicant's plan for

personnel training and qualification meets the requirements of 10 CFR Part 70

11.4.4 Procedures

As described in Section 11.3.4 of this SER, the staff evaluated the applicant's description of the

process for the development, approval, and implementation of procedures in accordance with

the criteria in Section 11.4.3.4 of NUREG-1 520 (NRC, 2002) to ensure that the applicant's

procedures will address the operation of IROFS and all management measures supporting

those IROFS and the applicant has described a process for developing, implementing, and

controlling procedures.

The staff concludes that the applicant's plan for procedures meets the requirements of 10 CFR

Part 70.

11.4.5 Audits and Assessments

As described in Section 11.3.5 of this SER, staff reviewed Section 11.5 of the SAR (AES, 2009)

and compared the applicant's plan and processes to the review acceptance criteria in

Section 11.4.3.5 of NUREG-1520 (NRC, 2002). Based on this review, the NRC staff has

concluded that the applicant has adequately described its audits and assessments process and

its policy directives, plans, and procedural requirements considering: (a) the general structure

of the audits and assessments program, (b) the activities to be audited or assessed, (c) the

scheduling of audits and assessments, (d) the procedures for audits and assessments; and

(e) the qualifications and responsibilities for audits and assessments.

The staff concludes that the applicant's plan for audits and assessments meets the

requirements of 10 CFR Part and provides reasonable assurance of protection of the

health and safety of the public, the workers, and the environment.

11-38
11.4.6 Incident Investigations

As described in Section 11.3.6 of this SER, the staff reviewed the applicant's description of the

organization responsible for: (1) performing incident investigations of abnormal events that may

occur during operation of the facility; (2) determining the root cause(s) and generic implications

of the event; and (3) recommending corrective actions for ensuring a safe facility and safe

facility operations and found that it is consistent with the acceptance criteria in Section 11.4.3.6

of NUREG-1520 (NRC, 2002).

The applicant will monitor and document corrective actions through to completion. The

applicant will also take follow-up actions to verify the proper and timely implementation

of the corrective actions.

The applicant will maintain documentation so that "lessons learned" may be applied to

future operations of the facility.

Accordingly, the staff concludes that the applicant's description of the incident investigation

process complies with applicable NRC regulations, and is adequate.

11.4.7 Records Management

As described in Section 11.3.7 of this SER, the staff has reviewed the applicant's records

management system against the acceptance criteria in Section 11.4.3.7 of NUREG-1 520

(NRC, 2002) and concluded that the system: (1) will be effective in collecting, verifying,

protecting, and storing information about the facility and its design, operations, and maintenance

and will be able to retrieve the information in readable form for the designated lifetimes of the

records; (2) will provide a records storage area(s) with the capability to protect and preserve

health and safety records that are stored there during the mandated periods, including

protection of the stored records against loss, theft, tampering, or damage during and after

emergencies; and (3) will provide reasonable assurance that any deficiencies in the records

management system or its implementation will be detected and corrected promptly.

11.4.8 Other QA Elements

In a letter dated October 30, 2009, AES submitted a request to the NRC for the expedited

review and approval of the QAPD for the EREF (AES, 2009c). AES requested the expedited

approval in order to be able to apply the QAPD language during its procurement of services and

material. The staff completed a technical review of the QAPD on April 8, 2010 (NRC, 2010).

Based on that review, documented in the accompanying Staff Evaluation Report, the staff found

the program acceptable for application to the design, construction, operation, including

maintenance and modification, and decommissioning of the proposed EREF (NRC, 2010). The

Staff Evaluation Report discusses the staff review of the QAPD which was based on NUREG1520 (NRC, 2002) and documents the staffs conclusion that the QAPD adequately describes

the application of other QA elements and has adequately established other QA elements as part

of Management Measures as required by 10 CFR Part 70.62(d) (NRC, 2010).

11-39
11.5 References

(AES, 2009a) AREVA Enrichment Services LLC, "Eagle Rock Enrichment Facility Safety

Analysis Report," 2009.

(AES, 2009b) AREVA Enrichment Services LLC, "AREVA Enrichment Facility Responses to

U.S. Nuclear Regulatory Commission's Requests for Additional Information," 2009.

(AES, 2009c) AREVA Enrichment Services LLC, "Request for Expedited Approval of Quality

Assurance Program Description," 2009.

(NRC, 2010) U.S. Nuclear Regulatory Commission, "AREVA Enrichment Services Quality

Assurance Program Description for the Eagle Rock Enrichment Facility (TAC L32738)," 2010.

(NRC, 2009) U.S. Nuclear Regulatory Commission, "Request for Additional Information -

AREVA Enrichment Services LLC License Application for the Eagle Rock Enrichment Facility

(TAC L32707)," 2009.

(NRC, 2002) U.S. Nuclear Regulatory Commission, NUREG-1520, "Standard Review Plan for

the Review of a License Application for a Fuel Cycle Facility," 2002.

11-40
CHAPTER 12.0 PHYSICAL PROTECTION

By letter, dated April 23, 2009, AREVA Enrichment Services, LLC (AES) submitted Revision 1

of the license application for review and approval by the U.S. Nuclear Regulatory Commission

(NRC) in accordance with Title 10 of the Code of Federal Regulations (CFR) 70.22(k). The staff

subsequently sent a Request for Additional Information (RAI), by letter, dated August 26, 2009

(Agencywide Documents Access and Management System (ADAMS) Accession

No. ML092190320), that contained 12 issues to be addressed by AES. AES provided a

response to the RAI. By letter, dated January 20, 2010, AES submitted the Physical Security

Plan (PSP) Revision 2, which incorporated the responses to the RAI. The PSP also included

the Transportation Security Plan (TSP) for the staffs review.

12.1 Regulatory Requirements

The following regulatory requirements apply to physical protection at the proposed facility:

1. 10 CFR 73.67 governs licensee fixed site and in-transit requirements for the physical

protection of special nuclear material (SNM) of moderate and low strategic significance

(LSS).

2. 10 CFR 73.71 requires reporting of safeguards events.

3. 10 CFR 73.73 specifies the requirements for advance notice and protection of export

shipments of SNM of LSS.

4. 10 CFR 73.74 describes the requirement for advance notice and protection of import

shipments on nuclear material from countries that are not party to the Convention on the

Physical Protection of Nuclear Material.

12.2 Regulatory Guidance

The following guidance applies to physical protection at the proposed facility:

1. Regulatory Guide 5.59, "Standard Format and Content for a Licensee Physical Security Plan

for the Protection of Special Nuclear Material of Moderate or Low Strategic Significance."

(NRC, 1983)

2. NRC Regulatory Issue Summary 2005-22, "Requirements for the Physical Protection During

Transportation of Special Nuclear Material of Moderate and Low Strategic Significance:

10 CFR Part 73 vs. Regulatory Guide 5.59 (1983)." (NRC, 2005)

12.3 Staff Review and Analysis

The PSP and TSP were reviewed to ensure that the applicant has an acceptable understanding

of the security requirements as outlined in 10 CFR Part 73; and has committed to implementing

12-1
the general performance objectives of 10 CFR 73.67(a), the specific fixed site and in-transit

security requirements of 10 CFR 73.67(c), (f), and (g), and the notification requirements of

10 CFR 73.71, 10 CFR 73.73, and 10 CFR 73.74.

The PSP was reviewed to ensure that the applicant committed to storing SNM only in a

controlled access area, monitoring the protection of SNM by an intrusion detection system or

other approved procedures, maintaining both a guard force and offsite response force available

to respond to unauthorized penetrations or activities, establishing and maintaining response

procedures for dealing with threats of theft or theft of SNM, and reporting safeguards events as

required by regulation.

The TSP for the transportation of SNM of LSS was reviewed to ensure that the applicant

committed to providing advance notification, receiving confirmation from receivers prior to

shipping material; providing in-transit security; transporting the material in tamper-indicating,

sealed containers; verifying the integrity of seals on shipments received; notifying shippers of

receipt of shipment; and implementing the proper reporting and investigations in the event of

lost or unaccounted for material.

12.4 Evaluation Findings

The NRC staffs review of the applicant's PSP for the protection of SNM of LSS contains

information that has been marked as "Security-Related Information" by the applicant, pursuant

to 10 CFR 2.390. The methods and procedures as outlined in the PSP satisfy the performance

objectives, systems capabilities, and reporting requirements specified in 10 CFR 73.67 and

73.71. The PSP for the facility is acceptable and provides reasonable assurance that the

requirements for the physical protection of SNM of LSS will be met.

The NRC staffs review of the applicant's TSP for the transportation of SNM of LSS contains

information that has been marked as "Security-Related Information" by the applicant, pursuant

to 10 CFR 2.390. The NRC staff also reviewed the applicant's TSP for SNM of LSS shipments

originating from, or arriving at, the facility. The approaches and procedures as outlined in the

TSP satisfy the performance objectives, systems capabilities, and event and advance

notification requirements specified in 10 CFR 73.67(a), 10 CFR 73.67(f) and (g)(1)-(5),

10 CFR 73.71, 10 CFR 73.73, and 10 CFR 73.74. The NRC staff concludes that the facility

TSP is acceptable and provides reasonable assurance that the requirements for the physical

protection of SNM of LSS in transit will be met.

12-2
CHAPTER 13.0 SAFETY EVALUATION REPORT PREPARERS

The individuals and organizations listed below are the principal contributors to the preparation of

this Safety Evaluation Report. The U.S. Nuclear Regulatory Commission's (NRC's) staff

directed the effort and contributed to the technical evaluations. The staff also used contractor

support from the Center for Nuclear Waste Regulatory Analyses and ICF Consulting in the

preparation of this document.

U.S. NUCLEAR REGULATORY COMMISSION CONTRIBUTORS

Breeda Reilly, Office of

Nuclear Material Safety and Safeguards (NMSS)

NRC Project Manager

General Information and Organization and

Administration

Thomas Hiltz, NMSS

Mary Adams, NMSS

Branch Chief, AFCB

Decommissioning

Damaris Arroyo, NMSS

James Bongarra, Office of New Reactors

Gregory Chapman, NMSS

Jonathan De Jesus-Segarra, NMSS

Ira Dinitz, Office of

Nuclear Reactor Regulation (NRR)

James Downs, NMSS

Stan Echols, NMSS

Keith Everly, Office of

Nuclear Security and Incident Response (NSIR)

Herman Graves, Office of

Nuclear Regulatory Research

Shawn Harwell, NRR

Doug Hase, NSIR

Kenneth Kline, Office of Federal and State

Materials and Environmental Management

Programs (FSME)

Management Measures and Quality

Assurance

Human Factors Engineering

Health Physics, Accident Analysis, and

Environmental Protection

Chemical Safety

Liability Insurance

Fire Safety

Environmental Protection

Protection of Classified Matter

Seismic and Structural

Financial Qualification

Protection of Classified Matter

Decommissioning Financial Assurance

13-1
Mike Norris, NSIR

Tom Pham, NMSS

Blake Purnell, NMSS

Roman Przygodzki, FSME

David Rahn, NRR

Christopher Tripp, NMSS

Glenn Tuttle, NMSS

Steve Ward, NMSS

Rex Wescott, NMSS

Barry Wray, NSIR

CENTER FOR NUCLEAR WASTE REGULATORY

Asadul Chowdhury

Sui-Min (Simon) Hsiung

John Stamatakos

ICF CONSULTING CONTRIBUTORS

Craig Dean

Paul Bailey

Emergency Management

Material Control and Accountability

Nuclear Criticality Safety

Decommissioning Financial Assurance

Electrical Engineering and Instrumentation

and Controls

Nuclear Criticality Safety

Material Control and Accountability

Material Control and Accountability

Integrated Safety Analysis

Physical Protection

ANALYSES CONTRIBUTORS

Structural Design

Natural Phenomena

Seismic Engineering

Decommissioning Financial Assurance

Decommissioning Financial Assurance

13-2
APPENDIX A

INTEGRATED SAFETY ANALYSIS (ISA) AND ISA SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) staff's review of the applicant's ISA and ISA

Summary is documented in Chapter 3 and Appendix A to this safety evaluation report.

Appendix A contains information that has been marked by the applicant as "Security-Related

Information" and "Export Controlled Information," pursuant to Title 10 of the Code of Federal

Regulations (CFR) 2.390 and 10 CFR 810, respectively.

As discussed in Chapter 3 of this report, the NRC staff concluded that the applicant's

commitment to conduct and maintain an ISA is in accordance with the requirements of

10 CFR 70.62(c)(1). Based on its review of the ISA Summary, the staff also concludes that the

applicant provided an acceptable ISA Summary for the proposed facility that will meet the

requirements of 10 CFR 70.65(b)(1) through (9); 10 CFR 70.23(a); and 10 CFR 70.64 (b).

A-1

APPENDIX B

ACCIDENT ANALYSIS

The U.S. Nuclear Regulatory Commission's (NRC) staff independently evaluated the

consequences of a subset of potential accident sequences identified in the applicant's

Integrated Safety Analysis (ISA) Summary. This evaluation is documented in Appendix B to this

safety evaluation report. Appendix B contains information that has been marked by the

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

Regulations (CFR) 2.390.

A list of the six accident sequences that the staff evaluated is presented in Appendix B. The

selected accident consequences vary in severity from high- to low-consequence events, and

include accidents initiated by natural phenomena, operator error, and equipment failure. The

most significant consequences are associated with the release of uranium hexafluoride and

nuclear criticality.

The NRC staff concluded that the proposed Eagle Rock Enrichment Facility (EREF) design

would reduce the risk (likelihood) of these accidents by using items relied on for safety (IROFS)

and defense-in-depth measures. In addition, the facility Emergency Plan addresses these types

of events. The NRC staff independently verified the applicant's accident analysis by performing

confirmatory calculations and modeling. The NRC staff concluded that through the combination

of plant design, passive- and active-engineered IROFS, administrative IROFS, and defense-indepth features, the consequences of potential accidents at the proposed EREF will pose an

acceptably low safety risk to workers, the public, and the environment.

B-1

APPENDIX C

COST ESTIMATE TO CONSTRUCT AND OPERATE

The U.S. Nuclear Regulatory Commission staff's review of the applicant's cost estimate to

construct and operate the Eagle Rock Enrichment Facility is documented in Section 1.2.3.3.1 of

this report. Appendix C contains the estimated costs of the two phases of construction which

that have been marked by the applicant as "Proprietary Information," pursuant to Title 10 of the

Code of Federal Regulations (CFR) 2.390.

The NRC staff finds that, based on the financial information submitted in the application and

subsequent RAI response, AES meets the financial qualifications for the proposed activities in

accordance with 10 CFR 70.23(a)(5).

C-1

APPENDIX D

HUMAN FACTORS

The U.S. Nuclear Regulatory Commission's (NRC) staff independently evaluated the human

factors engineering. This evaluation is documented in Appendix D to this safety evaluation

report. Appendix D contains information that has been marked by the applicant as "SecurityRelated Information," pursuant to Title 10 of the Code of Federal Regulations (CFR) 2.390.

The NRC staff reviewed general design criteria and applicable standards related to humansystem interfaces for the proposed Eagle Rock Enrichment Facility (EREF). The applicant will

establish and maintain a process to design and implement the human-system interfaces for the

EREF. The staff concluded that the design and implementation of humans-system interface

activities will ensure that the human-system interfaces perform their intended functions, thereby,

meeting the regulatory requirements in 10 CFR 70.61(e).

D-1

APPENDIX E

ELECTRICAL SYSTEM AND INSTRUMENTATION AND CONTROL

The U.S. Nuclear Regulatory Commission (NRC) staff's review of the electrical system and

instrumentation and controls (I&C) is documented in Appendix E to this safety evaluation report.

Appendix E contains information that has been marked by the applicant as "Security-Related

Information," pursuant to Title 10 of the Code of Federal Regulations (CFR) 2.390.

The staff reviewed the electrical system and I&C and their application to items relied on for

safety (IROFS) needed to prevent or mitigate the identified accident sequences. This included

reviewing how the design of these systems address the baseline design criteria, general design

criteria, applicable Institute of Electrical and Electronics Engineers' standards, and applicable

American National Standards Institute's standards. The proposed Eagle Rock Enrichment

Facility (EREF) design reduces the risk (likelihood) of the accident by identifying IROFS and

defense-in-depth features. The NRC staff concluded that through the combination of plant

design, passive- and active-engineered IROFS, administrative IROFS, defense-in-depth

features, and the application of management measures, electrical system and I&C at the

proposed EREF will enable the facility to pose an acceptably low safety risk to workers, the

public, and the environment. The staff reviewed the ISA summary and other information as it

pertains to the electrical power system and I&C used as IROFS, and finds that it provides

reasonable assurance that the applicant has identified appropriate IROFS and established

engineered and administrative controls to ensure compliance with the performance

requirements of 10 CFR 70.61.

E-1

APPENDIX F

STRUCTURAL DESIGN

The U.S. Nuclear Regulatory Commission (NRC) staff independently evaluated the protection of

building structures against natural phenomena in the applicant's Integrated Safety Analysis

Summary. This evaluation is documented in Appendix F to this safety evaluation report.

Appendix F contains information that has been marked by the applicant as "Security-Related

Information," pursuant to Title 10 of the Code of Federal Regulations (CFR) 2.390.

The NRC staff reviewed the structural design of the proposed facility. This included reviewing

the building designs and structural design criteria, bases, and methodology. The proposed

design criteria, bases, and methodology are based on industry-accepted codes, standards, and

procedures. As a result, the staff determined that the applicant meets the requirements in

10 CFR 70.64(a)(2), 70.64(a)(4) and 70.64(b), as they apply to structural design.

F-1

APPENDIX G

NUCLEAR CRITICALITY SAFETY

The U.S. Nuclear Regulatory Commission (NRC) staffs Nuclear Criticality Safety (NCS) review

of the applicant's Integrated Safety Analysis (ISA) Summary is documented in Appendix G to

this safety evaluation report. Appendix G contains information that has been marked by the

applicant as "Security-Related Information" and "Export Controlled Information," pursuant to

Title 10 of the Code of Federal Regulations (CFR) 2.390 and 10 CFR 810, respectively.

The staff reviewed the ISA Summary and selected ISA documents related to NCS. Based on

the review, the staff determined that the applicant's ISA met the applicable requirements of

10 CFR 70.70 through 70.65, as they apply to criticality safety.

G-1

APPENDIX H

MATERIAL CONTROL AND ACCOUNTING

The U.S. Nuclear Regulatory Commission (NRC) staff's review of the applicant's nuclear

Material Control and Accountability (MC&A) program is documented in Appendix H to this safety

evaluation report. Appendix H contains information that has been marked by the applicant as

"Security-Related Information," pursuant to Title 10 of the Code of Federal Regulations (CFR)

2.390.

The NRC staff concluded that the applicant provided an acceptable Fundamental Nuclear

Material Control (FNMC) Plan for the proposed facility that will meet the applicable

10 CFR Part 74 requirements. The FNMC Plan 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 10 CFR Part 74.

H-1

NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

(9-2004) (Assigned by NRC, Add Vol., Supp., Rev.,

NRCMD 3.7 and Addendum Numbers, if any.)

BIBLIOGRAPHIC DATA SHEET

(See instructions on the reverse) NUREG-1951

2. TITLE AND SUBTITLE 3. DATE REPORT PUBLISHED

Safety Evaluation Report for the Eagle Rock Enrichment Facility in Bonneville County, Idaho MONTH YEAR

September 2010

4. FIN OR GRANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT

Final

7. PERIOD COVERED (inclusive Dates)

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address; if contractor,

provide name and mailing address.)

Division of Fuel Cycle Safety and Safeguards

Office of Nuclear Material Safety and Safeguards

U.S, Nuclear Regulatory Commission

Washington, DC 20555-0001

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (if NRC, type "Same as above"; if contractor, provide NRC Division, Offce or Region, U. S. Nuclear Regulatory Commission,

and mailing address.)

Same as above.

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words or less)

This report documents the U.S. Nuclear Regulatory Commission (NRC) staff's review and safety and safeguards evaluation of

the AREVA Enrichment Services LLC (AES) application for a license to construct a gas centrifuge uranium enrichment facility

and possess and use byproduct material, source material, and special nuclear material (SNM). The proposed facility is known as the Eagle Rock Enrichment Facility (EREF). AES proposes that the EREF be located in Bonneville County, Idaho, about 32

kilometers (20 miles) west northwest of the city of Idaho Falls. The EREF will possess natural, depleted, and enriched uranium,

and will be authorized to enrich uranium up to a maximum of 5 weight percent uranium-235.

The objective of the NRC's review is to evaluate the facility's potential adverse impacts 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 management organization, administrative programs, and financial

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

The NRC 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.

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report.) 13. AVAILABILITY STATEMENT

nuclear fuel cycle unlimited

uranium enrichment 14. SECURITY CLASSIFICATION

centrifugation (This Page)

AREVA Enrichment Services LLC unclassified

Safety Evaluation (This Report)

unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (9-2004) PRINTED ON RECYCLED PAPER
FPdnl Rncycllng Progrm

R EG&j.,RE'r UNITED STATES

NUCLEAR REGULATORY COMMISSION

0 WASHINGTON, DC 20555-0001

.0 OFFICIAL BUSINESS